[Federal Register Volume 62, Number 68 (Wednesday, April 9, 1997)]
[Notices]
[Pages 17223-17252]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-8916]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving no Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 17, 1997 through March 28, 1997. The
last biweekly notice was published on March 26, 1997 (62 FR 14457).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed no Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By May 9, 1997, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert
[[Page 17224]]
opinion which support the contention and on which the petitioner
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. Petitioner
must provide sufficient information to show that a genuine dispute
exists with the applicant on a material issue of law or fact.
Contentions shall be limited to matters within the scope of the
amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner to relief. A petitioner who fails
to file such a supplement which satisfies these requirements with
respect to at least one contention will not be permitted to participate
as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: March 5, 1997.
Description of amendments request: The proposed amendments would
incorporate a new Technical Specification (TS) for instrumentation
associated with automatic isolation of a pathway for release of non-
condensible gases from the main condenser. At power levels of 5 percent
or less, mechanical vacuum pumps are used to remove non-condensible
gases from the condenser using a pathway to the release stack that
bypasses the normal holdup and filter train. The proposed TS will
require that four channels of the main steam line radiation--high
isolation function be capable of tripping the mechanical vacuum pumps
and closing an isolation valve in the release pathway. Surveillance
requirements are included in the TS to ensure the isolation
instrumentation will perform its intended function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change incorporates a new Technical Specification
3/4.3.8, ``Condenser Vacuum Pump Isolation Instrumentation.'' This
specification will require that the main steam line radiation--high
isolation function be capable of tripping the condenser vacuum
pump(s) and isolate the associated common isolation valve. Four
instrumentation channels of this function are required to be
operable when the unit is in OPERATIONAL CONDITION 1 or 2 with a
condenser vacuum pump in operation. Adding the requirement to trip
the condenser vacuum pumps does not affect the probability of an
accident previously evaluated. The probability of component failure
of the proposed design for condenser vacuum pump isolation devices
is the same as that of the original licensing basis. As a result,
the capability to isolate the condenser vacuum pump will not be
significantly impacted.
CP&L contracted Scientech-NUS to recalculate the main control
room doses resulting from a control rod drop accident assuming main
steam line radiation monitors isolate the condenser vacuum pump(s)
and determined the dose to be 23.2 rem thyroid and 0.05 rem whole
body, which is less than the General Design Criterion (GDC) 19/
Standard Review Plan (SRP) Section 6.4 limits of 30 rem thyroid and
5 rem whole body. The offsite doses at the exclusion area boundary
after 2 hours are 0.16 rem thyroid and 0.015 rem whole body, which
is less than the SRP Section 15.4.9 limits. The low population zone
(LPZ) dose is estimated to be about 1 rem thyroid, which is also
well below regulatory limits. Therefore, the proposed [amendments
do] not increase the consequences of an accident previously
evaluated.
2. The proposed amendment[s] would not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
The proposed [amendments add] new requirements to ensure the
capability to trip the condenser vacuum pump(s). The proposed
[changes do] not affect the operability of equipment designed to
mitigate the consequences of an accident nor [do they] create a
potential to initiate a new type of accident. Therefore, the
proposed [changes do] not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed license [amendments do] not involve a
significant reduction in a margin of safety.
The safety-related main steam line radiation monitors provide a
highly reliable means to detect radioactivity resulting from a
control rod drop accident and will provide automatic trip of the
condenser vacuum pumps and isolation of the associated isolation
valve. Use of the main steam line radiation monitors for this
application is consistent with the original Brunswick Steam Electric
Plant design for condenser pump and associated valve isolation. CP&L
contracted Scientech-NUS to recalculate the main control room doses
resulting from a control rod drop accident assuming main steam line
radiation monitors isolate the condenser
[[Page 17225]]
vacuum pump(s) and determined it to be 23.2 rem thyroid and 0.05 rem
whole body, which is less than the GDC 19/SRP Section 6.4 limits of
30 rem thyroid and 5 rem whole body. The offsite doses at the
exclusion area boundary after 2 hours are 0.16 rem thyroid and 0.015
rem whole body, which is less than the SRP Section 15.4.9 limits.
LPZ dose is estimated to be about 1 rem thyroid, which is also well
below regulatory limits. Therefore, the proposed [changes do] not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: February 18, 1997.
Description of amendment request: The proposed change revises the
Plant System Turbine Cycle Technical Specification (TS) 3/4.7.1 by
revising the power range high neutron flux setpoint values in TS Table
3.7-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The high flux setpoints are being revised to provide additional
margin against secondary side overpressurization for LOL/TT [loss-
of-load/turbine trip] events. The proposed revision will not create
any loss or reduction in redundancy or diversity in the reactor
protection systems that would increase the probability of a
previously evaluated accident. The high flux setpoints are being
revised to ensure that the consequences of a previously evaluated
accident do not increase.
Therefore, there would be no increase in the probability or
consequences of an accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No new or previously unanticipated failure mechanisms are
introduced by the proposed change. No new failure modes have been
created by the proposed change. No new credible event or initiating
factor is introduced. Reactor power is limited to ensure that the
secondary system is not overpressurized.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The margin of safety as defined in the basis of the Technical
Specification does not decrease. This change is proposed to ensure
that the secondary system pressure will be limited to within 110% of
its design pressure during the most severe anticipated operational
transient. The revised high flux setpoints are intended to bound the
allowable operating configurations of TS Table 3.7-1.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: February 21, 1997.
Description of amendment request: The proposed change adds a
definitive time limit to Technical Specification 3.3.2 in Action 16 of
Table 3.3-3 to place an inoperable channel into bypass.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change does not affect the operation or design of
the plant in any way. The requirement to place the channel into
bypass already exists and this change simply provides a specific
time limit. This logic circuit is not an initiator of any event and
with no change in logic or operation there is no change in
consequences.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed specific time limit does not involve any physical
alterations or additions to plant equipment or alter the manner in
which any safety-related system performs its function. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed change replaces an indeterminate time period with a
specific limit of six hours. Six hours is a reasonable period in
which to complete this requirement and is identical to the time
allowed for these functions in NUREG-1431 [Standard Specifications
Westinghouse Plants]. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: February 27, 1997.
Description of amendment request: The proposed change adds sleeve
installation as an alternative to tube plugging for repairing degraded
steam generator tubes to Technical Specification 3/4.4.5, Steam
Generators.
[[Page 17226]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The only equipment affected by sleeving is the steam generator
tubes. The most severe malfunction of a steam generator tube is a
tube rupture. The consequences of a ruptured sleeve are no greater
than the consequences of a ruptured tube. Sleeving does not increase
the probability of a steam generator tube failure because the
sleeved tube has been shown to have a significant safety factor for
burst and collapse pressures as well as demonstrated acceptable
resistance to corrosion and fatigue loading. Thus, a steam generator
with sleeved tubes would perform in the same manner as one without
sleeved tubes.
A sleeved tube is functionally equivalent to an unsleeved tube
except for less effective heat transfer due to the air gap and a
slightly higher pressure drop due to the primary flow restriction.
These differences are bounded by the current tube plugging limits.
Analysis and testing have demonstrated that the sleeves are
structurally adequate to withstand the load existing within the
steam generator tubes whether the original tube is still intact or
is breeched.
There is no increase in the possibility for increased fatigue
loadings. There is no possibility for the sleeve to become dislodged
from its plugging location and enter the RCS [Reactor Coolant
System] flow path.
The plant safety analysis for tube plugging bounds tube
sleeving.
The proposed change has no significant effect on the
configuration of the plant. The proposed change does not affect the
way in which the plant is operated. Therefore, there would be no
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
A sleeved tube is functionally equivalent to an unsleeved tube
except for less effective heat transfer due to the air gap and a
slightly higher pressure drop due to the primary flow restriction.
These differences are bounded by the current tube plugging limits.
The sleeved tube has been shown to have a significant safety
factor for burst and collapse pressures as well as demonstrated
acceptable resistance to corrosion and fatigue loading. Thus, a
steam generator with sleeved tubes would perform in the same manner
as one without sleeved tubes.
The proposed change has no significant effect on the
configuration of the plant. The proposed change does not affect the
way in which the plant is operated. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed revision to permit the installation of tube sleeves
does not reduce the margin of safety as presently defined in
Technical Specification BASES section 3/4.4.5. This margin of safety
includes primary to secondary leakage limits and tube plugging
limits which are not changed by the proposed amendment. The analyses
and testing of the proposed sleeve design demonstrates that the
structural integrity of the RCS is maintained. Design of the tube
sleeve considers mechanical/structural aspects, water chemistry and
metallurgical aspects as well as thermal/hydraulic considerations.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: March 10, 1997.
Description of amendment request: The proposed changes to Technical
Specification 3.5.1 provide an optional method of meeting surveillance
requirements by allowing the use of instrument readings to meet
surveillance 4.5.1.1.a.1, and adds a new Action c to cover a condition
in which one accumulator has a boron concentration not within limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The accumulators are not initiators of any event and so the
probability of occurrence of an event is unaffected by either of the
proposed changes. The use of actual instrumentation readings to
comply with the surveillance does not change the function or
performance of the accumulators and thus does not affect any
accident consequences. The increase in the allowed time to restore
the boron concentration to within limits is consistent with allowed
out of service times for other Emergency Safeguards equipment.
It will not have a significant impact on subcriticality during
reflood. Therefore, there will be no increase in the consequences of
an accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to the accumulator specification do not
involve any physical alterations or additions to plant equipment or
alter the manner in which any safety-related system performs its
function. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed change to the surveillance requirement provides an
equivalent means of meeting the requirement. Since there is no
change in either the accumulator limits or the surveillance
frequency, there is no reduction in safety margin. The new Action c
to address returning the boron concentration of a single accumulator
to within limits allows an out of service time commensurate with the
times allowed for other Engineered Safeguards Features. The boron
concentration of one accumulator does not have a significant impact
on subcriticality during reflood and thus does not involve a
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: March 14, 1997.
Description of amendment request: The amendment will revise the
Final Safety Analysis Report to include the
[[Page 17227]]
evaluation of a spent fuel cask drop analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes described do not impact the probability of
occurrence of accidents previously analyzed. Removal of the valve
box covers and all but four of the cask closure head sleeve nuts has
no impact on accident initiators. Dose assessments using maximum
potential releases assuming failure of the spent fuel and
radionuclide release through the gap between the cask closure head
and the cask or damage to the valves show that no significant
increase in consequences of an accident previously evaluated would
occur. [Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.]
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Compromising the integrity of the cask by removing the valve box
covers and closure head sleeve nuts in preparation for unloading the
spent fuel from the cask does not create the possibility of a new
type of accident or equipment malfunction. No safety-related
equipment, safety function, or operations of plant equipment will be
altered as a result of this change. Therefore, the proposed changes
do not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in the margin of safety.
The NRC basis for acceptance of a spent fuel cask drop is
documented in Section 15.7.5 of the Safety Evaluation Report, NUREG-
1038, dated November 1983. It states, ``* * * no loss of cask
integrity is postulated to occur in the event of a drop, and the
staff concludes there will be no significant radiation released to
the environment. The radiological consequences will be less than a
small fraction of the 10 CFR 100 exposure guideline values.''
As described in the proposed change, even though complete cask
integrity may not be preserved in the event of a loaded cask drop
with the valve box covers removed or with only four, rather than 32,
closure head sleeve nuts installed, the radiological consequences
calculated using conservative assumptions were determined to be a
small fraction of the 10 CFR 100 values. Therefore, the proposed
change does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602.
NRC Project Director: Mark Reinhart, Acting.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: June 20, 1996, as supplemented by
letters dated December 30, 1996, and March 5, 1997.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) by incorporating NRC approved
thermal limit licensing methodology in the list of approved
methodologies used in establishing the fuel cycle specific thermal
limits. In addition, the proposed amendment would correct mirror
editorial items in the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the
operability of plant systems designed to mitigate those
consequences. Limits will be established consistent with NRC
approved methods to ensure that fuel performance during normal,
transient and accident conditions is acceptable. The proposed
Technical Specifications amendment reflects NRC approved SPC
methodology used to analyze normal operations, including anticipated
operational occurrences (AOOs), and to determine the potential
consequences of accidents.
Licensing Methods and Models
The proposed amendment is to support operation with NRC approved
fuel and licensing methods supplied from Siemens Power Corporation
[SPC]. In accordance with [Updated Final Safety Analysis Report]
UFSAR Chapter 15, the same accidents and transients will be analyzed
with the new fuel and methods. The latest NRC approved revision to
the Siemens [loss-of-coolant accident] LOCA analysis methodology
(Reference: ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation
Methodology for Boiling Water Reactors EXEM BWR Evaluation Model)
will be used to evaluate the ATRIUM-9B and other co-resident fuel
types. The other licensing analysis methods and models are also NRC
approved. The approved methods and models are used to determine the
fuel thermal limits (e.g., average planar linear heat generation
rate, transient linear heat generation rate, minimum critical power
ratio and linear heat generation rate). The SPC core monitoring code
enables the site to monitor keff as well as control rod density
to perform the reactivity anomaly surveillance. Therefore, the
change in licensing analysis methods and models does not
significantly increase the probability of an accident or the
consequences of an accident previously identified. The support
systems for minimizing the consequences of transients and accidents
are not affected by the proposed amendment.
New Fuel Design
The use of reload quantities of ATRIUM-9B fuel at Dresden does
not involve a significant increase in the probability or
consequences of any accident previously evaluated in the [Final
Safety Analysis Report] FSAR. The ATRIUM-9B fuel is generically
approved for use as a reload BWR fuel type (Reference: ANF-89-
014(P)(A) Revision 1 Supplement 1, Generic Mechanical Design for
Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). Limiting
postulated occurrences and normal operation have been analyzed using
NRC-approved methods for the ATRIUM-9B fuel design to ensure that
safety limits are protected and that acceptable transient and
accident performance is maintained.
The reload fuel has no adverse impact on the performance of in-
core neutron flux instrumentation or CRD response. The ATRIUM-9B
fuel design will not adversely affect performance of neutron
instrumentation nor will it adversely affect the movement of control
blades relative to the current Dresden fuel type, the Siemens
manufactured 9x9-2. The exterior dimensions of the ATRIUM-9B fuel
have been evaluated by ComEd; the ATRIUM-9B fuel design provides
adequate clearances relative to the co-resident 9x9-2 fuel. Thus, no
increased interactions with the adjacent control blade or nuclear
instrumentation are created. Additionally, given the above mentioned
overall envelope similarities, no problems are anticipated with
other station equipment such as the fuel storage racks, the new fuel
inspection stand and the spent fuel storage pool fuel preparation
machine. Therefore, the probability of adverse interactions between
the ATRIUM-9B fuel and components in the core and fuel handling
equipment is not significantly increased.
The ATRIUM-9B design is neutronically compatible with the
existing fuel types and core components in the Dresden core. SPC
tests have demonstrated that the ATRIUM-9B fuel design is
hydraulically compatible with the co-resident 9x9-2 fuel. The bundle
pressure drop characteristics of the ATRIUM
[[Page 17228]]
9B bundle are similar to those of the 9x9-2 fuel design, hence core
thermal-hydraulic stability characteristics are not adversely
affected by the ATRIUM-9B design. Cycle stability calculations are
performed by SPC. Therefore, the probability of thermal hydraulic
instability is not significantly increased.
Evaluations of the Dresden Emergency Procedures and UFSAR
Chapter 15 AOOs are being performed to ensure that the use of the
ATRIUM-9B fuel at Dresden does not alter any assumptions previously
made in evaluating the radiological consequences of an accident at
Dresden Units 2 and 3. Therefore, the radiological consequences of
accidents are not significantly increased.
Methods approved by the NRC are being used in the evaluation of
fuel performance during normal and abnormal operating conditions.
The ComEd and SPC methods to be used for the cycle specific
transient analyses have been previously NRC approved. The proposed
methodologies are administrative in nature and do not significantly
affect any accident precursors or accident results; as such, the
proposed change to the listing of the SPC methodologies for Dresden
does not significantly increase the probability or consequences of
any previously evaluated accidents.
