[Federal Register Volume 63, Number 165 (Wednesday, August 26, 1998)]
[Notices]
[Pages 45521-45535]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-22766]
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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
[[Page 45522]]
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 August 3, 1998, through August 14, 1998. The
last biweekly notice was published on August 12, 1998 (63 FR 43200).
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 and
Directives Branch, Division of Administration 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 September 25, 1998, 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 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
[[Page 45523]]
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: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. 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.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert County,
Maryland
Date of amendments request: July 20, 1998.
Description of amendments request: The amendment incorporates the
changes described below into the Technical Specifications (TS) for
Calvert Cliffs Unit 2. Currently, Calvert Cliffs has four emergency
diesel generators (EDGs), two per Unit, to provide the onsite emergency
power supply for both Units. The Unit 2 EDGs rely on the Service Water
(SRW) System to provide their cooling water. During the Unit 2 1999
Refueling Outage, Baltimore Gas and Electric Company will replace the
SRW heat exchangers on Unit 2. During the period of the replacement, no
SRW cooling will be available for Unit 2. Therefore, both Unit 2 EDGs
would be inoperable during the replacement work. Unit 1 will continue
at full power operation during the Unit 2 refueling outage.
The loss of both EDGs on Unit 2 presents several challenges. First,
a number of outage activities require an EDG to be operable. BGE
proposes to provide an alternate cooling water supply to maintain the
EDGs operable to fulfill the TS requirements. One EDG will be provided
with cooling water from the Unit 1 SRW System. The other EDG will be
provided with cooling water from an independent external cooling
system. Second, Unit 1 is scheduled to be in Mode 1 operation during
this time. The No. 12 Control Room Emergency Ventilation System, No. 12
Control Room Emergency Temperature System, and a Hydrogen Analyzer are
affected by this work because they obtain their emergency power from a
Unit 2 EDG. These components support Unit 1 continued operation.
Therefore, the loss of both Unit 2 EDGs would impact operations on both
units.
There are several issues associated with this change that create an
Unreviewed Safety Question (USQ) as defined by 10 CFR 50.59. There is
an increase in the probability of a malfunction due to the use of an
independent cooling system that is non-safety-related and unprotected
from seismic or tornado events. The reliance of a Unit 2 EDG on Unit 1
SRW results in the increase of the probability of a malfunction, also.
Additionally, these SRW lineups affect the probability of a malfunction
for other equipment that relies on SRW during an outage. The approval
of these USQs, will permit a TS Bases change to the description of an
operable EDG while Unit 2 is in Modes 5 and 6.
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. Would not involve a significant increase in the probability
or consequences of an accident previously evaluated.
The EDGs are used to mitigate the consequences of an accident.
They are designed to start and load safety-related loads within a
specified time period. There are two EDGs for Unit 2. Only one is
required during the refueling outage, since a single failure
criterion does not apply during this time. However, it is desirable
for defense-in-depth and shutdown safety reasons to keep both EDGs
operable. Additionally, one of the EDGs supports operable equipment
on Unit 1 that remains at power. We are proposing an amendment that
would allow the EDGs to continue to be operable with an alternate
cooling water supply. Other than the change in cooling water supply,
we are not affecting or modifying the operation of the EDGs. The
EDGs are not an accident initiator for any previously evaluated
accident. Therefore, the proposed change does not involve an
increase in the probability of an accident previously evaluated.
The EDGs are designed to mitigate the consequences of an
accident. They will continue to perform that function while being
supplied with an alternate source of cooling water. The consequences
of a design basis accident during the period when the alternate
cooling water is being supplied is not increased because the
operation of the EDGs has not been adversely affected. Any
additional electrical loads (such as cooling tower pumps and fans)
or additional cooling loads (such as additional SRW flow to the No.
2A EDG) have been evaluated and found to be acceptable under
conditions postulated to exist during the outage. Therefore, the
proposed change does not significantly increase the consequences of
an accident previously evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated.
The EDGs are not being modified by this proposed change nor will
any unusual operator actions be required. The EDGs will continue to
operate in the same manner as before. However, the cooling water
supplies have been altered and were evaluated under the provisions
of 10 CFR 50.59 and determined to result in a USQ. These USQs are
evaluated below.
The first identified USQ is due to the realignment of a Unit 1
SRW subsystem to also support a Unit 2 EDG (2A). This alignment will
rely on two control valves (one to each EDG) to function properly in
order to provide adequate SRW flow to both EDGs. If one of the
valves should fail open, it may result in insufficient SRW flow or
increased SRW temperatures, as the EDGs share the same cooling
supply. This is an increase in the probability of a malfunction
because the operability of a EDG relies on both control valves
performing properly. We believe that this is an acceptable condition
because the control valves and their air supply are safety-related
and will be performing their design function. The control valves are
not being modified by the temporary configuration nor will any
operator action be required. The control valves will continue to
operate in the same manner. Therefore, because the malfunction is
the same as previously identified for these valves and only the
probability has increased, a new or different type of accident has
not been created.
The next USQ identifies a condition where a Unit 2 EDG is
dependent on a Unit 1 EDG for cooling water. The Unit 1 EDG powers
the pump for the cooling water system that will now provide cooling
to both EDGs. Although the consequences of a loss of cooling water
is the same (i.e., the EDG fails), the probability of a malfunction
for the Unit 2 EDG has increased because it now depends on the Unit
1 EDG to maintain its operability. We believe that this is an
acceptable condition because the Unit 1 EDG is safety-related and is
proven reliable through testing.
[[Page 45524]]
Additionally, the EDG will not be operated in a manner different
than it is currently. It is not being modified by the proposed
change nor will any additional operator actions be required. A
failure analysis shows that failure of the No. 1B EDG will not
result in the total loss of any safety function for either unit.
Therefore, the possibility of a new or different type of accident
has not been created.
A USQ has been identified related to the use of a temporary
cooling system to provide cooling to an EDG. The cooling system what
is proposed is not safety-related and is not protected from natural
phenomenon. This leads to an increase in the probability of a
malfunction because the cooling system is more likely to fail than a
safety-related, protected system. We believe that this is an
acceptable condition for the limited time we propose to use the
cooling system. The consequences of a cooling system failure are no
different than those of a failure of the SRW System. The events most
likely to cause the cooling system to fail are seismic events and
severe weather. Severe weather is not highly probable during this
time of year. Significant seismic events are not probable on this
part of the east cost. The cooling tower has been used before at
Calvert Cliffs to support testing of the EDGs during outages. The
cooling tower will have enhanced design features that will improve
its reliability, such as two pumps. The piping provided to and from
the cooling system will be steel and will be provided with flexible
joints making it rugged and flexible. Additionally, the cooling
tower will be placed close to the Auxiliary Building and the makeup
water piping will be run underground for part of its length. These
measures help to protect the cooling tower and its piping from
severe weather events. The EDG is not being altered by this
temporary configuration. It will continue to operate as before. No
additional operator action is required for the cooling tower to
perform its function. Therefore, the possibility of a new or
different type of accident has not been created.