The description of the fuel is modified to include the water box
design of the NRC approved ATRIUM-9B fuel type.
Review of the above concludes that the probability of occurrence
and the consequences of an accident previously evaluated in the
safety analysis report have not been significantly increased.
* * * * *
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated:
Creation of the possibility of a new or different kind of
accident would require the creation of one or more new precursors of
that accident. New accident precursors may be created by
modifications of the plant configuration, including changes in
allowable modes of operation.
Licensing Methods and Models
The proposed Technical Specification amendment reflects
previously approved SPC methodology used to analyze normal
operations, including AOOs, and to determine the potential
consequences of accidents. In accordance with FSAR Chapter 15, the
same accidents and transients will be analyzed with the new fuel and
method as have been previously performed. As stated above, the
proposed changes do not permit modes of reactor operation which
differ from those currently permitted; therefore, the possibility of
a new or different kind of accident is not created. Plant support
equipment is not affected by the proposed changes; therefore, no new
failure modes are created.
New Fuel Design
The basic design concept of a 9x9 fuel pin array with an
internal water box has been used in various lead assembly programs
and in reload quantities in Europe since 1986. WNP-2 has loaded
reload quantities since 1991. Eight lead ATRIUM-9B assemblies were
loaded into Dresden 2 during Cycle 15. Approximately 650 water box
assemblies have been irradiated in the United States through 1995,
with a substantially higher number being irradiated overseas. The
NRC has reviewed and approved the ATRIUM-9B fuel design (Reference:
ANF-89-014(P)(A) Revision 1 Supplement 1, Generic Mechanical Design
for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). The
similarities in fuel design and operation between the ATRIUM-9B and
the 9x9-2, and the previous Boiling Water Reactor experience with
Siemens fuel, indicate there would be no new or different types of
accidents for Dresden than have been considered for the existing
fuel. Therefore, the use of ATRIUM-9B fuel at Dresden does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
* * * * *
3. Involve a significant reduction in the margin of safety for
the following reasons:
The existing margin to safety is provided by the existing
acceptance criteria (e.g., 10 CFR 50.46 limits). The proposed
Technical Specification amendment reflects previously approved SPC
methodology used to demonstrate that the existing acceptance
criteria are satisfied. The revised LOCA methodology has been
previously reviewed and approved by the USNRC for application to
reload cores of BWRs. References for the Licensing Topical Reports
which document this methodology, and include the Safety Evaluation
Reports prepared by the USNRC, are added to the Reference section of
the Technical Specifications as part of this amendment.
Licensing Methods and Models
The proposed amendment does not involve changes to the existing
operability criteria. NRC approved methods and established limits
(implemented in the COLR) ensure acceptable margin is maintained.
The ComEd and SPC reload methodologies for the ATRIUM-9B reload
design are consistent with the Technical Specification Bases. The
Limiting Conditions for Operation are taken into consideration while
performing the cycle specific and generic reload safety analyses.
USNRC approved methods are listed in Specification 6.9.A of the
Technical Specifications.
Analyses performed with USNRC-approved methodology have
demonstrated that fuel design and licensing criteria will be met
during normal and abnormal operating conditions. The same margins of
safety will continue to be utilized by SPC (e.g., limits on peak
cladding temperature, cladding oxidation, plastic strain).
Therefore, there is not a significant reduction in the margin of
safety.
New Fuel Design
The exterior dimensions of the ATRIUM-9B fuel assembly result in
equivalent clearances relative to the co-resident 9x9-2 fuel. Thus,
no increased interactions with the adjacent control blade and
nuclear instrumentation are created. The change does not adversely
impact equipment important to safety; therefore the margin of safety
is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of amendment request: March 18, 1997.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TS) by increasing the High
Pressure Coolant Injection (HPCI) isolation setpoint from greater than/
equal to 80 psig to greater than/equal to 100 psig. The licensee has
requested the change to ensure consistency between the Updated Final
Safety Analysis Report (UFSAR), design basis documents and the TS. The
function of the setpoint is to assure the HPCI turbine steam supply is
isolated in the event that the reactor scram supply pressure falls
below the stall pressure of the HPCI turbine and the system seals are
no longer effective in controlling the release of potentially
contaminated steam.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
The Low Reactor Pressure isolation of the HPCI steam supply
lines is provided to prevent damage to the HPCI turbine when the
reactor steam pressure has decreased below that required to provide
adequate motive force to operate the system. The steam supply
isolation low reactor pressure setpoint is not an assumed initiator
or contributor to any previously evaluated accident and therefore
this change does not involve an increase in the probability of an
accident previously evaluated at Dresden Station.
The Lower Reactor Pressure isolation of the HPCI steam supply
lines is described in the
[[Page 17229]]
plant safety analysis as a backup protection to other system and
facility design features which provide assurance that accident
transients will not result in failures of the system which
contribute significantly to the consequences of the initiating
accident. The low reactor pressure isolation signal provides backup
to other isolation signals to ensure isolation will occur,
minimizing the radiation dose as a result of steam leakage past the
turbine seals in the event of a locked rotor due to damage from
liquid carryover due to postulated swell in the reactor vessel.
These analyses assume the isolation function occurs at 100 psig,
and the proposed setpoint of greater than or equal to 100 psig is
consistent and conservative with respect to these assumptions.
Because the isolation function is not an accident initiator and the
revised setpoint ensures that the isolation function continues to
minimize radiological consequences, the consequences of any accident
previously evaluated is not increased by the proposed changes.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
The proposed change administratively increases the Low Reactor
Vessel Pressure trip setpoint which initiates HPCI isolation. This
change does not result in any new or different modes of operation.
The proposed change increases the setpoint at which the HPCI turbine
steam supply will be isolated as the reactor vessel pressure
decreases following a postulated accident. The proposed new setpoint
is conservative with respect to the existing TS limit, i.e. the new
limit of greater than or equal to 100 psig is consistent and
permitted by the existing limit of greater than or equal to 80 psig.
The change assures that the Trip Setpoint in the TS accurately
reflects the design basis and UFSAR described limits.
Because the proposed change does not result in any new modes of
plant operation and administratively increases the system isolation
setpoint in a conservative manner, the proposed change does not
create the possibility of a new or different kind of accident from
those previously evaluated.
3. Involve a significant reduction in the margin of safety
because:
The Trip Setpoint provides assurance that the HPCI turbine
cannot be operated with a steam supply pressure too low to drive the
turbine and pump. The isolation assures that the turbine does not
stall and minimizes the potential for the release of radioactivity
which results from steam leakage past the turbine seals. The
proposed change increases the setpoint, ensuring that the required
isolation occurs at a higher pressure which is more conservative,
i.e. it assures the turbine is isolated before the inlet steam
pressure falls to the stall pressure of the HPCI turbine and leakage
occurs. The greater than or equal to 100 psig limit is well below
the range of reactor vessel pressure for which HPCI is required to
perform its safety function. Therefore, the margin of safety
provided by the function of the HPCI isolation on low reactor vessel
pressure is increased by the proposed TS change, and this change
will not involve a reduction in the margin of safety.
As described, the proposed amendment for Dresden will not reduce
the availability of systems required to mitigate accident
conditions. Neither are new or significantly different modes of
operation proposed. Therefore, the proposed change does not involve
a significant reduction in the margin of safety.
Guidance has been provided in ``Final Procedures and Standards
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744,
for the application of standards to license change requests for
determination of the existence of significant hazards
considerations. This document provides examples of amendments which
are and are not considered likely to involve significant hazards
considerations.
This proposed amendment does not involve any irreversible
changes, a significant relaxation of the criteria used to establish
safety limits, a significant relaxation of the bases for the
limiting safety system settings or a significant relaxation of the
bases for the limiting conditions for operations. Therefore, based
on the guidance provided in the Federal Register and the criteria
established in 10 CFR 50.92(c), the proposed change does not
constitute a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
NRC Project Director: Robert A. Capra.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: December 27, 1995, as supplemented
September 4, October 18, and November 26, 1996.
Description of amendment request: The proposed amendment would
revise technical specifications (TS) related to electrical power
systems. The proposed changes include revisions to limiting conditions
for operation (LCO), LCO applicability and action statements, allowed
outage times (AOT), surveillance requirements (SR), and administrative
controls. The changes add new requirements, revise or delete existing
requirements, relocate certain existing requirements to other licensee
controlled documents, and editorially restructure the proposed
requirements to closely emulate the electrical power system
requirements of NUREG-1432, ``Standard Technical Specifications for
Combustion Engineering Plants,'' (STS). The proposed requirements
differ from the requirements of the STS where necessary to reflect
features unique to the Palisades design. Each proposed change has been
classified by the licensee as Administrative, Relocated, More
Restrictive, or Less Restrictive.
Basis for proposed no significant hazards consideration
determination: A proposed amendment to an operating license for a
facility involves no significant hazards consideration if operation of
the facility in accordance with the proposed amendment would not: (1)
Involve a significant increase in the probability or consequences of an
accident previously evaluated; (2) create the possibility of a new or
different kind of accident from any previously evaluated; or (3)
involve a significant reduction in a margin of safety. As required by
10 CFR 50.91(a), the licensee has provided its analysis of the issue of
no significant hazards consideration, which is presented below:
Evaluation of ADMINISTRATIVE, RELOCATED, and MORE RESTRICTIVE
changes:
ADMINISTRATIVE changes and RELOCATED changes move requirements,
either within the TS or to documents controlled under 10 CFR 50.59,
or [clarify] existing TS requirements, without affecting their
technical content. Since ADMINISTRATIVE and RELOCATED changes do not
alter the technical content of any requirements, they cannot involve
a significant increase in the probability or consequences of an
accident previously evaluated, create the possibility of a new or
different kind of accident from any previously evaluated, or involve
a significant reduction in a margin of safety.
MORE RESTRICTIVE changes only add new requirements, or revise
existing requirements to result in additional operational
restrictions. Since the TS, with all MORE RESTRICTIVE changes
incorporated, will still contain all of the requirements which
existed prior to the changes; MORE RESTRICTIVE changes cannot
involve a significant increase in the probability or consequences of
an accident previously evaluated, create the possibility of a new or
different kind of accident from any previously evaluated, or involve
a significant reduction in a margin of safety.
Evaluation of LESS RESTRICTIVE changes:
1. Do these LESS RESTRICTIVE changes involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Change 3 revised the requirement for operable AC sources, using
more general wording than the existing TS. The existing LCO requires
that two explicitly specified transformers be operable; the proposed
LCO requires that two qualified offsite circuits be operable. The
proposed LCO will allow
[[Page 17230]]
substitution of Safeguards Transformer 1-1 for Station Power
Transformer 1-2 as a required AC source, but the quantity and
quality of required offsite AC sources is unaffected. Since the
capability and qualification of Safeguards Transformer 1-1 are
equivalent to those of the Station Power transformer, neither the
probability or consequences of an accident previously evaluated will
be increased.
Change 10 is less restrictive only in its allowance of a 72 hour
AOT for an inoperable offsite source instead of the 24 hour AOT
currently required. The change also makes a considerably more
restrictive change by eliminating the allowance, based on submittal
of a report, for continuous operation with Startup Transformer 1-2
inoperable. Changing an AOT, alone, cannot increase the probability
or consequences of an accident previously evaluated.
Change 14 allows, for an inoperable DG [diesel generator],
verification that no common cause failure is involved in lieu of
test starting the other DG. The intent of the test starting
requirement is to verify that there is no common cause failure which
also makes the other DG inoperable. The proposed action statement
thereby accomplishes the same objective as that it replaces. Since
the proposed action statement accomplishes the same objective as the
one it replaces, operation in accordance with the proposed change
will not increase the probability or consequences of an accident
previously evaluated.
Change 21 revises the SR for the DG starting test. [The ``Less
Restrictive'' elements of the change eliminate the requirement to
vary use of the A and B starting circuits for each monthly test,
because the DG is not assumed to be single failure proof; and
eliminate requirements that the DGs be manually started and that
they be synchronized from the control room, because no practical
alternatives exist for accomplishing these actions]. The proposed
change does not alter any plant operating conditions, operating
practices, equipment settings, or equipment capabilities. Therefore,
operation of the facility in accordance with the proposed change
will not involve an increase in the probability of an accident.
Change 21 requires more rigorous testing of the DGs than required by
the existing Technical Specifications. The more rigorous testing is
intended to provide additional assurance that the DGs are capable of
performing their design function and should, therefore, involve a
reduction, rather than an increase, in the consequences of those
accidents previously evaluated.
Change 25 revises the SR for testing the fuel transfer system.
The proposed change does not alter any plant operating conditions,
operating practices, equipment settings, or equipment capabilities.
Therefore, operation of the facility in accordance with the proposed
change will not involve an increase in the probability of an
accident. The only ``Less Restrictive'' feature of proposed SR is
test interval extension from ``each month'' to ``each 92 days.''
Changing a surveillance frequency, alone, cannot increase the
probability or consequences of an accident previously evaluated.
Change 26 revises the station battery SRs. The proposed monthly
and quarterly battery SRs contain all of the test requirements of
the existing SRs with two exceptions: (1) The proposed interval for
measuring each cell voltage is ``each 92 days'' instead of the
existing ``every month'' and (2) the requirement to record the
amount of water added has been deleted. Changing a surveillance
frequency or deleting a maintenance record cannot increase the
probability or consequences of an accident previously evaluated.
2. Do changes create the possibility of a new or different kind
of accident from any previously evaluated?
Change 3 only involves the specified offsite power sources.
Since the Loss of Offsite Power is already considered in the
accident analyses, operating the facility in accordance with Change
3 will not create the possibility of a new or different kind of
accident from any previously evaluated.
Change 10 revises an AOT; Change 14 revises a required action;
Change 21 revises a testing requirement; Changes 25 and 26 revise a
surveillance interval; and Change 26 deletes the requirement for a
maintenance record. None of these proposed changes alter any plant
operating conditions, operating practices, equipment settings, or
equipment capabilities. Therefore, operation of the facility in
accordance with the proposed changes will not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do changes involve a significant reduction in a margin of
safety?
Change 3 does not alter the quantity or quality of offsite
sources required to be available. Therefore, operating the facility
in accordance with the proposed change will not involve a reduction
in a margin of safety.
Change 10 revises an AOT; Change 14 revises a required action,
Change 21 revises a testing requirement; Changes 25 and 26 revise a
surveillance interval; and Change 26 deletes the requirement for a
maintenance record. These proposed changes do not alter any plant
operating conditions, operating practices, equipment settings, or
equipment capabilities. Therefore, operating the facility in
accordance with the proposed change will not involve a reduction in
a margin of safety.
The licensee's September 4, 1996, supplement stated that three of
the proposed changes contained in the supplement were not addressed in
the December 27, 1995, no significant hazards analysis. The changes
involved TS requirements that would be deleted. Equivalent requirements
would be incorporated in the FSAR or other documents subject to the
controls of 10 CFR 50.59. The licensee's analysis of the issue of no
significant hazards consideration for these changes is presented below:
1. Do changes which relocate a requirement from the TS to
documents which are controlled under 10 CFR 50.59 involve a
significant increase in the probability or consequences of an
accident previously evaluated?
10 CFR 50.59 specifically prohibits [without obtaining prior NRC
review and approval] changes to the facility as described in the
safety analysis report, and to procedures described in the safety
analysis report ``if the probability of occurrence or the
consequences of an accident or malfunction of equipment important to
safety previously evaluated in the safety analysis report may be
increased''. Since the conditions which limit changes performed
under 50.59 are more restrictive than the conditions which define
changes considered to involve a significant hazards consideration,
relocation of a requirement from the TS to the FSAR [Final Safety
Analysis Report] or to documents which are referenced by the FSAR
cannot involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Do changes which relocate a requirement from the TS to
documents which are controlled under 10 CFR 50.59 create the
possibility of a new or different kind of accident from any
previously evaluated?
10 CFR 50.59 specifically prohibits [without obtaining prior NRC
review and approval] changes to the facility as described in the
safety analysis report, and to procedures described in the safety
analysis report ``if a possibility for an accident or malfunction of
a different type than any evaluated previously in the safety
analysis report may be created''. Since the conditions which limit
changes performed under 50.59 are more restrictive than the
conditions which define changes considered to involve a significant
hazards consideration, relocation of a requirement from the TS to
the FSAR or to documents which are referenced by the FSAR cannot
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Do these changes which relocate a requirement from the TS to
documents which are controlled under 10 CFR 50.59 involve a
significant reduction in a margin of safety?
10 CFR 50.59 specifically prohibits [without obtaining prior NRC
review and approval] changes to the facility as described in the
safety analysis report, and to procedures described in the safety
analysis report ``if the margin of safety as defined in the basis
for any technical specification is reduced''. Since the conditions
which limit changes performed under 50.59 are more restrictive than
the conditions which define changes considered to involve a
significant hazards consideration, relocation of a requirement from
the TS to the FSAR or to documents which are referenced by the FSAR
cannot involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analyses and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
[[Page 17231]]
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: John N. Hannon.
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: March 10, 1997.
Description of amendment request: The proposed amendment would
modify Unit 1 Technical Specification (TS) 5.2.1 to add ZIRLO as fuel
assembly material and add reference to Nuclear Regulatory Commission
approved Topical Report, WCAP-12610, ``Vantage+ Fuel Assembly Reference
Core Report'', to TS 6.9.1.12 for both units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The methodologies used in the accident analyses have been
modified to reflect the requirements provided in WCAP-12610,
VANTAGE+ Fuel Assembly Reference Core Report. Reference to this NRC
approved ZIRLO topical report has been added to Specification
6.9.1.12, for both units to ensure the analytical methods used to
determine the core operating limits are consistent with those
previously approved by the NRC. The proposed changes do not change
or alter the design assumptions for the systems or components used
to mitigate the consequences of an accident. Use of ZIRLO fuel rod
material does not adversely affect fuel performance or impact
nuclear design methodology. Therefore, accident analysis results are
not impacted.
The operating limits will not be changed and the analysis
methods to demonstrate operation within the limits will remain in
accordance with NRC approved methodologies. Other than the changes
to the fuel assemblies, there are no physical changes to the plant
associated with this technical specification change. A safety
analysis will continue to be performed for each cycle to demonstrate
compliance with all fuel safety design bases.
VANTAGE 5H fuel assemblies with ZIRLO fuel rods meet the same
fuel assembly and fuel rod design bases as other VANTAGE 5H fuel
assemblies. In addition, the 10 CFR 50.46 criteria are applied to
the ZIRLO fuel rods. The use of these fuel assemblies will not
result in a change to the reload design and safety analysis limits.
Since the original design criteria are met, the ZIRLO fuel rods will
not be an initiator for any new accident. The fuel rod material is
similar in chemical composition and has similar physical and
mechanical properties as Zircaloy-4. Thus, the fuel rod integrity is
maintained and the structural integrity of the fuel assembly is not
affected. ZIRLO improves corrosion performance and dimensional
stability. No concerns have been identified with respect to the use
of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel
rods.
The dose predictions in the safety analyses are not sensitive to
the fuel rod material used; therefore, the radiological consequences
of accidents previously evaluated in the safety analysis remain
valid. A reload analysis is completed for each cycle, in accordance
with NRC approved methodologies. Therefore, the proposed change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
VANTAGE 5H fuel assemblies with ZIRLO fuel rods satisfy the same
design bases as those used for other VANTAGE 5H fuel assemblies. All
design and performance criteria continue to be met and no new
failure mechanisms have been identified. The ZIRLO fuel rod material
offers improved corrosion resistance and structural integrity.
The proposed changes do not affect the design or operation of
any system or component in the plant. The safety functions of the
related structures, systems, or components are not changed in any
manner, nor is the reliability of any structure, system, or
component reduced. The changes do not affect the manner by which the
facility is operated and do not change any facility design feature,
structure, or system. No new or different type of equipment will be
installed. Since there is no change to the facility or operating
procedures, and the safety functions and reliability of structures,
systems, or components are not affected, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The use of Zircaloy-4, ZIRLO, or stainless steel filler rods in
fuel assemblies will not involve a significant reduction in the
margin of safety because analyses using NRC approved methodology
will be performed for each configuration to demonstrate continued
operation within the limits that assure acceptable plant response to
accidents and transients. These analyses will be performed using NRC
approved methods that have been approved for application to the fuel
configuration.
Use of ZIRLO as fuel rod material does not change the VANTAGE 5H
reload design and safety analysis limits. The use of these fuel
assemblies will take into consideration the normal core operating
conditions allowed in the technical specifications. For each reload
core, the fuel assemblies will be evaluated using NRC approved
reload design methods, including consideration of the core physics
analysis peaking factors and core average linear heat rate effects.
Based on the above, it is concluded that the proposed license
amendment request does not result in a significant reduction in
margin with respect to plant safety as defined in the UFSAR [Updated
Final Safety Analysis Report] or any plant technical specification
BASES.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 17, 1996.
Description of amendment request: The proposed amendment would
reflect that the name of Louisiana Power & Light Company, which is
licensed to own and possess Waterford 3, has been changed to Entergy
Louisiana, Inc.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The proposed change documents changing the legal name of the
company. The proposed change will not affect any other obligations.
The company will still own all of the same assets, they still serve
the same customers, and all existing obligations and commitments
will continue to be honored. Therefore, the proposed change will not
involve a significant increase in the probability or consequences of
any accident previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No
The administrative changes in the Operating License requirements
do not involve any change in the design of the plant. Therefore, the
proposed change will not create the possibility of a new or
different
[[Page 17232]]
kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No
The proposed change is administrative in nature and does not
reduce the level of safety imposed by any current requirement.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 16, 1996.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) action requirements 3.2.1 and 3.2.4
and their associated surveillance requirements to extend the allowable
time for the Core Operation Limit Supervisory System (COLSS) to be out
of service by monitoring for adverse trends in the linear heat rate
(LHR) and departure from nucleate boiling (DNBR) limits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The proposed change does not modify the requirement to operate
within the alternate LHR and DNBR limits nor does it modify the
actual LHR or DNBR limits themselves. In the case of exceeding a
COLSS calculated [power operating limit] POL, Entergy agrees that
corrective action should be initiated promptly to bring the LHR and
DNBR within their respective limits and, in this case, a 15 minute
time limit is appropriate. However, in the case of exceeding a [core
protection calculator] CPC calculated operating limit following the
loss of COLSS, it is clear that simply because COLSS execution was
lost does not mean that the plant is operating outside the range of
conditions assumed in the Chapter 15 Safety Analysis and, in this
case, a 15 minute time limit is not appropriate. An increase from 2
hours to 8 hours to regain the monitoring capabilities of COLSS
would not significantly increase the probability of exceeding the
actual LHR or DNBR power operating limits since the increase in
COLSS out-of-service time will be compensated for by monitoring for
adverse trends of the important CPC calculated parameters (DNBR
Margin and LHR). Further, since the proposed change will result in
maintaining steady-state conditions while monitoring for adverse
trends, it will be easier for the operators to detect any abnormal
occurrence that has the potential to degrade either the LHR or the
DNBR.
The primary consideration in extending the COLSS out of service
time limit is the remote possibility of a slow, undetectable
transient that degrades the LHR and/or DNBR slowly over the 8 hour
period and is then followed by an [anticipated operational
occurrence] AOO or an accident. The parameters normally monitored by
COLSS which have the potential for degrading the LHR and DNBR if no
corrective action is taken are: Reactor Coolant System (RCS) flow
rate, axial and radial power distributions, core inlet temperature,
core power, RCS pressure and azimuthal tilt. Of these parameters,
core inlet temperature, core power, and RCS pressure are easily
monitored by the plant operators using various safety-grade,
Redundant Control Room indications and, therefore, changes in these
parameters are readily apparent. Further, operating experience at
Waterford 3 and other [Combustion Engineering] CE nuclear steam
supply systems using the same reactor coolant pumps (RCPs) as
Waterford has shown that measurable changes in RCP Ps
(which COLSS uses to calculate RCS flow) are very rare and when they
do occur, involve abrupt step changes in flow which are readily
apparent; hence, the probability of a slow degradation in the RCS
flow rate is exceedingly small. Thus, the parameters that
comparatively (although still remote) pose the most potential for a
degradation in the core thermal margin when COLSS is out of service
relate to the axial and radial core power distributions and the
azimuthal tilt. These parameters are discussed below.
Axial xenon oscillations are a normal consequence of the
Waterford 3 core design, particularly near the end of core life. As
a result, Waterford 3 operations personnel are instructed, per
operating procedure OP-10-001, General Plant Operations, to maintain
strict control over the axial power shape in the core. Although the
primary reason for axial shape control is to maintain an even fuel
burnup throughout the core, it also results in maintaining the axial
power shapes well within the limits assumed in the safety analysis.
Typically, axial shape control practiced at Waterford 3 maintains
the axial shape index (ASI) within 0.05 ASI units of the equilibrium
shape index (ESI), which is normally very near 0.0.
Hypothetically, the most severe situation which could be
postulated to occur, although again remote, would be if COLSS
execution was lost just when the plant operators were ready to take
manual action to return the ASI value to within the ESI + 0.05
control band. Since a full xenon oscillation takes approximately 26
hours, there would be about 6 hours from the time that control
action would normally be taken to the time that the ASI reached its
peak value (i.e., it takes one quarter cycle for the ASI to travel
from its ESI value to its peak value). Since abnormal operating
procedure OP-901-501, PMC or Core Operating Limit Supervisory System
Inoperable, will be revised to require the CPC calculated LHR and
DNBR trends to be monitored every 15 minutes (see below), any
significant change in the axial shape index will be apparent through
a change in these CPC calculated values. Hence, due to the attention
given the axial power distribution, both when COLSS is in service as
well as when COLSS is out of service it is very improbable that a
change in ASI during eight hours of steady-state operation with
COLSS out of service could be either undetected or lead to a
condition that placed the reactor outside the range of initial
conditions that were assumed in the safety analysis.
With regards to azimuthal tilt, there is very rarely any
significant change in this parameter as long as all [control element
assembly] CEAs are properly aligned. The only real contributor to a
rapid increase in azimuthal tilt would be an inadvertent CEA drop;
however, since the probability of a CEA drop is very low, the
likelihood of this event occurring within the eight hour time limit
is even lower. In the unlikely event that a CEA drop did occur, the
Control Element Assembly Calculators (CEACs) provide a safety-grade,
redundant means of alerting the operators that corrective action is
necessary. Thus, the potential for a degradation in azimuthal tilt
during eight hours of steady-state operation following the loss of
COLSS is both highly unlikely and relatively easy to detect using
instrumentation already available in the Control Room.
As previously stated, upon approval of the proposed change plant
personnel will revise abnormal operating OP-901-501, PMC or Core
Operating Limit Supervisory System Inoperable, to monitor for
adverse trends of the CPC calculated values of LHR and DNBR.
Currently, this procedure requires that the monitoring frequency for
LHR and DNBR be increased to once every 15 minutes on a loss of
COLSS.
Extending the time to restore the CPC calculated LHR and DNBR to
within the acceptable operating range from 2 hours to 8 hours is
being proposed to assure that COLSS can be restored thus decreasing
the probability of an avoidable challenge to the reactor protection
system (RPS) during a power reduction. It is possible that the
required power reductions may exceed 25% near the end of the fuel
cycle. These large power reductions result in a rapid increase in
xenon concentration, changes in ASI, and a subsequent decrease in
cold leg temperature (T-cold) that may be difficult to control.
Accordingly, given the potential for
[[Page 17233]]
power reductions of this magnitude, it is appropriate to extend the
time allowed to restore COLSS so that a power reduction may be
unnecessary.
Taken in total, the proposed changes will reduce the number of
potentially unnecessary power reductions by allowing more time for
COLSS to be restored along with the advantages of trend monitoring
in detecting an adverse trend expeditiously. The proposed change
will result in significant operational benefits while continuing to
maintain a high degree of confidence that the core conditions remain
well within the range of values assumed in the safety analysis.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No
The proposed change does not alter the current power operating
limits nor does it involve any changes to COLSS or CPC software.
There has been no physical change to plant systems, structures or
components nor will the proposed change affect the ability of any of
the safety-related equipment required to mitigate AOOs or accidents.
The only significant change associated with the proposed amendment
involves changes to the operating procedures used when COLSS is out-
of-service. All revisions to operating procedures will be reviewed
and approved by appropriate plant personnel as required by the
Administrative Controls (Section 6) in the Waterford 3 Technical
Specifications. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No
The intent of [limiting conditions for operation] LCOs 3.2.1 and
3.2.4 is to maintain the reactor within the range of initial
conditions that was assumed in the Safety Analysis. Maintaining the
LHR within the specified range ensures that in the event of a LOCA,
the fuel cladding temperature will not exceed the 2200 deg.F limit
imposed by 10CFR46 [10 CFR Part 46]. Maintaining the DNBR within the
specified range ensures that no AOO will result in a violation of
the [Specified Acceptable Fuel Design Limits] SAFDLs and that no
postulated accident will result in consequences more severe than
those described in Chapter 15 of the [Final Safety Analysis Report]
FSAR. Since there has been no change to the requirement to operate
the reactor within the LHR and DNBR limits and no change to the
actual LHR and DNBR limits themselves, the accident analyses
described in Chapter 15 of the FSAR will not be affected and will
therefore remain bounding.
The proposed change will reduce the number of potentially
unnecessary power reductions along with the rate at which the power
reductions are accomplished. Maintaining steady-state conditions for
up to eight hours after the loss of COLSS while monitoring the CPC
LHR/DNBR for trends, provides plant personnel with a reasonable
period of time to return COLSS to service while continuing to
maintain a high degree of confidence that the core conditions remain
well within the range of values assumed in the safety analysis. In
fact, monitoring for trends in LHR and DNBR Margin increases the
margin of safety by allowing the anticipation of degradation in LHR
or DNBR Margin. Moreover, by reducing the number of plant transients
there will be an attendant reduction in probability of an AOO and
subsequent RPS actuation. Therefore, the proposed change will not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400
L Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 16, 1996.
Description of amendment request: The following changes to the
Waterford Steam Electric Station, Unit 3, Technical Specifications are
proposed: 1) Relocation of certain administrative controls to the
Quality Assurance Program Manual (QAPM) as described in Nuclear
Regulatory Commission Administrative Letter 95-06, ``Relocation of
Technical Administrative Controls related to Quality Assurance''; 2)
Change shift coverage from 8-hour day, 40-hour weeks to an option of 8
or 12 hour days and nominal 40-hour weeks; 3) Make certain editorial
changes to the titles of certain organizational positions.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No
The conditions as they exist in the present Technical
Specifications do not have an affect on either the probability or
consequences of a previously evaluated accident. These changes also
will have no impact to increase either the probability or
consequences of a previously evaluated accident.
The proposed changes will have no affect on design basis
accidents nor will the change directly affect any material condition
of the plant that could directly contribute to causing or mitigating
the effects of an accident.
Therefore, the proposed changes will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No
The proposed changes will not alter the operation of the plant
or the manner in which it is operated. The changes do not involve a
design change and do not introduce any new failure modes.
Therefore, the proposed changes will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in margin of safety?
Response: No
The proposed changes are administrative in nature and affect
only Section 6.0 of the Technical Specifications. The Waterford 3
margins of safety are defined in Sections 2 through 5 and are
unaffected by these changes. Moving the reviews from the TS to the
QAPM will have no affect on the margin of safety because reviews
will still be performed. The only difference is the reviews will be
administratively controlled by the QAPM. The QAPM is controlled by
10CFR50.54 so no changes can be made which would lessen these
commitments (i.e., remove or reduce the requirement for procedure
reviews) without prior NRC approval.
Changing from an 8 hour to an 8 or 12 hour shift will not have
an adverse impact on personnel performance. The NRC study documented
in NUREG CR-4248 has identified that personnel errors have decreased
and productivity has increased where this change has been
implemented.
Therefore, the proposed changes will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
[[Page 17234]]
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 27, 1997.