This USQ exists because the piping from the cooling tower to the
EDG is not safety-related and could break, causing a flood in the
EDG room. This creates an increase in the probability of a
malfunction because of the increased probability of flooding in the
room. We believe that this increase is acceptable because the piping
is constructed from rugged materials and is flexibly connected to
the EDG. This reduces the chance that flooding will occur. If
flooding were to occur and the contents of the cooling system were
spilled into the room, it would not impact safety-related components
in the room because the water would not be deep enough. Therefore,
the possibility of a new or different accident has not been created.
Therefore, the possibility of a new or different type of
accident from any accident previously evaluated has not been
created.
3. Would not involve a significant reduction in a margin of
safety. The operability of the EDGs in Modes 5 and 6 ensures that
emergency power is available to mitigate the consequences of a fuel
handling accident and a boron dilution accident. Additionally, it
provides emergency power for shutdown cooling and spent fuel pool
cooling. One of the Unit 2 EDGs provides power to the shared Control
Room Emergency Ventilation System, Control Room Emergency
Temperature System, and the Hydrogen Analyzer needed to Support Unit
1 power operation. The proposed changes do not affect the function
of the EDGs. Because of the increased probability of a malfunction
of equipment important to safety (SRW support for the EDGs), the
margin of safety is reduced. However, the reduction is not
significant. As described above, each USQ has been evaluated and
determined to not have a significant impact on safety.
To provide additional assurance that all reasonable steps have
been taken to ensure the operability of the Unit 2 EDGs while in the
temporary configuration, the following actions will be taken in
addition to the installation of the temporary modifications as
described above:
To prevent the loss of the normal power supply to the Control
Room Emergency Ventilation System and Control Room Emergency
Temperature System, we will restrict maintenance activities on three
of the four offsite transmission lines until the Unit 2 EDGs are
returned to normal configuration.
To monitor risk, Unit 1 and 2 equipment taken out-of-service
during this period will be evaluated in the Unit 1 weekly quarterly
system schedule evaluations.
To ensure that weather-related events cannot cause a loss of all
emergency power on Unit 2 during periods of reduced inventory, the
No. 2A EDG will remain operable during reduced inventory periods.
To ensure that backup power is available to any of the safety-
related buses, the No. 0C Diesel Generator will not be taken out-of
service for planned maintenance and will remain available to be
connected to any of the safety-related buses.
We believe that the reduction in the margin of safety
represented by this temporary license amendment is not significant
based on our evaluation and management of plant risk, the
reliability of the EDGs, the availability of redundant EDGs, the
availability of the Station Blackout Diesel Generator and the
mitigating features described above. 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: S. Singh Bajwa, Director.
Duke Energy Corporation (DEC or licensee), Docket Nos. 50-369 and 50-
370, McGuire Nuclear Station, Units 1 and 2, Mecklenburg County, North
Carolina
Date of amendment request: May 27, 1997, as supplemented by letters
dated March 9, March 20, April 20, June 3, June 24, July 7, July 21,
and July 22, 1998.
Description of amendment request: The proposed amendments would
revise the Technical Specifications (TS) of each unit to conform with
NUREG-1431, Revision 1, ``Standard Technical Specifications--
Westinghouse Plants.'' The Commission had previously issued a Notice of
Consideration of Issuance of Amendments in the Federal Register on July
15, 1997 (62 FR 37940) covering all the proposed changes that were
indeed within the scope of NUREG-1431. In DEC's May 27, 1997,
submittal, there are proposed changes that are beyond the scope of
NUREG-1431, which were, thus, not covered by the staff's July 15, 1997,
notice. The following description and no significant hazard analysis
covers a beyond-scope change.
The licensee proposed to change Section 3.4.6.1 regarding reactor
coolant leakage detection systems; a system comprising diverse
instruments such as gaseous radioactivity monitoring, containment floor
and equipment sump monitoring, etc. In addition to the instruments
specified by this section, the plant has other installed instruments
such as monitors for humidity, temperature, etc., which can provide
indication for reactor coolant leakage. Currently, this specification
allows operation up to 30 days if the containment floor and equipment
sump monitoring system is inoperable. The proposed change would impose
a requirement to perform a precision water balance of the reactor
coolant system every 24 hours during this period. The proposed change
would also reduce the number of monitors required operable provided
compensatory measures are performed or diverse instruments continue to
be available.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analyses of the issue of no significant hazards
consideration for each of the above proposed changes. The NRC staff has
reviewed the licensee's analyses against the standards of 10 CFR
50.92(c). The NRC staff's analysis is presented below.
1. Will the change involve a significant increase in the
probability or consequences of an accident previously
[[Page 45525]]
evaluated? The proposed change will not affect the safety function of
the subject systems. There will be no direct effect on the design or
operation of any plant structures, systems, or components. No
previously analyzed accidents were initiated by the functions of these
systems, and the systems were not factors in the consequences of
previously analyzed accidents. Therefore, the proposed change will have
no impact on the consequences or probabilities of any previously
evaluated accidents.
2. Will the change create the possibility of a new or difference
kind of accident from any accident previously evaluated?
The proposed change would not lead to any hardware or operating
procedure change. Therefore, no new equipment failure modes or
accidents from those previously evaluated will be created.
3. Will the change involve a significant reduction in a margin of
safety? Margin of safety is associated with confidence in the design
and operation of the plant. The proposed change to the TS do not
involve any change to plant design, operation, or analysis. Thus, the
margin of safety previously analyzed and evaluated is maintained.
Based on this analysis, it appears that the three standards of 10
CFR 50.92(c) are satisfied for each of the proposed change. Therefore,
the NRC staff proposes to determine that the amendment request involves
no significant hazards consideration.
Local Public Document Room location: J. Murrey Atkins Library,
University of North Carolina at Charlotte, 9201 University City
Boulevard, Charlotte, North Carolina.
Attorney for licensee: Mr. Albert Carr, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina.
NRC Project Director: Herbert N. Berkow.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: July 21, 1998.
Description of amendment request: The proposed change to the
Technical Specifications would: (1) modify Specification 6.2.2.2(a) to
provide some flexibility to accommodate unexpected absence of on-duty
shift crew members, (2) eliminate reference to the Manager, Plant
Operations in Specification 6.2.2.2(j) as the position has been
eliminated, (3) reduce the maximum time in which to forward audit
reports to the responsible manager from 60 days to 30 days, (4) replace
the term ``Vice President'' with the term ``Corporate Officer'' in
several places in Section 6, and (5) correct several typographical
errors.
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. State the basis for the determination that the proposed
activity will or will not increase the probability of occurrence or
consequences of an accident.
The activity does not alter the design, function or manner of
operation of any structures, systems or components. Therefore, this
activity does not increase the probability or consequences of an
accident.
2. State the basis for the determination that the activity does
or does not create the possibility of an accident or malfunction of
a different type than any previously identified in the SAR.
The activity does not alter the design, function, or manner of
operation of any structures, systems or components. Therefore, this
activity does not create the possibility of an accident or
malfunction of a different type than any previously identified in
the SAR.
3. State the basis for the determination that the margin of
safety is not reduced.
The activity does not alter the design, function or manner of
operation of any structures, systems or components. In addition, a
decrease in staff for a short period of time on limited occasions is
not safety significant and permitted by 10 CFR 50.54 (m). Therefore,
this activity will not reduce 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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of amendment request: July 23, 1998.
Description of amendment request: Amend facility license to
establish that the existing Safety Limit Minimum Critical Power Ratio
(SLMCPR) contained in Technical Specification 2.1.A is applicable for
the next operating cycle (Cycle 17).
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 TS change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The derivation of the Cycle 17 SLMCPR for Oyster Creek for
incorporation into the TS, and its use to determine cycle-specific
thermal limits, has been performed using NRC-approved methods.
Additionally, interim implementing procedures, which incorporate
cycle-specific parameters, have been used. Based on the use of these
calculations, the Cycle 17 SLMCPR of 1.09 will not increase the
probability or consequences of an accident.
The basis of the MCPR 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. A SLMCPR of 1.09 preserves
adequate margin to transition boiling and fuel damage in the event
of a postulated accident. The probability of fuel damage is not
increased.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
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. The
limit cannot create the possibility of any new type of accident. The
Cycle 17 SLMCPR has been calculated using NRC-approved methods.
Additionally, interim procedures, which incorporate cycle-specific
parameters, have been used. Therefore, the proposed TS change does
not create the possibiliy of a new or different kind of accident,
from any accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the TS Bases will remain the
same. The Cycle 17 SLMCPR is calculated using NRC-approved methods,
which are in accordance with the current fuel design and licensing
criteria. Additionally, interim implementing procedures, which
incorporate cycle-specific parameters, have been used. The MCPR
Safety Limit remains high enough to ensure that greater than 99.9%
of all fuel rods in the core will avoid transition boiling if the
limit is not violated, thereby preserving fuel cladding integrity.
Therefore, the proposed TS change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this
[[Page 45526]]
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: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Cecil O. Thomas.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: August 6, 1998.
Description of amendment request: The proposed amendment would
allow changes to the Updated Safety Analysis Report (USAR) to reflect
the as-built configuration of the reactor building isolation dampers.
These changes would clarify the USAR discussion of secondary
containment isolation and revise the calculated offsite dose
consequences resulting from a postulated refueling accident. No changes
to the Technical Specifications (TS) are required; the TS Bases,
Sec. 3.6.4.2, will be revised under the licensee's Bases control
program to reflect the changes in the USAR 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:
The enclosed proposed license amendment for the as-built design
of the Secondary Containment (Reactor Building) isolation dampers is
judged to involve no significant hazards based on the following:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The existing plant design does not involve a significant
increase in the probability of an accident previously evaluated in
the Updated Safety Analysis Report (USAR). The current configuration
does not affect the performance and reliability of the Secondary
Containment and the Reactor Building Isolation and Control System or
any system interface in a way that could lead to an accident
occurring. The current configuration and analysis do not affect any
accident precursors or initiators, and therefore, does not increase
the probability of an accident.
The present plant configuration also does not involve a
significant increase in the consequences of an accident previously
evaluated in the USAR. The current design will require a
clarification to the Secondary Containment safety design basis as
described in the USAR to reflect the as-built configuration and
analysis of the plant by stating that the Reactor Building Isolation
and Control System is designed to limit the release of fission
products through the normal ventilation discharge path during a
postulated Refueling Accident.
The original analysis determined that the consequences of the
Refueling Accident were significantly less than 1 Rem to the thyroid
and whole body (maximum off-site dose). When this analysis was
revised to account for the 90 second motor-operated damper closure
time, the calculated whole body off-site dose increased, but was
still less than 1 Rem; the calculated off-site dose to the thyroid,
however, increased to 2.7 Rem. While this change in the analysis
represents an order of magnitude increase in consequences (thyroid
dose increase from 17 milliRem to 2.7 Rem), the actual increase is
minimal because this increase in consequences is still less than 1
percent (1%) of the limits specified in 10 CFR 100. Thus the
consequences still remain well within the regulatory threshold
specified in 10 CFR 100 and thus pose no undue hazard to the health
and safety of the public. This proposed amendment does not alter the
Control Room dose from that which was submitted to the NRC in
support of Amendment 167.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
This proposed license amendment is administrative in nature in
that it reflects the effects of a revised analysis for the Refueling
Accident, which is an accident previously analyzed as a Design Basis
Accident (DBA) in the SAR, based on the present configuration of the
plant. The current configuration does not create the possibility of
a new or different kind of accident from any accident previously
evaluated in the USAR. The proposed license amendment does not
introduce any new equipment or hardware changes, nor does it require
existing equipment or systems to perform a different type of
function than they are presently designed to perform. The as-built
configuration does not introduce any new mode of plant operation,
thus there are no new accident failure paths created.
The as-built configuration does not affect any accident
precursors or initiators and does not create the possibility of a
new or different kind of accident.
3. Does not create a significant reduction in the margin of
safety.
The present plant configuration does not involve a significant
reduction in a margin of safety. Technical Specification Bases
section 3.2.D.2, Reactor Building Isolation and Standby Gas
Treatment (SGT) Initiation, states that the trip settings for the
Reactor Building exhaust plenum radiation monitors are based on
initiating normal ventilation system isolation and SGT System
operation so that none of the activity released during the refueling
accident leaves the Reactor Building via the normal ventilation
path, but rather all the activity is processed by the SGT System.
This basis statement remains true unless there is a single failure
of the air-operated Secondary Containment isolation damper. Under
single failure conditions there would be the potential for a limited
release through the normal ventilation system prior to complete
isolation of the secondary containment and initiation of the SGT
System.
The significance of this change is minimal, as Technical
Specification requirements to isolate Secondary Containment are
still met. The overall function of the Secondary Containment and
Reactor Building Isolation and Control System, in conjunction with
other accident mitigation systems, is to limit fission product
release during and following postulated DBAs. High radiation in the
Secondary Containment exhaust is an indication of possible gross
failure of the fuel cladding, possibly due to a Refueling Accident.