Description of amendment request: The proposed change modifies
Technical Specification 3/4.5.2, ``ECCS Subsystems Modes 1, 2, and 3.''
The proposed change adds a surveillance requirement to verify the
Emergency Core Cooling System (ECCS) piping is full of water at least
once per 31 days. A change to the Technical Specification Basis 3/4.5.2
and 3/4.5.3 has been included to support this change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed change will not affect the assumptions, design
parameters, or results of any accident previously evaluated. The
proposed change does not add or modify any existing equipment. The
proposed change adds a new surveillance requirement which will
minimize the likelihood of a pressure transient occurring during
system startup and provide increased assurance that the ECCS will
perform its design basis function when needed. The new [low pressure
safety injection] LPSI and [high pressure safety injection] HPSI
vent valves which may be manipulated during this surveillance will
be administratively controlled and will be locked close when not in
use to prevent the possibility of a flow diversion. This
surveillance requirement is consistent with NUREG 1432.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
While new vent lines are being installed under 10CFR50.59, this
proposed change adds only a new surveillance requirement to
Technical Specification 3/4.5.2 and therefore does not involve
modifications to any existing equipment. The new vent valves, when
required, will be operated and controlled in the same manner as
existing LPSI and HPSI vent valves. The new LPSI and HPSI vent
valves will be administratively controlled and will be locked close
when not in use. This surveillance requirement is consistent with
NUREG 1432.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The functionality of ECCS is maintained such that it is capable
of performing its design function as assumed in the Updated Final
Safety Analysis Report. Verifying the ECCS is full of water at least
once per 31 days will minimize the likelihood of a pressure
transient occurring during system startup and provide increased
assurance that the ECCS will perform its design basis function when
needed. This surveillance requirement is consistent with NUREG 1432.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
Electric Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: March 27, 1997.
Description of amendment request: The proposed change modifies
Technical Specification (TS) surveillance requirements 4.5.2.d.3 and
4.5.2.d.4. The proposed change specifies granular trisodium phosphate
dodecahydrate (TSP), increases the minimum required amount of TSP that
is maintained in containment during power operation, and adjusts the
TSP sampling requirement accordingly. A change to the TS Basis 3/4.5.2
has been included to support this change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
Granular trisodium phosphate dodecahydrate is stored in the
containment lower level to raise the pH of the sump and spray water
following a LOCA. As the pH of the water increases, more radioactive
iodine is kept in solution and the amount of airborne radioactive
leakage is decreased. This also lessens the potential for boric acid
solution reacting with galvanized metal in containment to release
hydrogen. An additional advantage of a higher pH is the beneficial
reduction in chloride stress corrosion cracking of metal components
in the containment following an accident.
This chemical is an accident mitigator, not an accident
initiator in that it is not used until after an accident has
occurred. At the time it goes into solution, the accident has
occurred, containment spray has been activated and water has
collected in the sump. Therefore, increasing the Technical
Specification minimum amount verified to be in containment or
changing the sample solution and sample size will not involve a
significant increase in the probability of an accident previously
evaluated.
At the time TSP goes into solution, the accident has occurred,
containment spray has been activated and water has collected in the
containment sump. At Waterford 3, the iodine partition factor is a
constant 50% and does not vary with pH as allowed in the Standard
Review Plan (SRP) revision 1. The curve in SRP 6.5.2 revision 1
allows a partition factor of at least 50% for containment water at a
pH of 6.5 or less. The partition factor increases as pH rises. But,
the curve is based on sodium hydroxide which is much more reactive
than TSP. Therefore, increasing the Technical Specification minimum
amount verified to be in the containment, and corresponding sample
size, will not involve any significant increase in the consequences
probability of an accident because no credit is taken for reducing
the amount of volatized iodine normally associated with a 7.0 pH
solution.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Will operation of the facility in accordance with this
proposed change create the possibility of a new or different type of
accident from any accident previously evaluated?
Response: No.
The addition of more TSP does not represent a significant change
in the configuration or operation of the plant. Trisodium phosphate
dodecahydrate is currently present in the containment lower level.
Design Change 3491 which increases the storage capacity of the TSP
storage baskets was evaluated in accordance with 10 CFR 50.59 and
found not to involve an unreviewed safety question.
[[Page 17235]]
Boric acid acts as a buffer to prevent the pH from rising above
approximately 8.1 as TSP is dissolved. An internal study (EC-S96-013
revision 0) has shown that given the ``ratio of grams of TSP to
liters of 3000 ppm boron solution'' stays less than 5.6, TSP cannot
increase pH above 8.2. As pH increases, components composed of
aluminum, zinc, or copper become vulnerable to corrosion. Branch
Technical Position MTEB 6-1 implies that a solution pH greater than
7.5 enhances the chance for hydrogen generation as a result of
aluminum corrosion. Waterford 3 administratively limits the amount
of aluminum in containment to minimize the amount of hydrogen
expected during a DBA. Zinc is a component of the paint applied to
surfaces inside containment. The hydrogen recombiner design basis
includes 464 square feet (1040 pounds) of aluminum and 419,300
square feet (17,252 pounds) of metallic zinc. Estimates of the
amount of hydrogen produced by the aluminum assumes that the
corrosive agent is sodium hydroxide--a much more active chemical
than is TSP. Thus, the amount of hydrogen expected in the FSAR for
the hydrogen recombiner bounds what would actually be produced by
TSP even at a pH of approximately 8.1.
The 4.5.2.d.3 proposed TSP to boron ratio assures that pH cannot
rise above 8.1 as long as post accident in-containment boric acid
solution concentration is no greater than 3011 ppm boron and no less
than 1504 ppm boron. The main variable in post accident
concentration (the difference between 1504 and 3011) is the
concentration in the RCS at the time of the accident.
Therefore, the proposed change will not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
Trisodium phosphate dodecahydrate is stored in the containment
lower level to raise the pH of the sump and spray water following a
LOCA. As the pH of the water increases, more radioactive iodine is
kept in solution and the amount of airborne radioactive leakage is
decreased. A neutral pH also reduces the hydrogen generation from
the corrosion of the galvanized materials in containment. An
additional advantage of a higher pH is the beneficial reduction in
chloride stress corrosion cracking of metal components in the
containment following an accident.
Technical Specification 4.5.2.d.3 requires verification that a
minimum volume of TSP is contained in the storage baskets in
containment. Nine previous runs of surveillance requirement
4.5.2.d.4 (and similar tests) showed that the TSP actually used in
the plant properly neutralized a sample of water borated within RWSP
boron concentration limits. Boron concentrations of eight of the
sample solutions used in these tests ranged from 1753 ppm to 2217
ppm and resulted in a pH of 7.02 or greater. (The boron
concentration of one test performed in 1986 was unavailable.) The
ratio 4 grams to 4 liters is the amount of TSP needed to bring the
solution to a pH of at least 7.0 given that the solution is in the
1753 to 2217 ppm Boron range.
The amount of TSP in containment currently is adequate assuming
that RCS boric acid concentration stays below 454 ppm. However, the
fuel cycle is nearly over and a restart with a refreshed core would
require substantially more boric acid. We expect that the
containment water would reach approximately 2400 ppm under ideal
circumstances during cycle 9. During cycle 10, boron concentration
in containment could reach 3011 under those same ideal conditions.
As the maximum boron concentration increases, there is a non-linear
increase in the amount of TSP needed to raise solution pH to 7.0.
Thus, we request that the minimum amount of TSP in containment
required by 4.5.2.d.3 to be increased from 97.5 cubic feet to 380
cubic feet. This change also proposes to adjust the 4.5.2.d.4
specified increase that sample solution and the TSP sample size
accordingly. This change will ensure the safety injection
containment sump, when filled with water, will have an acceptable pH
following a LOCA. The test will not only demonstrate that TSP is in
the baskets but also shows that the amount of TSP in containment can
neutralize the solution expected in containment during any DBA.
Therefore, the proposed change will not involve a significant
reduction in a margin of safety. The amount of iodine kept in
solution during a DBA is limited to 50%. Note, the pH scale is
logarithmic so that the amount of TSP needed to raise pH to 7.0 is
more than three times the amount needed to reach 6.5. Furthermore,
the amount of hydrogen generated during a DBA is over estimated by
the analysis when it used sodium hydroxide as the corrosive agent
rather than TSP.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room Location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400
L Street N.W., Washington, D.C. 20005-3502.
NRC Project Director: William D. Beckner.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: May 17, 1996, as supplemented March 17,
1997.
Description of amendment request: The proposed amendment would
modify Technical Specification (TS) Section 3/4.4.5, Steam Generators,
3/4.4.6, Reactor Coolant System Leakage, and associated Bases to allow
the installation of tube sleeves as an alternative to plugging to
repair defective steam generator tubes. The proposed change would also
specify the Westinghouse topical reports to be used for sleeve design
and inspection, and identify the inspection sample size for repaired
tubes. This application was previously published in the Federal
Register on May 29, 1996, (61 FR 26938).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Listing the specific Westinghouse topical reports in the
TS binds the South Texas Project (STP) to the sleeve design and
inspection techniques identified in that revision of the topical
report. Any changes to sleeve design or inspection technique would
require a separate TS amendment.
New TS Table 4.4-3, Steam Generator Repaired Tube Inspection,
identifies the inspection sample size for steam generator tubes that
have already been repaired. This table simply identifies inspection
criteria and associated actions for repaired tubes and does not
increase the probability or consequences of an accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated. Implementation of laser welded sleeving maintains overall
tube bundle structural and leakage integrity conditions. Providing
specific Westinghouse topical report references in the TS only
serves to identify which sleeve design and inspection techniques are
being employed at STP. Likewise, the addition of Table 4.4-3
clarifies the expected inspection samples for previously repaired
tubes. The addition of Table 4.3-3 provides assurance that
previously repaired tubes will be inspected at regular intervals and
appropriate action taken if the tube is found defective. Neither of
these additions to the TS will create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety. Both of these changes are being added to
clarify the STP steam generator tube inspection program and provide
more specific detail regarding steam generator tube inspection
samples and inspection techniques. By requiring inspection of
previously repaired tubes, the margin of safety is increased rather
than decreased.
[[Page 17236]]
Based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment request involves no significant hazards
consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: William D. Beckner.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas, Docket
Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, Matagorda
County, Texas
Date of amendment request: January 28, 1997.
Description of amendment request: The proposed amendment would
relocate the details of Technical Specification (TS) Section 6.2.3 on
the Independent Safety Engineering Group (ISEG) from the Administration
Controls section of the TSs and place these details in the Updated
Final Safety Analysis Report (UFSAR) for South Texas Project, Units 1
and 2. This relocation is administrative only, and would not render any
changes to the existing plant philosophy toward the ISEG or any safety
analysis. Section 6.2.3 would be deleted from the TSs and removed from
the table of contents for Administrative Controls. Currently UFSAR
Section 13.4.2.2 describes the ISEG, but not in the detail as the
current TSs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes move details from the Technical
Specifications [TSs] to the Updated Final Safety Analysis Report
(UFSAR). The changes do not result in any hardware or operating
procedure changes. The details being removed from the Technical
Specifications [TSs] are not assumed to be an initiator of any
analyzed event. The UFSAR, which will contain the removed Technical
Specification [TS] details, will be maintained using the provisions
of 10 CFR 50.59 and is subject to the change control process in the
Administrative Controls Section of the Technical Specifications
[TSs]. [In addition] any changes to the UFSAR will be evaluated per
10 CFR 50.59, no increase in the probability or consequences of an
accident previously evaluated will be allowed without prior NRC
[Nuclear Regulatory Commission] approval. Therefore, the changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes move details from the technical
Specifications [TSs] to the Updated Final Safety Analysis Report
(UFSAR). The changes will not alter the plant configuration (no new
or different type of equipment will be installed) or make changes in
methods governing plant operation. The changes will not impose
different requirements, and adequate control of information will be
maintained. The changes will not alter assumptions made in the
safety analysis and licensing basis. Therefore, the changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes move detail from the Technical
Specifications [TSs] to the Updated Final Safety Analysis Report
(UFSAR). The changes do not reduce the margin of safety since the
relocation of details [is an administrative action and] has no
impact on any safety analysis assumptions. In addition, the detail
transposed from the Technical Specifications [TSs] to the UFSAR are
the same as the existing Technical Specification [TS] [6.2.3]. [In
addition] any future changes to the FSAR will be evaluated per the
requirements of 10 CFR 50.59, no reduction in a margin of safety
will be allowed without prior NRC approval. [Therefore, the licensee
concluded that the changes will not involve a significant reduction
in a margin of safety.]
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
NRC Project Director: William D. Beckner.
Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone
Nuclear Power Station, Unit No. 1, New London, Connecticut
Date of amendment request: February 7, 1997.
Description of amendment request: The proposed Technical
Specification changes would clarify and/or modify instrument
calibration, functional, and response time requirements for resistance
temperature detector and thermocouple testing. Also, certain
definitions would be clarified and/or modified using applicable wording
from NRC's NUREG-1433, ``Standard Technical Specifications,'' Revision
1, and industry recommendations. Additionally, the change would
relocate the reactor protection system logic response time value
utilizing the guidance provided by NRC's Generic Letter 93-08,
``Relocation of Technical Specification Tables of Instrument Response
Time Limits,'' with the exception of relocating the value to the
Technical Specifications Bases Section instead of the Updated Final
Safety Analysis Report. The proposed amendment is intended to clarify
instrumentation surveillance requirements, thereby helping to ensure
proper testing of safety-related components.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Pursuant to 10 CFR 50.92, NNECO [Northeast Nuclear Energy
Company] has reviewed the proposed changes and concludes that the
changes do not involve a significant hazards consideration (SHC)
since the proposed changes satisf[y] the criteria in 10 CFR
50.92(c). That is, the proposed changes do not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment continues to ensure the surveillance
requirements satisfy the licensing basis. The current TS [technical
specifications] definition for Instrument Functional Test requires
injection of a simulated signal into the primary sensor to verify
proper response. Current TS exempt the sensors of specific
instrument channels where it is not practical to include them within
the functional test boundaries. Some examples of these exemptions
include neutron monitoring system, turbine control valve fast
closure, and standby gas treatment initiation radiation monitors. In
these cases, TS permit the performance of the functional test by
injection of a simulated electrical signal into the measurement
channel. The proposed definition, which is consistent with the STS
[standard technical specifications]
[[Page 17237]]
definition, for CHANNEL FUNCTIONAL TEST requires injection of the
simulated signal ``as close to the sensor as practicable.''
Therefore, the proposed definition is consistent with the current TS
definition and its exemptions. The primary sensor is the transmitter
or switch or radiation monitor. The definition does not include
sensing elements such as radiation detectors, flow elements,
acceleration relays or reference legs.
This change will allow the channel functional test to be
performed by means of any series of sequential, overlapping, or
total channel steps and aligns this methodology with industry
practice. This change does not affect accident precursors and thus
does not involve a significant increase in the probability of an
accident previously evaluated. The proposed change will allow a
simulated or actual signal to be used to perform an Instrument or
Channel Functional Test. This change does not impose a requirement
to create an actual signal, nor does it eliminate any restriction on
producing an actual signal. While creating an ``actual'' signal
could increase the probability of an event, existing procedures (and
the 10 CFR 50.59 control of revisions to them) dictate the
acceptability of generating this signal. The proposed change does
not affect the procedures governing plant operations or the
acceptability of creating these signals; it simply would allow such
a signal to be utilized in evaluating the acceptance criteria for
the Instrument or Channel Functional Test requirements. Therefore,
the change does not involve a significant increase in the
probability of an accident previously evaluated. Because the method
of initiation will not affect the acceptance criteria of the
Instrument or Channel Functional Test, the change does not involve a
significant increase in the consequences of an accident previously
evaluated.
Minor word differences from STS are required to provide
consistency with current TS wording and support the current
licensing basis. These minor word differences including Industry/
TSTF [Technical Specification Task Force] Standard Technical
Specification Change Traveler (TSTF-64) do not alter the meaning of
instrument testing in the STS or change the current licensing basis.