The trip settings for the Reactor Building (Secondary Containment)
radiation monitors are such that initiation of secondary containment
isolation and SGT would still occur in sufficient time (within 90
seconds of detection) to maintain postulated off-site releases well
within the limits of 10 CFR 100. As stated previously, the effects
of the 90 second motor-operated damper closure time on Control Room
dose have already been taken into consideration in the District's
submittals supporting Amendment 167.
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: Auburn Memorial Library, 1810
Courthouse Avenue, Auburn, NE 68305.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Project Director: John N. Hannon.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: August 4, 1998.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) relating to the Condensate
Storage Tank (CST) and also add a new TS section that would establish
requirements for the atmospheric steam dump valves (ASDVs) to assure
their operability. The applicable TS Bases section for the CST would
also be changed to reflect the proposed changes and a new TS Bases
section would be added to discuss the new TS section for the ASDVs.
Specifically, the proposed changes would modify TS 3.7.1.3, ``Plant
[[Page 45527]]
Systems--Condensate Storage Tank,'' by increasing the minimum required
CST level from 150,000 gallons to 165,000 gallons to account for the
discharge nozzle pipe elevation above the tank bottom and vortex
formation in the CST at the auxiliary feedwater supply piping entrance.
TS 3.7.1.7, ``Plant Systems--Atmospheric Steam Dump Valves,'' would be
added to provide the requirements necessary to assure that the ASDVs
will be available to either maintain the unit in hot standby or cool
down the unit to shutdown cooling entry conditions if the condenser
steam dump valves are not available. As previously noted, the TS Bases
would be modified to reflect the proposed 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. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change to increase the minimum required Condensate
Storage Tank (CST) level of Technical Specification 3.7.1.3 will
ensure sufficient water is available for the Auxiliary Feedwater
(AFW) System to function as designed to mitigate design basis
accidents. There will be no adverse effect on equipment important to
safety. Therefore, the proposed change will not result in a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed change to add a Technical Specification for the
Atmospheric Steam Dump Valves (ASDVs) will provide additional
assurance that the ASDVs will be available to either maintain the
unit in hot standby, or cool down the unit to Shutdown Cooling (SDC)
entry conditions if the condenser steam dump valves are not
available. The proposed change does not alter the way any structure,
system, or component functions. There will be no adverse effect on
any design basis accident previously evaluated or on any equipment
important to safety. Therefore, the proposed change will not result
in a significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes have no adverse effect on any of the design
basis accidents previously evaluated. Therefore, the license
amendment request does not impact the probability of an accident
previously evaluated nor does it involve a significant increase in
the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes will not alter the plant configuration (no
new or different type of equipment will be installed) or require any
new or unusual operator actions. They do not alter the way any
structure, system, or component functions and do not alter the
manner in which the plant is operated. The proposed changes 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. Involve a significant reduction in a margin of safety.
The proposed change to increase the minimum required CST level
will ensure the AFW System will function as designed to mitigate
design basis accidents. The proposed change to add a Technical
Specification for the ASDVs will provide additional assurance that
the ASDVs will be available, if needed. There will be no adverse
effect on equipment important to safety. Therefore, there will be no
significant reduction of margin of safety as defined in the Bases
for Technical Specifications affected by these proposed changes.
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, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
Connecticut.
NRC Deputy Director: William M. Dean.
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: March 18, 1998.
Description of amendment request: The proposed amendment would
revise the Bases for Technical Specification (TS) 3/4.6.2.1,
``Containment Spray System,'' of the combined technical specifications
for the Diablo Canyon Power Plant, Unit Nos. 1 and 2, to clarify that
containment spray is not required to be actuated during recirculation,
but may be actuated at the discretion of the Technical Support Center.
Additionally, the Bases would be clarified to state that the ability to
spray containment using the residual heat removal (RHR) system is
demonstrated by opening the RHR Spray Ring Cross Connect Valve 9003 A
or B. The Bases will also be clarified to state that flow to the spray
headers can be established with only one operable RHR pump by closing
the cold leg discharge valve 8809 A or B.
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.
Containment spray (CS) in the recirculation mode of post-loss-
of-coolant accident (LOCA) safety injection (SI) is used only after
the accident has already occurred. Its availability or
unavailability is unrelated to, and is not a precursor for, an
accident that has already been initiated. The availability or
unavailability of CS recirculation spray does not involve any
physical change in plant systems, structures, or components, and
there is no change in preaccident operating procedures, so there is
no change to the probability of an accident occurring as a result of
any such changes. The recirculation mode of emergency core cooling
is only used following a LOCA; therefore, an evaluation of the
effects of the use or absence of CS in the recirculation mode
applies only to a LOCA and not to any other type of accident
analyzed in the Final Safety Analysis Report (FSAR).
The peak post-LOCA containment pressure and temperature
conditions occur prior to the recirculation phase of SI, and are not
affected by CS operation during the recirculation mode of SI. The
long term pressure and temperature profiles are slightly increased
if recirculation spray is unavailable but are still within the dose
analysis and equipment qualification requirements. There is no
effect on the offsite dose analysis or on equipment operability.
If CS is not operated in the recirculation mode, there is no
reduction in the amount of emergency core cooling system (ECCS)
water pumped into the reactor vessel. Since the flow to the reactor
is not reduced, core cooling is not adversely affected if
recirculation spray is not used. If recirculation spray is used
under Technical Support Center (TSC) direction with only one train
of residual heat removal (RHR) in operation, ECCS flow to the
reactor will be reduced, but analysis has shown that the flow to the
reactor in this situation is still in excess of that needed to
supply the required core cooling. Therefore, although it is not
required, it would still be possible to establish CS in the
recirculation mode with only one train of RHR in operation, if
considered desirable by the TSC.
From the above discussion, it can be seen that the consequences
of an accident analyzed in the FSAR are not increased because the
absence of recirculation spray has no effect on the dose analyses
and the effect on other accident parameters is within limits.
Therefore, the proposed changes do not involve a significant
increase in the
[[Page 45528]]
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 possibility of a malfunction of a different type than
previously evaluated is not created for the following reasons:
Data provided in the FSAR can be used to determine that the
iodine removal function for the CS system is completed in
approximately 26 minutes, and prior to completing switchover to the
recirculation mode after a LOCA. The statements in previous
revisions of the FSAR that recirculation spray will continue for 2
hours to remove iodine are considered to be descriptive in nature,
explaining an additional capability of the CS system, but not relied
upon or evaluated in the FSAR.
The post-LOCA containment environmental conditions without
recirculation spray remain bounded by those for which safety-related
equipment inside containment is qualified; therefore, there is no
resulting increase in the probability that it will malfunction.
There is no other new mechanism created by the unavailability of
recirculation spray that would lead to any greater probability of
malfunction of safety-related equipment.
The peak post-LOCA containment pressure and temperature
conditions occur prior to the recirculation phase of SI, and are not
affected by CS operation during the recirculation mode of SI. Also,
the Diablo Canyon Power Plant (DCPP) design bases and accident
analyses do not assume any contribution to post-accident containment
hydrogen mixing from recirculation spray. The DCPP design basis has
always assumed that hydrogen mixing is achieved by containment fan
cooler unit operation alone.
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.
Technical Specification (TS) 3/4.6.2.1, ``Containment Spray
System,'' requires the operability of two trains of CS with each
train capable of taking suction from the refueling water storage
tank (RWST) and transferring spray function to an RHR train taking
suction from the containment sump. With the proposed changes, the
capability to perform the required alignment remains unaffected.
However, the ability to actually provide CS in the recirculation
mode of SI is limited by procedure in the event of failure of a
train of auxiliary saltwater, component cooling water, or RHR. This
does not affect the margin of safety as defined in the TS Bases. The
Bases for CS operability are to ensure pressure reduction, cooling
capability, and iodine removal from the containment atmosphere
consistent with the assumptions used in the safety analyses.
All pressure reduction, cooling, and iodine removal parameters
assumed in the accident analyses continue to bound those resulting
in the event that recirculation spray is not used. The accident
analyses require that the peak post-accident pressure does not
exceed 47 psig, and that post-accident pressure be reduced to less
than half the peak within 24 hours. These requirements are still
met, but the long term pressure is slightly higher. Since these
requirements are based on minimizing leakage rates and on
environmental qualification concerns, and since the leakage rate in
the offsite dose analysis and pressures for which safety-related
equipment inside containment is qualified still bound the analysis
results, a slightly higher long term pressure has no effect on
safety margins. Although the long term temperature profile increases
slightly with no recirculation spray, the equipment is still
environmentally qualified for these temperatures, so again margin is
maintained. The use of recirculation spray is not credited in the
offsite or control room dose analyses since the containment
atmospheric iodine decontamination factor reaches 1000 prior to the
time recirculation spray is placed in service, so there is no loss
of margin in the offsite and control room dose analyses. None of the
accident analysis limits are exceeded in the absence of
recirculation spray.
The function of CS to inject NaOH into the containment
atmosphere and sump is not affected by the proposed changes. The
same amount of RWST water will be pumped into the containment via
the CS system for a given size LOCA with or without recirculation
spray, so the same amount of NaOH is injected into the containment,
and hence there is no effect on sump pH, iodine retention, or the
dose analysis.
In the event that recirculation spray is established under TSC
direction with only one train of RHR in operation, there is no
reduction in the margin of safety from the resulting reduced flow to
the core since analysis has demonstrated that even with no RHR flow
to the RCS, the resulting flow to the core will still be greater
than that required to maintain adequate core cooling and maintain
peak clad temperatures within limits.
The functions specified for the CS system in the TS Bases are to
ensure post-accident pressure reduction, cooling capability, and
iodine removal from the containment atmosphere consistent with the
assumptions used in the safety analyses. Since these functions are
maintained within the limits of the safety analyses even in the
absence of recirculation spray, the operability of the CS system as
required by TS 3.6.2.1 is maintained.
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 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 &
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Project Director: William H. Bateman.
Pennsylvania Power and Light Company, Docket No. 50-388, Susquehanna
Steam Electric Station, Unit 2, Luzerne County, Pennsylvania
Date of amendment request: August 5, 1998.
Description of amendment request: The amendment to Unit 2 Technical
Specifications (TS) involves the addition of a new section entitled
``Oscillation Power Range Monitoring (OPRM) Instrumentation'' and
revisions to Section 3.4.1 ``Recirculation Loops Operating'' to remove
the specifications related to thermal power stability which will not be
required after the installation of the OPRM instrumentation. Unit 2 is
currently operating under Interim Corrective Actions (ICAs) defined in
TS 3.4.1 that specify restrictions on plant operation and actions by
operators in response to instability events. The OPRM system provides
an automatic long-term solution to the instability issue and eases the
burden on the operator.
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.
This proposal does not involve an increase in the probability or
consequences of an accident previously evaluated.
The OPRM most directly affects the [Average Power Range Monitor]
APRM and [Local Power Range Monitor] LPRM portions of the Power
Range Neutron Monitoring system. Its installation does not affect
the operation of these sub-systems. None of the accidents or
equipment malfunctions affected by these sub-systems are affected by
the presence or operation of the OPRM.
The APRM channels provide the primary indication of neutron flux
within the core and respond almost instantaneously to neutron flux
changes. The APRM Fixed Neutron Flux-High function is capable of
generating a trip signal to prevent fuel damage or excessive reactor
pressure. For the [American Society of Mechanical Engineers] ASME
overpressurization protection analysis in [Final Safety Analysis
Report] FSAR Chapter 5, the APRM Fixed Neutron Flux-High function is
assumed to terminate the main steam isolation valve closure event.
The high flux trip, along with the safety/relief valves, limit the
peak reactor pressure vessel
[[Page 45529]]
pressure to less than the ASME Code limits. The control rod drop
accident (CRDA) analysis in Chapter 15 takes credit for the APRM
Fixed Neutron Flux-High function to terminate the CRDA. The
Recirculation Flow Controller Failure event (pump runup) is also
terminated by the high neutron flux trip. The APRM Fixed Neutron
Flux-High function is required to be OPERABLE in MODE 1 where the
potential consequences of the analyzed transients could result in
the Safety Limits (e.g., [Minimum Critical Power Ratio] MCPR and
Reactor pressure) being exceeded.
The installation of the OPRM equipment does not increase the
consequences of a malfunction of equipment important to safety. The
APRM and [Reactor Protection System] RPS systems are designed to
fail in a tripped (fail safe) condition; the OPRM will have no
affect on the consequence of the failure of either system. An
inoperative trip signal is received by the RPS any time an APRM mode
switch is moved to any position other than Operate, an APRM module
is unplugged, the electronic operating voltage is low, or the APRM
has too few LPRM inputs. These functions are not specifically
credited in the accident analysis, but are retained for the RPS as
required by the NRC approved licensing basis.
The OPRM allows operation under current operating conditions
presently restricted by the current Technical Specifications by
providing automatic suppression functions in the area of concern in
the event an instability occurs. The consequences of any accident or
equipment malfunction are not increased by operating under those
conditions. Although protected by the OPRM from thermal-hydraulic
core instabilities above 30% core power, operation under natural
core recirculation conditions is not allowed. No accidents or
transients of a type not analyzed in the FSAR are created by
operating under these conditions with the protection of the OPRM
system.