Moving the RPS [Reactor Protection System] Logic Response Time
LCO [Limiting Condition of Operation] description to the TS
definition section is an administrative change and does not alter
the original intent or licensing basis.
Relocation of the RPS Logic Response Time value from the TS to
the Bases section involves the use of an alternate regulatory
process for controlling the instrument response time limit. The
change does not introduce any new modes of plant operation, make any
physical changes, alter any operational setpoints, or change the
surveillance requirements. Any change in the RPS logic response time
value would be evaluated pursuant to the requirements of 10 CFR
50.59.
The surveillance section editorial change does not alter the
meaning of surveillance applicability. Providing RPS Logic Response
Time surveillance frequency and applicable trip functions ensures
proper testing of RPS components and is consistent with industry
practice. An evaluation completed by GE [General Electric] verified
the applicable RPS trip functions that require a specific logic
response time using the current accident analysis as the basis. For
trip functions where no explicit credit is taken in the safety
analysis, the measurement of logic response time is not important,
and therefore, not warranted. In addition, we have concluded, that
instrumentation response time requirements (specified limits) other
than RPS logic are not important to test, especially considering the
long delays already accounted for in the accident analyses
associated with the start of emergency power sources, ECCS
[Emergency Core Cooling System] components, and containment
isolations, and that the non-RPS logic response times, including
response times of other instrumentation such as radiation monitors,
are not part of the Millstone Unit No. 1 licensing basis. The
sensors associated with all TS instrumentation are functionally
tested and calibrated to ensure proper operation.
No physical change is being made to instrument channels, or to
any systems or component that interfaces with the instrumentation
channels, therefore there is no change in the probability or
consequences of any accident analyzed in the UFSAR [Updated Final
Safety Analysis Report].
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change does not result in any design or physical
configuration changes to the instrumentation channels. Operation
incorporating the proposed change will not impair the
instrumentation channels from performing as provided in the design
basis.
Changing the TS to be consistent with current industry practice
adopted in STS will help to prevent unnecessary removal and
potential damage of the temperature detectors (for sensor
calibration). Clarification of RPS Logic Response Time testing
requirements consistent with the current licensing basis will ensure
proper testing of safety-related components.
Wording changes to Instrument Calibration and Functional Test
definitions do not involve a physical modification to the plant. The
injection of an actual or simulated signal as close to the sensor as
practical minimizes the likelihood of any transients.
Minor word differences from STS are required to provide
consistency with current TS wording and support the current
licensing basis. These minor word differences, including Industry/
TSTF Standard Technical Specification Change Traveler (TSTF-64), do
not alter the meaning of instrument testing in the STS or change the
current licensing basis.
Moving the RPS Logic Response Time LCO description to the TS
definition section is an administrative change and does not alter
the current licensing basis.
Relocation of the RPS Logic Response Time value involves the use
of an alternate process for controlling the instrument response time
limits. Therefore, the above change does not introduce any accident
initiators as it does not involve any new modes of plant operation,
make any physical changes, alter any operational setpoints, or
change the surveillance requirements.
The surveillance section editorial change does not alter the
meaning of surveillance applicability. Providing RPS Logic Response
Time surveillance frequency and applicable trip functions ensures
proper testing of RPS components and is consistent with industry
practice.
Since the proposed changes in the Technical Specifications do
not adversely impact the reliability of the RPS and other automatic
actuations, no new or different kind of accident is created.
3. Involve a significant reduction in a margin of safety.
Because the proposed change does not involve the addition or
modification of plant equipment, is consistent with the existing
Technical Specifications, current industry practices as outlined in
NUREG 1433, ``Standard Technical Specifications GE Plants, BWR/4,''
Revision 1, and with the current design and licensing basis of the
Protective Instrumentation systems including the accident analysis,
no action will occur that will involve a significant reduction in a
margin of safety.
The proposed change to allow the use of an actual signal in
addition to the existing requirement, which limits use to a
simulated signal, will not affect functional test acceptance
criteria. Therefore, the proposed change does not adversely affect
the reliability of the RPS or other automatic actuation and does not
involve a significant reduction in a margin of safety.
Relocation of the RPS Logic Response Time value from the TS to
the Bases section involves the use of an alternate regulatory
process for controlling the instrument response time limit. Any
change in the RPS logic response time value would be evaluated
pursuant to the requirements of 10 CFR 50.59.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis, and based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Deputy Director: Phillip F. McKee.
[[Page 17238]]
Northern States Power Company, Docket No. 50-263, Monticello
Nuclear Generating Plant, Wright County, Minnesota
Date of amendment request: November 25, 1996, as supplemented
December 12, 1996.
Description of amendment request: The proposed amendment would make
changes to Section 2.1.A for the Safety Limit Minimum Critical Power
Ratio (SLMCPR) and to Section 3.11.C for the Operating Limit Minimum
Critical Power Ratio (OLMCPR). The proposed change to Section 2.1.A
revises the SLMCPR value from 1.07 to 1.08 for two recirculation pump
operation and from 1.08 to 1.09 for single loop operation. The proposed
change to Section 3.11.C deletes the sentence that specifies the OLMCPR
limit penalty for single recirculation loop operation and adds a
statement that references the Core Operating Limits Report (COLR) as
the source for this information.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The basis of the MCPR [minimum critical power ratio] Safety
Limit calculation is to ensure that greater than 99.9% of all fuel
rods in the core avoid transition boiling if the limit is not
violated. The new SLMCPRs preserve the existing margin to transition
boiling and fuel damage in the event of a postulated accident. The
probability of fuel damage is not increased. The derivation of the
revised SLMCPRs for Monticello for incorporation into the Technical
Specification, and its [their] use to determine cycle-specific
thermal limits, have been performed using NRC-approved methods as
identified in Technical Specification 6.7.A.7.b. NSP [Northern
States Power] methodology established OLMCPR such that integrity of
the SLMCPR is maintained for the bounding analyzed transients.
Additionally, GENE [General Electric Nuclear Energy] interim
implementing procedures, which incorporate cycle-specific
parameters, have been used. Based on the use of these calculations,
the calculation of the revised SLMCPRs maintains the integrity of
the safety limits and therefore cannot increase the probability or
severity of an accident. The single loop OLMCPR evaluation was
performed using NSP methodology approved by the NRC. Relocating the
OLMCPR value to the COLR establishes appropriate control on a core
operating limit which may vary from cycle to cycle because it is
cycle dependent. Since OLMCPR is developed using procedures approved
in the Technical Specifications, placing the OLMCPR in the COLR
cannot result in a change not controlled by the Technical
Specifications. The change does not affect failure modes of
equipment, therefore, this amendment will not cause a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
analyzed.
The MCPR Safety Limit is a Technical Specification numerical
value, designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident. It
cannot create the possibility of any new type of accident. The new
SLMCPRs have been calculated using NRC-approved methods and the
OLMCPR values are more conservative. Additionally, interim
procedures, which incorporate cycle-specific parameters, have been
used. Therefore, the proposed Technical Specification change does
not create the possibility of a new or different kind of accident,
from any accident previously evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The MCPR Safety Limit is a Technical Specification numerical
value, designed to ensure that fuel damage from transition boiling
does not occur as a result of the limiting postulated accident.
Increasing the SLMCPR and OLMCPR values results in an increase in
the margin of safety to fuel failure, and does not affect other
plant systems. Therefore, the proposed Technical Specification
change does not involve a significant reduction in the margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
NRC Project Director: John N. Hannon.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: November 20, 1996, as supplemented by
letter dated February 20, 1997.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) to allow the Vice President to
designate the Safety Audit and Review Committee (SARC) Chairperson, to
change the work hours limitation in accordance with guidance in GL 82-
12, ``Nuclear Power Plant Staff Working Hours;'' to change radioactive
shipments record retention requirements to comply with recent 10 CFR
Part 20 changes; to revise position titles to reflect organizational
changes; and other editorial changes. The February 20, 1997,
supplemental letter differs from the November 20, 1996, application
which was noticed in the Federal Register on January 2, 1997 (62 FR
131), in that the previous application did not propose changes to TS
5.3, 5.5, 5.6, 5.7, and 5.11 reflecting recent organizational changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The changes requested are administrative in nature. Paragraph
3.D was placed in the License by Amendment No. 155 to authorize
Omaha Public Power District (OPPD) to increase the storage capacity
of the FCS spent fuel pool. Amendment No. 155 stated that the TS as
issued would be effective when the last new rack was installed.
Since the last new rack was installed on August 8, 1994, Paragraph
3.D is no longer necessary and should be deleted from the License.
Table of Contents, Section 6.0, ``Interim Special Technical
Specifications,'' Subsections 6.1 through 6.4 are proposed for
deletion because all of the Specifications referred to have been
deleted by previous Amendments.
The revision proposed for TS 2.15 (Item 2C of Table 2-3 & Item
1C of Table 2-4) will insert the correct terminology (Pressurizer
Low/Low Pressure) into the Functional Unit description.
The revision proposed for TS 5.2 will delete the specific
working hours as stated and relocate these requirements to the
Updated Safety Analysis Report (USAR). Overtime will remain
controlled by plant administrative procedures with the USAR
generally following the guidance of the NRC's Policy Statement on
working hours contained in Generic Letter 82-12, ``Nuclear Power
Plant Staff Working Hours.'' Specifying personnel working hours in
TS does not meet any of the four criteria contained in 10 CFR 50.36
for inclusion in the TS. Revisions to plant procedures containing
these requirements are required to be evaluated in accordance with
10 CFR 50.59. The proposed relocation is similar to recent
Amendments issued to the Davis-Besse Nuclear Power Station and the
San Onofre Nuclear Generating Station.
The revision proposed for TS 5.5.2.2 will replace the specific
title of the Chairperson of the Safety Audit and Review Committee
[[Page 17239]]
and replace it with ``Member as appointed by the Vice President.''
This will allow the flexibility to change chairmanship of the
committee amongst the members.
The revisions proposed to TS 5.3, 5.5, 5.6, 5.7, and 5.11 revise
position titles and reporting responsibilities to reflect
organizational changes. Qualifications for individuals in these
positions meet or exceed the previous requirements.
The revision to TS 5.10 concerning retention of records of
radioactive shipments will update the TS to current 10 CFR 20
requirements. Plant procedures already comply with current 10 CFR 20
record retention requirements. The addition of the Section 5.0 title
corrects a minor format discrepancy.
These proposed revisions are administrative in nature. The
proposed revisions have no effect on any initial assumptions or
operating restrictions assumed in any accident, nor do these changes
have any effect on equipment required to mitigate the consequences
of an accident. Therefore the proposed revisions do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed revisions correct minor errors, remove outdated
information, are consistent with changes in organizational
structure, 10 CFR Part 20, or the criteria contained in 10 CFR
50.36. These changes will not result in any physical alterations to
the plant configuration, changes to setpoint values, or changes to
the application of setpoints or limits. No new operating modes are
proposed as a result of these changes. Therefore the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The revisions listed above correct minor errors, remove outdated
information, or are consistent with changes in organizational
structure, 10 CFR Part 20, or the criteria contained in 10 CFR
50.36. These changes will not result in any physical alterations to
the plant configuration, changes to setpoint values, or changes to
the application of setpoint or limits. Therefore the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L
Street, N.W., Washington, DC 20005-3502.
NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: February 26, 1997.
Description of amendment request: The proposed amendment would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.4.5 and 3.4.6.2,
including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the
implementation of steam generator (SG) tube voltage based repair
criteria for outside diameter stress corrosion cracking (ODSCC)
indications at tube-to-tube support plate (TSP) intersections. The
allowed primary-to-secondary operational leakage from any one SG would
be reduced from 500 gpd to 150 gpd.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Structural Integrity Considerations
The structural criteria ensure that all indications subjected to
voltage-based repair limits will be able to withstand pressure
loading consistent with the criteria of NRC Regulatory Guide (RG)
1.121.
Tube burst criteria are inherently satisfied during normal
operating conditions because of the proximity of the tube support
plate (TSP). It is conservatively assumed that the entire crevice
region is uncovered during the secondary side blowdown of a main
steam line break (MSLB). Therefore, during a postulated MSLB
accident, tube burst capability must exceed the RG 1.121 criterion
requiring a margin of 1.43 times the steam line break pressure
differential on tube burst.
Based on the latest industry database, the RG 1.121 criterion is
satisfied by bobbin coil indications of outside diameter stress
corrosion cracking (ODSCC) with signal amplitudes less than 8.7
volts. The latest NRC-approved database will be used for repair and
analysis applications.
Industry testing of model boiler and operating plant tube
specimens for free-span tubing (no tube support plate (TSP)
restraint) at room temperature conditions show typical burst
pressures in excess of 5,000 psi for ODSCC indications with voltage
measurements at or below 8.7 volts. This tube burst capability
exceeds the RG 1.121 criterion.
The lower voltage repair limit is conservatively defined to be
2.0 volts in accordance with NRC Generic Letter (GL) 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator
Tubes Affected by Outside Diameter Stress Corrosion Cracking,''
August 3, 1995. This 2.0 volt repair limit is very conservative
because it contains a large safety margin, based on a structural
limit of 8.7 volts. A maximum allowable upper repair limit (URL) is
also established using the guidance of GL 95-05. The URL is
calculated before each inspection by subtracting the NDE uncertainty
and growth rate allowances from the current structural limit. The
URL for near term inspections at DCPP Units 1 and 2 is expected to
be about 5.0 volts. Bobbin indications greater than 2.0 volts and
less than or equal to 5.0 volts that are confirmed by RPC will be
repaired. Bobbin indications greater than 5.0 volts will be
repaired.
Following each inspection, burst probability analyses are
performed for the end of cycle (EOC) distribution. In accordance
with GL 95-05, the projected MSLB burst probability must be less
than the threshold value of 1 x 10 x 2. Based on the relatively
small number and voltages of ODSCC indications identified to date at
DCPP Units 1 and 2, it is expected that the near term EOC
conditional burst probability for a faulted SG will be much less
than this threshold value, providing further assurance of acceptable
structural integrity.
Leakage Considerations
PG&E will implement reduced operational leakage limits as
recommended in GL 95-05. PG&E will revise the TS to implement a
maximum leakage rate of 150 gpd for any one SG to help preclude the
potential for excessive leakage during power operation in Modes 1
and 2. The TS has also been changed to specify that the 150 gpd leak
limit is not necessarily a limiting condition for operation in Modes
3 and 4. The 150 gpd leak rate per steam generator has been
established for normal operation. This leakage rate provides added
assurance against tube rupture at normal and faulted conditions. In
Modes 3 and 4, there is less differential pressure across the tube
and the potential source term from a tube failure is much less than
in Modes 1 and 2. The operational leak rate monitoring program is a
defense-in-depth measure that provides a means for identifying leaks
during power operation to allow for repair before such leaks can
result in tube failure. The leakage criteria ensure that for
indications subjected to voltage-based repair criteria, induced
leakage under worst-case MSLB conditions will not result in offsite
and control room dose releases that exceed the applicable guideline
values of 10 CFR 100 and GDC 19.
Relative to the expected leakage during accident condition
loadings, a postulated MSLB outside of containment, but upstream of
the main steam isolation valve (MSIV), represents the most limiting
radiological condition for implementation of voltage-based repair
criteria. The steam generator tubes are subjected to an increase in
differential pressure following a MSLB, resulting in a postulated
increase in leakage
[[Page 17240]]
and associated offsite doses. Leakage following a MSLB bypasses
containment.
PG&E will calculate the primary-to-secondary leakage for
degradation subjected to the voltage repair criteria under worst-
case postulated MSLB conditions. The leak rate will be compared to
the maximum allowable leak rate limit of 12.8 gpm to ensure that a
postulated MSLB occurring at EOC would not result in radiological
consequences that are in excess of the applicable offsite and
control room dose guidelines of 10 CFR 100 and GDC 19. Based on the
relatively small number of ODSCC indications identified to date at
DCPP Units 1 and 2, it is expected that the near term EOC predicted
leak rates for a faulted SG will be much less than the maximum
allowable leak rate limit.