This change does not increase the probability of an accident as
previously evaluated. The OPRM is designed and installed to not
degrade the existing APRM, LPRM, and RPS systems. These systems will
still perform all of their intended functions. The new equipment is
tested and installed to the same or more restrictive environmental
and seismic envelopes as the existing systems.
The new equipment has been designed and tested to the
electromagnetic interference (EMI) requirements of Reference 2,
which assures correct operation of the existing equipment. The new
system has been designed to single failure criteria and is
electrically isolated from equipment of different electrical
divisions and from non-1E equipment. The electrical loading is
within the capability of the existing power sources and the heat
loads are within the capability of existing cooling systems. The
OPRM allows operation under operating conditions presently forbidden
or restricted by the current Technical Specifications. No other
transient or accident analysis assumes these operating restrictions.
Based upon the analysis presented above, PP&L concludes that the
proposed action does not involve an 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.
This proposal does not create the probability of a new or
different type of accident from any accident previously evaluated.
The OPRM system is a monitoring and accident mitigation system that
cannot create the possibility for an accident.
The OPRM will allow operation in conditions currently restricted
by the current Technical Specifications. Although protected by the
OPRM from thermal-hydraulic core instabilities above 30% core power,
operation under natural circulation conditions is not allowed. No
accidents or transients of a type not analyzed in the FSAR are
created by operating under these conditions with the protection of
the OPRM system. No new failure modes of either the new OPRM
equipment or of the existing APRM equipment have been introduced.
Quality software design, testing, implementation and module self-
health testing provides assurance that no new equipment malfunctions
due to software errors are created. The possibility of an accident
of a new or different type than any evaluated previously is not
created.
The new OPRM equipment is designed and installed to the same
system requirements as the existing APRM equipment and is designed
and tested to have no impact on the existing functions of the APRM
system. Appropriate isolation is provided where new interconnections
between redundant separation groups are formed. The OPRM modules
have been designed and tested to assure that no new failure modes
have been introduced.
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 change does not involve a significant reduction
in the margin of safety.
There has been no reduction in the margin of safety as defined
in the basis for the Technical Specifications. The OPRM system does
not negatively impact the existing APRM system. As a result, the
margins in the Technical Specifications for the APRM system are not
impacted by this addition.
Current operation under the ICAs provides an acceptable margin
of safety in the event of an instability event as the result of
preventive actions and Technical Specification controlled response
by the control room operators. The OPRM system provides an increase
in the reliability of the protection of the margin of safety by
providing automatic protection of the MCPR safety limit, while the
protection burden is significantly reduced for the control room
operators. This protection is demonstrated as described above, and
in the NRC reviewed and approved Topical Reports NEDO-32465-A and
CENPD-400-P-A.
Replacement of the ICA operating restrictions from Technical
Specifications with the OPRM system does not affect the margin of
safety associated with any other system or fuel design parameter.
Therefore, the change 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: Robert A. Capra.
Southern Nuclear Operating Company, Inc., 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 amendments request: July 22, 1998.
Description of amendments request: The proposed amendments would
change Technical Specification Tables 3.3.6.1-1 and 3.3.6.2-1 by
increasing the Allowable Values for the high radiation trip for the
exhaust monitors for the reactor building and the refueling floor.
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. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The Unit 1 and Unit 2 reactor building and refueling floor
ventilation exhaust radiation monitors perform no function in
preventing, or decreasing the probability of, a previously evaluated
accident. The monitors are designed to monitor ventilation exhaust
for indications of a release of radioactive material resulting from
a design basis accident and initiate appropriate protective actions.
Because the proposed changes affect only the ventilation exhaust
radiation monitors, the probability of an accident previously
evaluated remains the same.
The function of the reactor building and the refueling floor
ventilation exhaust radiation monitors, in combination with other
accident mitigation systems, is to limit fission product release
during and following postulated design basis accidents. The proposed
new Allowable Values for the high radiation trip will continue to
ensure the offsite doses resulting from a design basis accident do
not exceed the NRC-approved
[[Page 45530]]
licensing basis and FSAR [Final Safety Analysis Report] limits.
Therefore, the proposed changes do not involve a significant
increase in the consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed changes increase the radiation level at which the
ventilation exhaust monitors actuate; however, the manner in which
their actuation logic functions and the systems that isolate or
actuate as a result are unaffected by the proposed changes.
Furthermore, the ventilation exhaust monitors will continue to
perform their design function of limiting offsite doses to NRC-
approved licensing basis and FSAR limits at the higher Allowable
Values. Therefore, the proposed changes cannot create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The Bases for Unit 1 and Unit 2 Technical Specification Tables
3.3.6.1-1 and 3.3.6.2-1 state that the Allowable Values for the
reactor building and refueling floor ventilation exhaust radiation
monitors ``are chosen to ensure radioactive releases do not exceed
offsite dose limits.'' The proposed Allowable Values ensure the
radiation monitors actuate at a radiation level sufficient to ensure
offsite doses are within the NRC-approved licensing basis and FSAR
limits. The proposed Allowable Values comply with the margin of
safety defined in the Technical Specifications Bases for the
ventilation exhaust radiation monitors; therefore, the proposed
changes do not reduce 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: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC.
NRC Project Director: Herbert N. Berkow.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: July 7, 1998.
Description of amendment request: The proposed amendment would
revise the spent fuel pool criticality analysis and rack utilization
schemes by allowing credit for spent fuel pool soluble boron.
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 presence of soluble boron in the spent fuel pool water for
criticality control does not increase the probability of a fuel
assembly drop accident in the spent fuel pool. The handling of the
fuel assemblies in the spent fuel pool has always been performed in
borated water.
The criticality analysis shows the consequences of a fuel
assembly drop accident in the spent fuel pool are not affected when
considering the presence of soluble boron. The rack Keff
[K effective] remains less than or equal to 0.95.
There is no increase in the probability of an accident. The
proposed change does allow a greater number of fuel storage
configurations in the spent fuel pool. While this could increase the
probability of a fuel misloading, the presence of sufficient soluble
boron in the spent fuel pool precludes criticality as a result of
the misloading. Fuel assembly placement will continue to be
controlled pursuant to approved fuel handling procedures and will be
in accordance with the Technical Specification spent fuel rack
storage configuration limitations.
There is no increase in the consequences of the accidental
misloading of spent fuel assemblies into the spent fuel pool racks.
The criticality analyses demonstrate that the pool Keff
will remain less than or equal to 0.95 following an accidental
misloading due to the boron concentration of the pool. The proposed
Technical Specification limitation will ensure that an adequate
spent fuel pool boron concentration is maintained.
There is no increase in the probability of the loss of normal
cooling to the spent fuel pool water when considering the presence
of soluble boron in the pool water for subcriticality control since
a high concentration of soluble boron has always been maintained in
the spent fuel pool water.
Reactivity changes due to spent fuel pool temperature changes
have been evaluated. The basic case criticality analysis covers a
``normal'' spent fuel pool temperature range of 50 degrees F to 160
degrees F. Spent fuel pool temperature accidents are considered
outside the normal temperature range extending from 32 degrees F to
240 degrees F. In all spent fuel pool temperature accident cases,
sufficient reactivity margin is available to the 0.95
Keff limit without requiring additional soluble boron
above the base case level. Because adequate soluble boron will be
maintained in the spent fuel pool water to maintain Keff
less than or equal to 0.95, the consequences of a loss of normal
cooling to the spent fuel pool will not be increased.
Therefore, based on the conclusions of the above analysis, 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.
Spent fuel handling accidents are not new or different types of
accidents, they have been analyzed in Section 15.7.4 of the Updated
Final Safety Analysis Report.
Criticality accidents in the spent fuel pool are not new or
different types of accidents, they have been analyzed in the Updated
Final Safety Analysis Report and in Criticality Analysis reports
associated with specific licensing amendments for fuel enrichments
up to and exceeding the nominal 4.95 weight percent U\235\ [Uranium-
235] that is assumed for the proposed change.
Current Technical Specifications contain limitations on the
spent fuel pool boron concentration. The actual boron concentration
in the spent fuel pool has been maintained at a higher value. The
proposed changes to the Technical Specifications establish new boron
concentration requirements for the spent fuel pool water consistent
with the results of the new criticality analysis (Attachment 2).
Since soluble boron has always been maintained in the spent fuel
pool water, and is currently required by Technical Specifications,
the implementation of this new requirement will have little effect
on normal pool operations and maintenance. A dilution of the spent
fuel pool soluble boron has always been a possibility; however, it
was shown in the spent fuel pool dilution evaluation (Attachment 3)
that there are no credible dilution events for which the spent fuel
pool Keff could increase to greater than 0.95. Therefore,
the implementation of new limitations on the spent fuel pool boron
concentration will not result in the possibility of a new kind of
accident.
The proposed changes to Technical Specifications 3.9.13, 4.9.13,
and 5.6 continue to specify the requirements for the spent fuel rack
storage configurations. Since the proposed spent fuel pool storage
configuration limitations will be similar to the current ones, the
new limitations will not have any significant effect on normal spent
fuel pool operations and maintenance and will not create any
possibility of a new or different kind of accident. Verifications
will continue to be performed to ensure that the spent fuel pool
loading configuration meets specified requirements.
The misloading of a fuel assembly in the required storage
configuration has been evaluated. In all cases, the rack
Keff remains less than or equal to 0.95. Removal of an
[sic] Rod Control Cluster Assembly from a checkboard storage
configuration has been analyzed and found to be bounded by the
misloading of a fuel assembly.
As discussed above, the proposed changes will not create the
possibility of a new or different kind of accident. There is no
significant change in plant configuration, equipment design or
equipment.
Under the proposed amendment, no changes are being made to the
racks themselves, any other systems, or to the physical structures
of the Fuel Handling
[[Page 45531]]
Building itself. 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 change does not involve a significant reduction
in a margin of safety.
The Technical Specification changes proposed by this License
Amendment Request and the resulting spent fuel storage operation
limits will provide adequate safety margin to ensure that the stored
fuel assembly array will always remain subcritical. Those limits are
based on a plant specific criticality analysis (Attachment 2)
performed in accordance with Westinghouse spent fuel rack
criticality analysis methodology.
While the criticality analysis utilized credit for soluble
boron, storage configurations have been defined using 95/95
Keff calculations to ensure that the spent fuel rack
Keff will be less than 1.0 with no soluble boron. Soluble
boron credit is used to offset uncertainties, tolerances, and off-
normal conditions and to provide subcritical margin such that the
spent fuel pool Keff is maintained less than or equal to
0.95.
The loss of substantial amounts of soluble boron from the spent
fuel pool which could lead to Keff exceeding 0.95 has
been evaluated (Attachment 3) and shown to be not credible. A safety
evaluation has been performed which shows that dilution of the spent
fuel pool boron concentration from 2500 ppm to 700 ppm is not
credible. Also, the spent fuel rack Keff will remain less
than 1.0 (with a 95/95 confidence level) with the spent fuel pool
flooded with unborated water. These safety analyses demonstrate a
level of safety comparable to the conservative criticality analysis
methodology required by Westinghouse WCAP-14416.
Based on the above evaluation, the South Texas Project concludes
that the proposed changes to the Technical Specifications involve no
significant hazards consideration.
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: 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 Energy Corporation, Docket No. 50-287, Oconee Nuclear Station,
Unit No. 3, Oconee County, South Carolina
Date of amendment request: July 20, 1998.
Description of amendment request: The proposed amendment would
extend, on a one-time basis, Technical Specification Surveillance
4.18.3 for hydraulic and mechanical snubber testing. The tests are
required to be performed at a frequency of 18 months, with a maximum
allowed frequency of 22 months, 15 days. The amendment would extend
this to a maximum of 25 months.
Date of publication of individual notice in Federal Register: July
27, 1998 (63 FR 40137).
Expiration date of individual notice: August 26, 1998.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
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.
CBS Corporation, Docket No. 50-22, Westinghouse Test Reactor, Waltz
Mill, Pennsylvania
Date of application for amendment: December 22, 1997 supplemented
on June 15, 1998.
Brief description of amendment: This amendment changes the legal
name of the licensee for the Westinghouse Test Reactor from
Westinghouse Electric Corporation to CBS Corporation.
Date of issuance: July 31, 1998.
Effective Date: July 31, 1998.
Amendment No: 7.
Facility Operating License No. TR-2: This amendment changes the
legal name of the licensee for the Westinghouse Test Reactor from
Westinghouse Electric Corporation to CBS Corporation.
Date of initial notice in Federal Register: July 15, 1998 (63 FR
38207).
The Commission has issued a Safety Evaluation for this amendment
dated July 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document: N/A.
Commonwealth Edison Company, Docket No. 50-249, Dresden Nuclear Power
Station, Unit 3, Grundy County, Illinois
Date of application for amendment: May 6, 1998.
Brief description of amendment: The proposed amendment would amend
Technical Specification (TS) 4.6.E to allow a one-time extension of the
40-month surveillance interval requirement to set pressure test or
replace all Main Steam Safety Valves (MSSVs) to a maximum interval of
60 months as currently allowed by the American Society of Mechanical
Engineers
[[Page 45532]]
(ASME) Boiler and Pressure Vessel Code (Code).