Therefore, based on the structural integrity and leakage
considerations discussed above, the proposed changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Implementation of the proposed voltage-based repair criteria for
ODSCC at TSP intersections does not introduce any significant change
to the plant design basis. Use of the criteria does not create a
mechanism which could result in an accident in the free span because
the repair criteria do not apply to tubes containing ODSCC located
outside the thickness of the TSPs. Based on the burst probability
acceptance limit of 1 x 10-2, it is expected that for all plant
conditions, neither a single nor multiple tube rupture event would
likely occur in a steam generator where voltage-based repair
criteria have been applied.
Steam generator tube integrity is continually maintained through
inservice inspection and primary-to-secondary leakage monitoring.
Any tubes with ODSCC degradation in excess of the URL are repaired.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of the bobbin probe to disposition ODSCC degraded tubes
within TSP intersections by voltage-based repair criteria is
demonstrated to maintain SG tube integrity in accordance with the
requirements of RG 1.121. RG 1.121 describes a method acceptable to
the NRC Staff for meeting GDCs 14, 15, 31, and 32 by reducing the
probability or the consequences of SG tube rupture. This is
accomplished by determining the limiting conditions of degradation
of SG tubing, as established by inservice inspection, for which
tubes with unacceptable degradation are removed from service. Upon
implementation of the voltage-based repair criteria, even under the
worst case conditions, the occurrence of ODSCC at TSP intersections
is not expected to lead to a SG tube rupture during normal or
faulted plant conditions, nor is it expected to lead to unacceptable
primary-to-secondary leakage.
In addressing the combined effects of a loss of coolant accident
(LOCA) and safe shutdown earthquake (SSE) on the SGs, as required by
GDC 2, it has been determined that tube collapse may occur based on
analysis for a large break LOCA plus SSE. The analysis identifies a
maximum of 7.5 percent of tubes per SG located adjacent to wedge
regions that are subject to potential collapse during combined LOCA
and SSE. Tubes located in the wedge region exclusion zone will be
excluded from application of voltage-based repair criteria. Thus,
existing tube integrity requirements apply to these tubes and the
margin of safety is not reduced.
Implementation practices using voltage-based repair criteria
bounds RG 1.83 considerations. Specifically, GL 95-05 requires the
following: (1) enhanced eddy current inspection guidelines are
implemented to provide consistency in voltage normalization; (2) 100
percent bobbin coil inspections are performed each cycle for all hot
leg TSP intersections and all cold leg TSP intersections down to the
lowest cold leg TSP with known ODSCC indications; and (3) rotating
pancake coil (RPC) inspection of indications greater than 2 volts
are performed to characterize the principal degradation as ODSCC.
DCPP's proposed voltage-based repair criteria implementation
practices meet the above requirements, and in some areas exceed them
(for example, 100 percent bobbin coil inspections are routinely
performed each cycle on every TSP intersection).
Implementation of voltage-based repair criteria at TSP
intersections will decrease the number of tubes which must be
repaired. Since the installation of tube plugs to remove ODSCC
degraded tubes from service reduces RCS flow margin, voltage-based
repair criteria implementation will help preserve the margin of RCS
flow.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: February 27, 1997.
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Power Plant, Unit Nos. 1 and 2 by revising Technical Specifications
(TS) 3/4.8.1.1, ``A.C. Sources--Operating,'' to clarify that emergency
diesel generator (EDG) testing is initiated from standby conditions
rather than ``ambient'' conditions. The associated TS Bases will be
revised to discuss the temperature range that satisfies EDG standby
conditions. This amendment also proposes to revise TS 3/4.3.2,
``Instrumentation--Engineering Safety Features Actuation System
Instrumentation.'' This revision clarifies that when one or both of the
first level load shed relays, or one or both of the second level
undervoltage relays are inoperable, the associated EDG for that bus
shall be declared inoperable.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the technical specifications (TS) do not
change the function or operation of any plant equipment or affect
the response of that equipment if it is called upon to operate.
The proposed change to TS 4.8.1.1.2a.2 and the Bases will
clarify the term ``ambient conditions'' as used in the emergency
diesel generator (EDG) surveillance requirement. EDG testing will
still be completed on a frequency commensurate with the current TS.
The proposed change to TS 3.3.2, Table 3.3-3, will permit time
to restore the load shed first level undervoltage relays (FLURs) and
second level undervoltage relays (SLURs) to operable status that is
consistent with times allowed for outage of other safety-related
equipment affecting one train of vital equipment. This proposed
change maintains a high degree of equipment availability without
requiring unnecessary initiation of a plant shutdown for partial
equipment outages.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of
[[Page 17241]]
accident from any accident previously evaluated.
The proposed change to TS 4.8.1.1.2a.2 and the Bases will
clarify the term ``ambient conditions'' as used in the EDG
surveillance requirement. EDG testing will still be completed on a
frequency commensurate with the current TS, and will be more
representative of the conditions under which the EDGs would be
required to start in an accident condition.
The proposed change to TS 3.3.2, Table 3.3-3, will provide time
to restore the load shed FLURs and SLURs to operable status that is
consistent with times allowed for outage of other safety-related
equipment affecting one train of vital equipment. The load shed FLUR
and SLUR sets for one 4 kV bus only affect one train of vital
equipment. If an accident occurred while the relays were inoperable,
the redundant trains (two remaining EDGs and vital buses) would
complete the safety function. The proposed allowed outage time (AOT)
for the load shed FLURs and SLURs is bounded by the time allowed for
an EDG supporting the vital 4 kV bus and is consistent with AOTs for
other safety-related components.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change to TS 4.8.1.1.2a.2 and its Bases, clarifies
the term ``ambient conditions'' as used in the EDG surveillance
requirement. EDG testing will still be completed on a frequency
commensurate with the current TS. Use of temperatures in the standby
range result in no significant variation in EDG start times as
indicated by the diesel vendor and by PG&E test results. Standby
conditions are representative of actual starting conditions that
would be in effect if the EDGs started in an accident.
The proposed change to TS 3.3.2, Table 3.3-3, will provide time
to restore the load shed FLURs and SLURs to operable status that is
consistent with times allowed for outage of other safety-related
equipment affecting one train of vital equipment. If an accident
occurred while the relays were inoperable, the redundant trains (two
remaining EDGs and vital buses) would complete the safety function.
The proposed change eliminates an unneccessary plant shutdown and
associated risk due to shutdown transient. It prevents a transient
that could require the EDGs at a time when less than all three EDGs
would be operable.
Therefore, neither of the proposed changes involves a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407.
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Portland General Electric Company, et al., Docket No. 50-344,
Trojan Nuclear Plant, Columbia County, Oregon
Date of amendment request: January 28, 1997.
Description of amendment request: The proposed amendment by
Portland General Electric (PGE or the licensee) clarifies the
administrative controls that are used for the revision and maintenance
of the Certified Fuel Handler Training Program. The change allows the
licensee to make changes to the certified fuel handlers program without
prior NRC staff approval. The text of the proposed change is taken from
the improved standard technical specifications, NUREG-1431, ``Standard
Technical Specifications, Westinghouse Plants.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
In accordance with the requirements of 10 CFR 50.92, ``Issuance
of amendment,'' this license amendment request is judged to involve
no significant hazards consideration based upon the following:
1. The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change is a clarification of the method of control
that will be used for the Certified Fuel Handler Training Program,
and as such, is administrative in nature and has no impact on the
probability or consequences of accidents previously evaluated. The
physical structures, systems, and components of the facility and the
operating procedures for their use are unaffected by this proposed
clarification. The proposed administrative controls provide adequate
confidence that personnel that perform the certified fuel handler
functions will have been adequately trained for the changing
conditions of the facility. Since the training program will prepare
the operations personnel for fuel handling operations, including
responses to abnormal events/accidents, there will be no increase in
the probability of occurrence or in the consequences of an accident
previously evaluated.
2. The proposed license amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
This change ensures the qualifications of the operations
personnel are commensurate with the tasks to be performed and the
conditions to which they may be required to respond. This change
does not affect plant equipment or the procedures for operating
plant equipment and, therefore, does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. The proposed license amendment does not involve a significant
reduction in a margin of safety.
This change ensures the qualification of the operations
personnel are commensurate with the tasks to be performed and the
conditions to which they may be required to respond. The assumptions
for a fuel handling accident in the Fuel Building are not affected
by the proposed change. The proposed amendment does not, therefore,
involve a reduction in a margin of safety.
The NRC staff has reviewed the analysis of the licensee and, based
on this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Branford Price Millar Library,
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151,
Portland, Oregon 97207.
Attorney for the Licensees: Leonard A. Girard, Esq., Portland
General Electric Company, 121 S.W. Salmon Street, Portland, Oregon
97204.
NRR Project Director: Seymour H. Weiss.
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: September 11, 1996.
Description of amendment request: The proposed amendment would
permit operation with increased safety relief valve (SRV) and safety
valve (SV) setpoint tolerance and permit operation up to 100% of rated
power with a single inoperable SRV.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed changes will permit operation with increased SRV
and SV setpoint tolerance and permit operation up to 100% of rated
power with a single inoperable SRV.
[[Page 17242]]
The valves are not related to the control rod system. The valves
are not involved in the initiation of a Control Rod Drop Accident.
The valves are part of the Reactor Vessel (RV) pressure boundary and
their failure could initiate a LOCA [loss-of-coolant accident].
However, the proposed changes do not constitute a change in the
design of the valves from a pressure boundary perspective. The
proposed changes do not affect the probability of a LOCA initiated
by valve failure. The valves are not a component, system, or
structure involved in refueling operations. The valves and their as-
found setpoint tolerance are not involved in the initiation of a
Refueling Accident.
The design basis Main Steam Line Break is a complete severance
of one main steam line outside the secondary containment. The SRVs
and SVs are located inside primary containment and cannot cause a
main steam line rupture outside secondary containment. The valves
are not involved in the initiation of a design basis Main Steam Line
Break. The probability or consequences of these accidents are not
affected.
Attachment C [see application dated September 11, 1996] includes
an analysis to demonstrate that margin exists to SV challenges
during an Abnormal Operational Transient (AOT). For this purpose a
Generator Load Rejection without Bypass (GLRWOBP) was identified as
the limiting AOT. The results confirm that SV challenges would not
occur with an inoperable SRV at rated power.
The current Technical Specification limit of 95% rated power or
less with an inoperable SRV is therefore not required to prevent SV
challenges during an AOT.
As discussed in Attachment C [see application], the impact of
the proposed as-found SRV setpoint tolerance increase on SRV piping/
supports and discharge loads to the Torus was evaluated. A
mechanical loads analysis confirmed the integrity of these
components, systems, and structures during SRV discharge with the
proposed changes.
Attachment C [see application] provides an evaluation of the
impact of the proposed changes on the consequences of the Loss of
Coolant Accident and the Main Steam Line Break. The limiting LOCA
event is a break in the recirculation loop, with a break area of 0.6
ft\2\, at the pump discharge location, with a loss of one train of
DC power as the single failure. For breaks in the recirculation line
larger than 0.4 ft\2\, the SRVs would not be challenged. Therefore,
in assessing the impact of the proposed changes on 10CFR50.46
acceptance criteria, only recirculation line breaks less than 0.4
ft\2\ were reevaluated. Results show that the 0.6 ft\2\
recirculation line break remains the limiting LOCA event and it is
not affected. The consequences of the limiting design basis LOCA are
not increased by the proposed changes. The design basis accident for
containment performance is a double-ended break in the recirculation
pump suction. For this size break, the SRVs are not challenged.
Therefore, the proposed changes do not have any effect on the design
basis accident for containment performance. The design basis
accident for radioactive material releases and radiological effects
is a complete severance of one main steam line outside the secondary
containment. For steam line breaks outside the containment, MSIVs
[main steam isolation valves] close and terminate radiological
releases outside the containment, SRVs are not challenged until
after MSIV closure and isolation. Therefore, the proposed changes do
not increase the radiological consequences of the design basis Main
Steam Line Break.
The SRVs and SVs are designed to mitigate the consequences of
malfunctions of equipment which result in a Nuclear System pressure
increase. These abnormal operational transients are defined and
analyzed in Section 14.5.1 of the VY [Vermont Yankee] FSAR [final
safety analysis report]. The impact of the proposed changes on these
abnormal operational transients was evaluated. Results are
documented in Attachment C [see application] and show that
applicable acceptance criteria are met provided operating MCPR
[minimum critical power ratio] limits as specified in the COLR [core
operating limit report] are adjusted to reflect the effects of the
proposed changes. A hot channel analysis of the limiting delta CPR
overpressure transient confirmed that a 0.02 increase in the
operating MCPR limits bounds the combined effects of implementing
the proposed changes in the current cycle. The operating MCPR limits
in COLR have already been increased for the current cycle.
Appropriate operating MCPR limits for future cycles will be
determined from cycle-specific safety analyses performed with the
approved changes.
Current practice regarding SRV setpoints is to assure plus or
minus 1% tolerance is met as required by the ASME [American Society
of Mechanical Engineers] Boiler & Pressure Vessel Code referenced in
Technical Specification Surveillance Requirement 4.6.E.2. As-left
setpoints always meet the plus or minus 1% tolerance. The safety
analysis in Attachment C [see application] demonstrates that as-
found setpoints within plus or minus 3% are acceptable. However,
valves re-installed after testing will continue, as previously, to
meet plus or minus 1% tolerance as required by the ASME Boiler &
Pressure Vessel Code. Thus, the probability of SRV actuation (and
the associated risk of failure to reseat properly) is not increased
by the proposed change.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from an accident previously
evaluated.
The proposed changes will permit operation with increased Safety
Relief Valve (SRV) and Safety Valve (SV) setpoint tolerance and
permit operation up to 100% of rated power with a single inoperable
SRV. The proposed changes:
(1) do not constitute a change in the design of the valves;
(2) will not cause the valve or associated systems and
structures to be operated beyond their original design envelopes;
and,
(3) do not involve new plant equipment.
Therefore, this amendment does not create the possibility of a
new or different kind of accident.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
Technical Specification Basis 3.6 and 4.6D identifies the
minimum critical power ratio (MCPR) safety limit. Operational
restraints on MCPR are placed in the COLR to assure no violation of
the MCPR safety limit during AOTs. The impact of the proposed
changes on MCPR limits was determined by performing a hot channel
analysis for the overpressure transient which yields the largest
transient drop in CPR [critical power ratio] (delta CPR). Results
are documented in Attachment C [see application], and show that a
0.02 increase in the operating MCPR limits bounds the combined
effects of the proposed changes and assures the MCPR safety limit is
not violated during AOTs. The margin of safety defined by the MCPR
safety limit is not reduced.
Technical Specification Basis 3.6 and 4.6D also identifies the
ASME Boiler and Pressure Vessel Code Section III-A limit which
permits pressure transients up to 10% over design pressure (110% x
1250 = 1375 psig). This margin of safety is not reduced by the
proposed changes. Attachment C [see application] documents new
overpressure transient analysis with results that demonstrate the
ASME overpressure limit of 110% of design is met. This license
amendment request does not propose to reduce the margin of safety
defined by the ASME Boiler & Pressure Vessel Code limit.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, VT 05301.
Attorney for licensee: R. K. Gad, III, Ropes and Gray, One
International Place, Boston, MA 02110-2624.
NRC Project Director: Patrick D. Milano, Acting.
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: February 3, 1997 as supplemented March
18, 1997.
Description of amendment request: The proposed change to Technical
Specification 4.15.B.1 is administrative in nature in that it revises
the Technical Specifications (TS) to be consistent with the NRC-
approved inservice inspection program. In addition, three TS pages
which were previously approved by NRC, and which were inadvertently
omitted in an earlier amendment (amendments 40 and 39 for units 1 and
2, respectively), are being reissued.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 17243]]
issue of no significant hazards consideration, which is presented
below:
1. Operation of Surry Units 1 and 2 in accordance with the
proposed Technical Specifications change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change is administrative in nature, and station
operations are not being affected. The ASME Section XI Code
requirements are thoroughly established, reviewed and approved by
ASME, the industry and ultimately endorsed by the NRC for inclusion
into 10 CFR 50.55a. Updates to the Code reflect advances in
technology and consider information obtained from plant operating
experience to provide enhanced inspection and examination techniques
for pipe welds. Therefore, performing weld examinations for the pipe
in our augmented inspection program to the requirements of the 1989
edition of the ASME Section XI Code provides a regulatory acceptable
and adequate level of assurance that the integrity of the pipe will
be maintained. By not referencing a specific Code edition in the
Technical Specifications, our examinations for pipe in the augmented
inspection program will consistently be performed to the Code of
record, consistent with the requirements [of] 10 CFR 50.55a.
Consequently, the probability or consequences of an accident
previously evaluated are not increased.
2. The proposed Technical Specifications change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
As noted above, the proposed change is administrative in nature,
and no new accident precursors are being introduced. Since the
augmented inspection program will continue to be performed to NRC
approved ASME Section XI Code requirements, adequate assurance is
provided to ensure the integrity of the pipe. Consequently, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed Technical Specifications change does not involve
a significant reduction in a margin of safety.
Performing weld examinations to the Code of record is prudent,
consistent with accepted industry and regulatory requirements, and
provides adequate assurance that piping integrity will be
maintained. The use of a general ASME Section XI Code reference in
Technical Specification 4.15.B.1 is consistent with the existing
wording in Technical Specifications 4.15.A and C, and ensures that
weld examinations are being consistently performed to the currently
approved edition of the ASME Section XI Code. This is an
administrative change and as such does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. The staff notes that the reissuance of three TS pages is a
purely administrative matter which involves no significant hazards
consideration and which has been considered previously. Therefore, the
NRC staff proposes to determine that the amendment request involves no
significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Mark Reinhart, Acting.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two
Creeks, Manitowoc County, Wisconsin
Dates of amendment requests: June 4, 1996, as supplemented August
5, September 26, October 21, November 13, November 20, and December 2,
1996, and January 16, March 5, and March 20, 1997 (TSCR 188 and 189).
Description of amendment requests: The proposed amendments would
revise License Nos. DPR-24 and DPR-27 to add commitments for control
room habitability and revise Technical Specification (TS) Sections
15.1, ``Definitions,'' 15.2.1, ``Safety Limit, Reactor Core,'' 15.2.3,
``Limiting Safety System Settings and Protective Instrumentation,''
Section 15.3.1, ``Reactor Coolant System,'' 15.3.4, ``Steam and Power
Conversion System,'' 15.3.5, ``Instrumentation System, 15.4.1,
``Operational Safety Review,'' 15.5.3, ``Design Features--Reactor,''
and 15.6.9, ``Plant Reporting Requirements,'' and modify the bases for
Section 15.2.2, ``Safety Limit, Reactor Coolant System Pressure,'' and
Section 15.3.1.C, ``Maximum Coolant Activity,'' to incorporate changes
associated with the operation of Point Beach Nuclear Plant (PBNP), Unit
2, with replacement steam generators. The new analyses performed for
replacing Unit 2 steam generators resulted in changes to the reactor
core safety limits and protective instrumentation setpoints for Unit 1
as well as Unit 2. Calculations are based on operation at either 2000
psia or 2250 psia and an average temperature limit of greater than or
equal to 557 degrees Fahrenheit and less than or equal to 573.9 degrees
Fahrenheit. New dose calculations were performed based on new setpoints
for low-low steam generator water level, new values of primary and
secondary steam generator volumes, and revised accident analyses for
steam generator tube rupture, main steam line break, locked rotor, and
control rod ejection. Additional license conditions are proposed to
document the commitments made to improve habitability of the control
room so that dose limits do not exceed 10 CFR Part 50, Appendix A,
General Design Criterion 19, without relying on the use of potassium
iodide pills and/or self-contained breathing apparatus. The original
applications were previously noticed in the Federal Register on July 3,
1996 (61 FR 34903 and 34904).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed TS changes reflect the replacement of steam
generators at PBNP, including new analyses and setpoints, and a
different standard and acceptance criteria for Dose Equivalent I-
131. The proposed setpoints maintain the margin to safe operation of
Unit 2 with the replacement steam generators. In order to maintain
one set of safety analyses for both units, the analyses for
operation of Unit 2 with the replacement steam generators were
performed to encompass the operation of Unit 1. Therefore, the
proposed changes apply to the operation of both units and maintain
the margin of safety for each. The staff independently performed an
evaluation of the dose consequences for steam generator tube
rupture, main steam line break, locked rotor accident, and a rod
ejection accident. The staff determined there are no significant
increases in dose for the low population zone or the exclusion area
boundary. The licensee had not previously analyzed these accidents
for control room habitability. As a result of the proposed changes,
limiting control room doses will require compensatory measures, use
of potassium iodide and self-contained breathing apparatus, which
have been previously approved, until such time that the control room
ventilation design is improved. The commitments to improve control
design/operation are included as license conditions. Therefore, the
proposed changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
(2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
Installation of new steam generators, with a small increase in
primary side volume and new setpoints for instrumentation, does not
create the possibility of a new or different kind of accident from
any accident previously evaluated. The proposed setpoints
[[Page 17244]]
maintain the margin to safe operation of Unit 2 with the replacement
steam generators. In order to maintain one set of safety analyses
for both units, the analyses for operation of Unit 2 with the
replacement steam generators were performed to encompass the
operation of Unit 1. Therefore, the proposed changes apply to the
operation of both units and maintain the margin of safety for each.
These changes do not affect any of the parameters or conditions that
contribute to initiation of any accidents. Therefore, the proposed
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
(3) The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed setpoints maintain the margin to safe operation of
Unit 2 with the replacement steam generators. In order to maintain
one set of safety analyses for both units, the analyses for
operation of Unit 2 with replacement steam generators were performed
to encompass the operation of Unit 1. Therefore, the proposed
changes apply to the operation of both units and maintain the margin
of safety for each. Compensatory measures will ensure control room
doses remain within the dose guidelines in 10 CFR Part 50, Appendix
A, General Design Criterion 19, until such time as the control
ventilation system design/operation is revised. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
Based on this review, it appears that the three standards of 10
CFR 50.92(c) are satisfied. Therefore, the NRC staff proposes to
determine that the amendment requests involve no significant hazards
consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Power Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of amendment request: September 30, 1996, as supplemented
November 26, and December 12, 1996, February 13, and March 5, 1997
(TSCR 192).
Description of amendment request: The proposed amendments would
revise License Nos. DPR-24 and DPR-27 to add commitments for control
room habitability and revise Technical Specification (TS) Sections
15.3.3, ``Emergency Core Cooling System, Auxiliary Cooling Systems, Air
Recirculation Fan Coolers, and Containment Spray,'' TS 15.3.7,
``Auxiliary Electrical Systems,'' 15.5.2, ``Design Features-
Containment,'' and associated TS Bases to reflect proposed changes to
the limiting conditions for operation, action statements, allowable
outage times, and design specifications for the Point Beach Nuclear
Plant (PBNP) TS associated with the containment accident fan coolers,
service water equipment (pumps and piping), component cooling water
pumps, and normal and emergency power supplies. Specifically, these
proposed changes increase the number of service water pumps and
component cooling water pumps required to be operable, change the
description of the service water system to define three separate loops,
modify the limiting conditions for operation of the containment cooling
and iodine removal systems and the component cooling water and service
water systems, modify the auxiliary electrical system requirements,
modify the associated TS Bases, and change the design value for each
containment ventilation/air coolers from 55,600 Btu/sec to 41,700 Btu/
sec. The original application was previously noticed in the Federal
Register on November 19, 1996 (61 FR 58905).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
(1) The proposed changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes involve components currently installed in
the facilities and reflect current capabilities of this equipment.
Increasing the number of service water and component cooling water
pumps required to be operable, changing the service water header
definitions and modifying the limiting conditions for operation for
service water and component cooling water, and modifying the
requirements for the 4160/480-volt safeguards buses does not
increase the probabilities of any accidents currently evaluated in
the final safety analysis report (FSAR). The probabilities of
accidents previously evaluated in the FSAR are based on the
probability of initiating events for these accidents. Initiating
events for accidents previously evaluated for Point Beach include:
Control rod withdrawal and drop, CVCS [chemical volume and control
system] malfunction (boron dilution), startup of an inactive reactor
coolant loop, reduction in feedwater enthalpy, excessive load
increase, losses of reactor coolant flow, loss of external
electrical load, loss of normal feedwater, loss of all AC power to
the auxiliaries, turbine overspeed, fuel handling accidents,
accidental releases of waste liquid or gas, steam generator tube
rupture, steam pipe rupture, control rod ejection, and primary
coolant system ruptures. The change to the heat removal capability
of the containment ventilation/air coolers from 55,600 Btu/sec to
41,700 Btu/sec was evaluated to ensure that containment design is
not challenged. Therefore, the proposed changes do not affect the
probability of occurrence or the consequences of any accident
previously evaluated in the FSAR. During review of the proposed
changes, the staff determined that other changes made to the
operation of the containment spray system and the control room
ventilation design and operation could affect the doses associated
with a loss-of-coolant accident. The staff has determined that there
is no significant increase in offsite doses. As a result of the
proposed changes and current plant design, limiting control room
doses will require compensatory measures, use of potassium iodide
and self-contained breathing apparatus, which have been previously
approved, until such time that the control room ventilation design/
operation is improved. The commitments to improve control design/
operation are included as license conditions.
(2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not introduce any new accidents from any
previously evaluated. Failures for the systems affected by the
proposed changes, service water system, component cooling water
system, containment ventilation/air cooling units, and the 4160/480-
volt safeguards buses are factored into the accident analyses
included in the FSAR. No new or different kinds of accidents are
created since no new or different accident initiators or sequences
are involved. Therefore, these proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated in the Point Beach FSAR.
(3) The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes provide the appropriate limiting conditions
for operation, action statements, allowable outage times, and design
specifications for service water, component cooling water,
containment cooling, and normal and emergency power supplies. This
ensures that the safety systems that protect the reactor and
containment will operate as required. The impact of changes to
design and operation of affected systems do not affect the reactor
and containment design. Therefore, the margins of safety for Point
Beach are not being reduced because the design and operation of the
reactor and containment are not being changed and the safety systems
and limiting conditions of operation for these safety systems that
provide their protection that are being changed will continue to
meet the requirements for accident mitigation for PBNP. Compensatory
measures will ensure control room doses remain within the dose
guidelines in 10 CFR Part 50, Appendix A, General Design Criterion
19, until such time as the control ventilation system design/
[[Page 17245]]
operation is revised. Therefore, the proposed changes will not
create a significant reduction in a margin of safety.
Based on this review, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the amendment requests involve no significant hazards
consideration.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts,
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 7, 1997.
Description of amendment request: The proposed amendments would
revise Technical Specification 3/4.7.1.6 and Section 15.6.3 of the
Updated Final Safety Analysis Report to require four instead of three
steam generator pressure operated relief valves operable.
Date of publication of individual notice in Federal Register: March
13, 1997 (62 FR 11931).
Expiration date of individual notice: April 14, 1997.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Unit 2, Luzerne County,
Pennsylvania
Date of amendment request: March 17, 1997.
Brief description of amendment request: The proposed amendment
would modify the Design Features Section 5.3.1 of the Technical
Specifications to reflect the Atrium-10 design and would include a
Siemens Power Corporation topical report reference in Section 6.9.3.2
to reflect mechanical design criteria for this fuel. This change would
allow this fuel to be loaded and maintained in the core only under
Condition 5, (refueling).
Date of publication of individual notice in Federal Register: March
25, 1996 (62 FR 14167).
Expiration date of individual notice: April 24, 1997.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of application for amendments: May 2, 1995, as supplemented by
letter dated March 7, 1996.
Brief description of amendments: These amendments modify the
licenses to authorize incorporation in the Updated Final Safety
Analysis Report (UFSAR) of certain changes to the description of the
facilities involving a revised large-break loss of coolant accident
(LOCA) analysis that addresses a previously unanalyzed release path
through the steam generators to the atmosphere.
Date of issuance: March 17, 1997.
Effective date: March 17, 1997, to be implemented within 60 days of
issuance.
Amendment Nos.: Unit 1--111; Unit 2--103; Unit 3--83.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Operating Licenses and Updated Final Safety
Analysis Report.
Date of initial notice in Federal Register: December 6, 1995 (60 FR
62487). The March 7, 1996, supplemental letter provided additional
clarifying information and did not change the initial no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
March 17, 1997. No significant hazards consideration comments received:
No.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of application for amendment: January 29, 1997, as
supplemented February 6, and February 21, 1997.
Brief description of amendment: The amendment adds a new Technical
[[Page 17246]]
Specification 3.0.5 to provide guidance for returning equipment to
service under administrative controls for the sole purpose of
performing testing to demonstrate operability.
Date of issuance: March 17, 1997.
Effective date: March 17, 1997.
Amendment No.: 69.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997
(62FR6569).
The February 6, and February 21, 1997 letters provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 17, 1997.
No significant hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: January 10, 1997, as
supplemented January 31, February 20, and March 3, 1997.
Brief description of amendment: The amendment revises Technical
Specification 4.8.1.1.2 to clarify pressure testing requirements for
the isolable and non-isolable portions of the diesel fuel oil piping.
Date of issuance: March 19, 1997.
Effective date: March 19, 1997.
Amendment No.: 70.
Facility Operating License No. NPF-63: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 5, 1997 (62 FR
5490). The January 31, February 20, and March 3, 1997, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 19, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina.
Date of application for amendments: November 4, 1996 and
supplemented February 5, 1997.
Brief description of amendments: The amendments revise Technical
Specification Section 4.7.13.1.c to eliminate the requirement that the
18-month Standby Shutdown System diesel generator inspection be
performed only during shutdown of both reactors.
Date of issuance: March 13, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--157--Unit 2--149.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64383) The supplemental letter dated February 5, 1997, provided
additional information that did not change the scope of the November 4,
1996, application and the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 13, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: January 13, 1997.
Brief description of amendments: The amendments revise the
Technical Specifications so that the containment integrated leak rate
Type A testing will now be performed consistent with the revised 10 CFR
Part 50, Appendix J, Option B, by referring to Regulatory Guide 1.163,
``Performance-Based Containment Leak-Test Program.'' No changes to
implement Option B for the Type B and Type C tests were requested by
the licensee at this time.
Date of issuance: March 21, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1--173--Unit 2--155.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6575) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, North Carolina 28223-0001.
Duke Power Company, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina
Date of application of amendments: February 15, 1996, as
supplemented by letter dated February 18, 1997.
Brief description of amendments: The amendments add operability and
surveillance requirements regarding operation and testing of the Keowee
Hydro Station to the Oconee Technical Specifications.
Date of Issuance: March 20, 1997.
Effective date: As of the date of issuance to be implemented within
30 days. Implementation shall include revision of the Selected Licensee
Commitment manual to incorporate the Keowee Hydro units' commercial
power operating restrictions curves in accordance with the application
for the amendments.
Amendment Nos.: Unit 1--222; Unit 2--222; Unit 3--219.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: March 27, 1996 (61 FR
13523) The February 18, 1997, letter provided clarifying information
that did not change the scope of the February 15, 1996, application and
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 20, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691.
[[Page 17247]]
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: September 18, 1992, as
supplemented October 6, 8, 15, 23, and November 13 and 20, 1992, March
5, May 24, June 10, and December 20, 1993, April 6 and July 28, 1995,
and September 11, October 1, December 13, 19 and 23, 1996.
Brief description of amendments: The amendments modify the Facility
Operating Licenses, Technical Specifications, Environmental Protection
Plan, and Antitrust conditions to add Southern Nuclear Operating
Company, Inc., as operator of the facilities, with exclusive
responsibility and control over its physical construction, operation,
and maintenance. The antitrust license conditions divorce Southern
Nuclear from marketing or brokering power or energy from the Hatch
Plant and holds Georgia Power Company accountable for the actions of
its agent, Southern Nuclear, to the extent Southern Nuclear's actions
contravene the Hatch antitrust license conditions. An Order Approving
Southern Nuclear Operating Company, Incorporated, As Exclusive Operator
was included along with the issuance of the amendments.
Date of issuance: March 17, 1997.
Effective date: To be implemented within 60 days of the date of
issuance.
Amendment Nos.: 203 and 144.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications and Operating Licenses.
Date of initial notice in Federal Register: October 14, 1992 (57 FR
47131). The October 6, 8, 15, 23, and November 13 and 20, 1992, March
5, May 24, June 10, and December 20, 1993, April 6 and July 28, 1995,
and September 11, October 1, December 13, 19 and 23, 1996, letters, did
not change the scope of the September 18, 1992, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated March 17, 1997, and an Environmental Assessment
dated October 27, 1992.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: October 7, 1996.
Brief description of amendments: The amendments revise Surveillance
Requirements (SRs) 3.1.7.7 and 3.4.3.1, and Limiting Conditions for
Operation 3.4.3, 3.5.1, and 3.6.1.6 to increase the nominal mechanical
pressure relief setpoints for all of the 11 safety/relief valves (SRVs)
to 1150 psig and allow operation with one SRV and its associated
functions inoperable. The change will reduce the potential for SRV
pilot leakage and the potential for forced outages due to an inoperable
SRV during a fuel cycle.
Date of issuance: March 21, 1997.
Effective date: As of the date of issuance to be implemented for
Unit 1 prior to startup from its refueling outage scheduled for fall
1997; and for Unit 2 prior to startup from its refueling outage
currently scheduled for March 15, 1997.
Amendment Nos.: 204 and 145.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
129). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of application for amendments: October 29, 1996, as
supplemented February 19, 1997.
Brief description of amendments: The amendments revise the
Technical Specifications associated with the installation of a digital
Power Range Neutron Monitoring system.
Date of issuance: March 21, 1997.
Effective date: As of the date of issuance to be implemented for
Unit 1 prior to its startup from the fall of 1997 refueling outage; and
implemented for Unit 2 prior to its startup from the spring of 1997
refueling outage.
Amendment Nos.: 205 and 146.
Facility Operating License Nos. DPR-57 and NPF-5: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
130). The February 19, 1997, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear
Date of application for amendment: November 27, 1996 (TSCR 232).
Brief description of amendment: The amendment changes the
acceptance criteria for the individual cell voltage from 2.0v to 2.09v,
the frequency for battery specific gravities to implement the
recommendations of IEEE 450-1995, deletes surveillance 4.7.B.4.d, and
adds a clarifing phrase ``while on a float charge . . .'' where
appropriate.
Date of Issuance: March 24, 1997.
Effective date: March 24, 1997.
Amendment No.: 189.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: February 12, 1997 (62
FR 6576) The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated March 24, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: November 20, 1996, as supplemented by
letters dated February 20, 1997, and March 25, 1997.
Brief description of amendment: The amendment revises Section 5.2
of the Fort Calhoun Station technical specifications to relocate
controls for working hours to the Updated Safety Analysis Report.
Date of issuance: March 27, 1997.
Effective date: March 27, 1997.
Amendment No.: 181.
[[Page 17248]]
Facility Operating License No. DPR-40: Amendment revised the
Technical Specifications and operating license.
Date of initial notice in Federal Register: January 2, 1997 (62 FR
131) The February 20, 1997, and March 25, 1997, supplemental letters
provided additional clarifying information that did not change the
portion of the initial no significant hazards consideration
determination that addressed this proposed change.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 27, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: January 11, 1996, as
supplemented by letters dated February 26, May 22, June 27, July 12,
December 23, 1996, and March 17, 1997
Brief description of amendment: The amendments revise Section 6.0
(Administrative Controls) of the Hope Creek TS to: (1) Relocate the
requirements of Section 6.5 (Station Operations Review Committee,
Nuclear Safety Review and Audit, and Technical Review and Control) to
the Quality Assurance Program, (2) replace specific management titles
with generic management functional positions, (3) change Operating
Engineer to Assistant Operations Manager, (4) require a Senior Reactor
Operator license be held by either the Operations Manager or one of the
Assistant Operations Managers, and (5) correct some typographical
errors in Section 6.0.
Date of issuance: March 21, 1997.
Effective date: As of date of issuance and shall be implemented
within 60 days.
Amendment No.: 97.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications and the license.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5817).
The supplemental letters provided clarifying information that did
not change the initial proposed no significant hazards consideration
determination nor the original notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of application for amendment: October 25, 1996, as
supplemented by letters dated December 4, 1996, and January 24, 1997.
Brief description of amendment: This amendment changes Hope Creek
Technical Specification (TS) 3/4.1.3.5, ``Control Rod Scram
Accumulator,'' in order to: 1) Permit a separate entry into a TS action
statement for each inoperable control rod; 2) provide more specific
applicability for required actions in Operational Condition 1 or 2 with
one inoperable control rod scram accumulator (reactor pressure of
900 psig would be specified); 3) provide more specific
actions (verify charging water pressure) for two or more inoperable
control rod scram accumulators when reactor pressure is 900
psig; 4) provide more specific actions when reactor pressure is < 900
psig and one or more control rod scram accumulators are inoperable
(verify insertion of control rods associated with inoperable
accumulators and verify that charging water header pressure is
940 psig); 5) provide specific actions in Operational
Condition 5 with one or more withdrawn control rods inoperable; and 6)
eliminate the requirements to perform an 18-month channel functional
test of the leak detectors and the 18-month channel calibration of the
pressure detectors.
Date of issuance: March 26, 1997.
Effective date: As of date of issuance, to be implemented within 60
days.
Amendment No.: 98.
Facility Operating License No. NPF-57: This amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64394) The December 4, 1996, and January 24, 1997, supplements did not
change the initial proposed no significant hazards consideration
determination. The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 26, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem
County, New Jersey
Date of application for amendments: January 11, 1996, as
supplemented by letters dated February 26, May 22, June 27, July 12,
December 23, 1996, and March 17, 1997.
Brief description of amendments: The amendments revise Section 6.0
(Administrative Controls) of the Salem TS to: 1) relocate the
requirements of Section 6.5 (Station Operations Review Committee,
Nuclear Safety Review and Audit, and Technical Review and Control) to
the Quality Assurance Program, 2) replace specific management titles
with generic management functional positions, 3) change Operating
Engineer to Assistant Operations Manager, 4) require a Senior Reactor
Operator license be held by either the Operations Manager or one of the
Assistant Operations Managers, and 5) correct some typographical errors
in Section 6.0.
Date of issuance: March 21, 1997.
Effective date: Both units, as of date of issuance, to be
implemented within 60 days.
Amendment Nos.: 192 and 175.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications and the license.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5818) The supplemental letters provided clarifying information that
did not change the initial proposed no significant hazards
consideration determination nor the original notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 21, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 Joseph
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama
Date of amendment request: December 26, 1997, as supplemented by
letter dated February 6, March 7, and March 21, 1997.
Brief Description of amendment: The amendment changes Technical
Specification 3/4.4.6, ``Steam Generators'' and associated Bases to
implement the voltage-based alternate repair criteria for steam
generator tubes in Farley Unit 1 in accordance with
[[Page 17249]]
Generic Letter 95-05, ``Voltage-Based Repair Criteria for Westinghouse
Steam Generator Tubes Affected by Outside Diameter Stress Corrosion
Cracking.''
Date of issuance: March 24, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 124.
Facility Operating License Nos. NPF-2 and NPF-8: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4353) By letter dated February 6, 1997, the licensee submitted
additional information to clarify the changes to the proposed repair
criteria, which did not change the scope of the December 26, 1996,
application and the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated March 24, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: January 10, 1997, as supplemented by
letter dated February 24, 1997.
Brief Description of amendments: The amendments revise the
Technical Specifications (TS) to incorporate the latest revised topical
reports governing the installation of laser welded steam generator tube
sleeves. In addition, the reference to a one-cycle implementation of
L*, which expired at the last Unit 2 outage was deleted from the Unit 2
TS.
Date of issuance: March 24, 1997.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 125 and 119.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: January 29, 1997 (62 FR
4355) The February 24, 1997, letter provided clarifying information
that did not change the original application and the initial proposed
no significant hazards consideration determination published in the
Federal Register on January 29, 1997 (62 FR 4355).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 24, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendments request: September 30, 1996.
Brief description of amendments: The amendments revise Technical
Specifications (TS) 3/4.1.1.1, 3/4.1.1.2, 3/4.1.1.3, 3/4.1.3.5,
3.1.3.6, 3.2.1, 3.2.2 and 3.2.3 and associated Bases to remove certain
cycle-specific parameter limits from the TS and relocate them to the
Core Operating Limits Report.
Date of issuance: March 25, 1997.
Effective date: As of the date of issuance to be implemented for
Unit 1 prior to entry into Mode 5 following the next scheduled
refueling outage, which should begin in March 1997; for Unit 2 prior to
entry into Mode 5 following the refueling outage scheduled to begin in
March 1998.
Amendment Nos.: 126 and 120.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications and License Conditions.
Date of initial notice in Federal Register: November 6, 1996 (61 FR
57491) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 25, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
Southern California Edison Company, et al., Docket No. 50-362, San
Onofre Nuclear Generating Station, Unit No. 3, San Diego County,
California
Date of application for amendment: February 18, 1997, as
supplemented by letter dated February 21, 1997.
Brief description of amendment: The amendment defers implementation
of Surveillance Requirement 3.3.5.6 of Technical Specifcation 3.3.5,
``Engineered Safety Features Actuation System (ESFAS)
Instrumentation,'' until the next SONGS Unit 3 shutdown, which will be
no later than the upcoming Cycle 9 refueling outage (currently
scheduled for April 12, 1997).
Date of issuance: March 17, 1997.
Effective date: March 17, 1997.
Amendment No.: 127
Facility Operating License No. NPF-15: The amendments revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (62 FR 9001 dated February 27, 1997). The notice
provided an opportunity to submit comments on the Commission's proposed
no significant hazards consideration determination. No comments have
been received. The notice also provided for an opportunity to request a
hearing by March 31, 1997, but indicated that if the Commission makes a
final no significant hazards consideration determination any such
hearing would take place after issuance of the amendment. The February
21, 1997, letter provided additional clarifying information and did not
change the original no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated March 17, 1997.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, P. O. Box 800, Rosemead, California 91770.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713.
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear
Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: February 14, 1997.
Brief description of amendment: This amendment revises Technical
Specification (TS) Section 3/4.5.2, ``Emergency Core Cooling Systems,
ECCS Subsystems--Tavg 280 deg.F.'' Surveillance
requirement 4.5.2.f would be modified to state that opening and closing
of the inspection port on the watertight enclosure for the decay heat
valve pit would not require this surveillance procedure to be
performed. This amendment also revises the applicable TS bases.
Date of issuance: March 24, 1997.
Effective date: Immediately, and shall be implemented no later than
120 days after issuance.
Amendment No.: 215.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (62 FR 8783 dated February 26, 1997). The
notice
[[Page 17250]]
provided an opportunity to submit comments on the Commission's proposed
NSHC determination. No comments have been received. The notice also
provided for an opportunity to request a hearing by March 30, 1997, but
indicated that if the Commission makes a final NSHC determination, any
such hearing would take place after issuance of the amendment.
The Commission's related evaluation of the amendment, finding of
exigent circumstances, and final determination of NSHC are contained in
a Safety Evaluation dated March 24, 1997.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606.
United States Department of Commerce, National Institute of Standards
and Technology, Docket No. 50-184, NIST Test Reactor
Date of application for amendment: January 17, 1997.
Brief description of amendment: This amendment revises the
Technical Specifications to change the name of the Reactor Radiation
Division to the NIST Center for Neutron Research and the Chief,
Radiation Division to Director, NIST Center for Neutron Research.
Date of issuance: March 31, 1997.
Effective date: March 31, 1997.
Amendment No.: 8.
Amended Facility License No. TR-5: This amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 26, 1997 (62
FR 8801). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 31, 1997.
No significant hazards consideration comments received: No.
Local Public Document Room Location: N/A.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) The
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 9, 1997, the licensee
may file a request for a hearing with respect to issuance of the
amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the
[[Page 17251]]
designated Atomic Safety and Licensing Board will issue a notice of a
hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-001, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-001, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: December 27, 1996, as
supplemented by letter dated March 18, 1997.
Brief description of amendments: The amendments modify the licenses
to authorize incorporation of certain changes to the description of the
facilities involving offsite power sources into the Updated Final
Safety Analysis Report (UFSAR) for the Palo Verde Nuclear Generating
Station (PVNGS).
Date of issuance: March 26, 1997.
Effective date: March 26, 1997, to be implemented within 60 days of
the date of issuance.
Amendment Nos.: Unit 1--112; Unit 2--104; Unit 3--84.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the operating licenses and the Updated Final Safety
Analysis Report.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendments, finding of
emergency circumstances, and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated March
26, 1997.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004.
NRC Project Director: William H. Bateman.
Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, Illinois
Date of application for amendments: March 26, 1997, as supplemented
on March 27, 1997.
Brief description of amendments: The proposed amendments provided
(1) An evaluation of the Unreviewed Safety Question (USQ) involving the
control room operator dose resulting from error in the secondary
containment volume, (2) a change in Technical Specification (TS)
4.7.P.2.b and 4.7.P.3 values for the allowed methyl iodide penetration
for the standby gas treatment charcoal adsorbers, and (3) change of TS
5.2.C to reflect the new calculated free volume of the secondary
containment.
Date of Issuance: March 27, 1997.
Effective date: March 27, 1997.
Amendment Nos.: 175, 171.
Facility Operating License Nos. DPR-29 and DPR-30: The amendments
revised the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: No.
The Commission's related evaluation of the amendments, finding of
emergency circumstances and final determination of no significant
hazards consideration are contained in a Safety Evaluation dated March
27, 1997.
[[Page 17252]]
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60603.
Local Public Document Room location: Dixon Public Library, 221
Hennepin Avenue, Dixon, Illinois 61021.
NRC Project Director: Robert A. Capra.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: January 29, 1997, as
supplemented February 11, 12, March 7, 10, 11, 19, and 20, 1997.
Brief description of amendments: The amendments authorize Northern
States Power Company to continue operation of Prairie Island Units 1
and 2 on an interim basis, through the incorporation of three license
conditions into its licenses, until a seismically qualified emergency
cooling water source is provided that will provide the basis to extend
the time for operator post-seismic cooling water load management. This
could be done either through a seismic evaluation of the intake canal,
physical modifications to the intake canal or plant, or some
combination of the two.
Date of issuance: March 25, 1997.
Effective date: March 25, 1997, with implementation of License
Condition 1 prior to Unit 2 entering Mode 2, with implementation of the
requirements of License Condition 2 by July 1, 1997, and December 31,
1998, and with implementation of License Condition 3 at the next
updated safety analysis report update following completion of License
Condition 2, but no later than June 1, 1999.
Amendment Nos.: 128 and 120.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the licenses.
Public comments requested as to proposed no significant hazards
consideration (NSHC): Yes (62 FR 5857 dated February 7, 1997). This
notice provided an opportunity to submit comments on the Commission's
proposed NSHC determination. The notice also provided for an
opportunity to request a hearing by March 10, 1997, but indicated that
if the Commission makes a final NSHC determination, any such hearing
would take place after issuance of the amendments. Because of the
significant revisions to the licensee's original application, NRC also
published a public notice of the proposed amendments, issued a proposed
finding of no significant hazards consideration, and requested that any
comments on the proposed finding be provided to the staff by close of
business on March 20, 1997. The notice was published in the St. Paul
Pioneer Press on March 15, 1997, the Minneapolis Star Tribune on March
16, 1997, and the Red Wing Republican Eagle on March 17, 1997. No
comments have been received. The Commission's related evaluation of the
amendments, finding of exigent circumstances, and final determination
of NSHC are contained in a Safety Evaluation dated March 25, 1997.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
NRC Project Director: John N. Hannon.
Dated at Rockville, Maryland, this 2nd day of April, 1997.
For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear
Reactor Regulation.
[FR Doc. 97-8916 Filed 4-8-97; 8:45 am]
BILLING CODE 7590-01-P