Date of issuance: August 7, 1998.
Effective date: Immediately, to be implemented within 30 days.
Amendment No.: 163.
Facility Operating License No. DPR-25: The amendment revised the
TSs.
Date of initial notice in Federal Register: June 3, 1998 (63 FR
30263).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Morris Area Public Library
District, 604 Liberty Street, Morris, Illinois 60450.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: November 2, 1994, as
supplemented January 4, 1995, February 19, 1998, April 28, 1998, and
June 5, 1998.
Brief description of amendment: The amendment revises the Technical
Specifications that have become unnecessary due to previous approved
amendments, make editorial changes, change managerial titles, update
references and reporting requirements.
Date of issuance: August 12, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 198.
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 8, 1995 (60 FR
56365).
The January 4, 1995, February 19, 1998, April 28, 1998, and June 5,
1998, 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 August 12, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: January 28, 1998 (NRC-98-0011)
as supplemented March 12 and June 9, 1998.
Brief description of amendment: The amendment revises Technical
Specification 3.4.2.1, ``Safety/Relief Valves,'' changing the safety
relief valve (SRV) setpoint tolerance from plus or minus 1 percent to
plus or minus 3 percent. An associated footnote is revised to indicate
that, although the as-found setpoint tolerance is now plus or minus 3
percent, the as-left settings of the SRVs shall be within plus or minus
1 percent of the specified setpoints prior to installation of the SRVs
after testing. Bases section 3/4.4.2 is also revised.
Date of issuance: July 31, 1998.
Effective date: July 31, 1998, with full implementation prior to
restart from the sixth refueling outage.
Amendment No.: 123.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9600). The March 12 and June 9, 1998, letters provided clarifying
information that was within the scope of the original Federal Register
notice and did not change the staff's initial proposed no significant
hazards considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 31, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: June 26, 1998 (NRC-98-0040) as
supplemented July 16, 1998 (NRC-98-0096), and July 23, 1998 (NRC-98-
0117).
Brief description of amendment: The amendment provides a one-time
extension of the interval for a number of technical specification (TS)
surveillance requirements that will be performed during the sixth
refueling outage. TS 4.0.2 and Index page xxii are revised and TS
tables 4.0.2-1 and 4.0.2-2 are replaced to reflect the extensions.
NRC has also granted the request of Detroit Edison Company to
withdraw a portion of its June 26, 1998, application. By letter dated
July 16, 1998, the licensee made some editorial changes and withdrew
the portion of the submittal related to TS 4.0.5 for the inservice
testing of two valves. A change to the schedule for these valves will
be handled within the Inservice Testing Program and a TS change is not
necessary. For further details with respect to this action, see the
application for amendment dated June 26, 1998, and the licensee's
letter dated July 16, 1998, which withdrew this portion of the
application for the license amendment, and the staff's safety
evaluation enclosed with the amendment. By letter dated July 23, 1998,
the licensee added an additional surveillance requirement for two
instruments to the amendment. The above documents 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 listed below.
Date of issuance: August 4, 1998.
Effective date: August 4, 1998, with full implementation within 30
days.
Amendment No.: 124.
Facility Operating License No. NPF-43: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: July 2, 1998 (63 FR
36273). The July 16 and July 23, 1998, letters provided clarifying
information and updated TS pages that were within the scope of the
original Federal Register notice and did not change the staff's initial
proposed no significant hazards considerations determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Monroe County Library System,
Ellis Reference and Information Center, 3700 South Custer Road, Monroe,
Michigan 48161.
Duke Energy Corporation, 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: July 8, 1998.
Brief description of amendments: The amendments revise TS
4.5.4.1.b.1 for testing the Penetration Room Ventilation System air
flow by adding a reference to the following statement that has been
added to the bottom of the TS page: ``A temporary noncompliance with
this surveillance requirement is allowed until August 30, 1998, to
complete necessary modifications to enable flow testing in accordance
with ANSI N510-1975.'' This action supersedes the Notice of Enforcement
Discretion that was issued by the staff on July 8, 1998.
Date of Issuance: August 7, 1998.
[[Page 45533]]
Effective date: As of the date of issuance.
Amendment Nos.: Unit 1--231; Unit 2--231; Unit 3--228.
Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The
amendments revise the Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes. (63 FR 38433 dated July 16, 1998). 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 August 17, 1998, but indicated that if the Commission makes
a final no significant hazards consideration determination, any such
hearing would take place after issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and a final no significant hazards consideration
determination are contained in a Safety Evaluation dated August 7,
1998.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of application for amendment: May 19, 1995, as supplemented by
letters dated February 27 and September 30, 1996.
Brief description of amendment: The amendment modifies the
technical specifications (TSs) to extend the allowed outage times
(AOTs) for a single inoperable Safety Injection Tank (SIT) from one
hour to 24 hours, and for a single inoperable SIT specifically due to
malfunctioning SIT water level or nitrogen cover pressure
instrumentation inoperability from one hour to 72 hours.
Date of issuance: August 7, 1998.
Effective date: August 7, 1998.
Amendment No.: 192.
Facility Operating License No. NPF-6: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 2, 1995 (60 FR
39439).
The February 27 and September 30, 1996, submittals provided
clarifying information that did not change the initial proposed NSHC
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, AR 72801.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida
Date of application for amendment: June 18, 1998, and supplemented
June 30, 1998.
Brief description of amendment: The amendment proposed to revise
the Improved Technical Specifications to allow operation with a number
of indications previously identified as tube end anolmalies and
multiple tube end anolmalies in the Crystal River Unit 3 Once Through
Steam Generator tubes.
Date of issuance: July 30, 1998.
Effective date: July 30, 1998.
Amendment No.: 169.
Facility Operating License No. DPR-72: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 30, 1998 (63 FR
35615). The June 30, 1998 supplement included clarifying information
which did not change the original no significant hazards determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Coastal Region Library, 8619
W. Crystal Street, Crystal River, Florida 34428.
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389,
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida
Date of application for amendments: May 27, 1998.
Brief description of amendments: The amendments revise the
Administrative Controls, Unit Staff Section 6.2.2.f of TS to authorize
the use of various controlled shift structures and durations during a
nominal (36 to 48 hours) work week. This includes the use of up to 12-
hour shifts without heavy use of overtime.
Date of Issuance: July 30, 1998.
Effective Date: July 30, 1998.
Amendment Nos.: 155 and 93.
Facility Operating License Nos. DPR-67 and NPF-16: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 1, 1998 (63 FR
35989).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 30, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
GPU Nuclear, Inc. et al., Docket No. 50-219, Oyster Creek Nuclear
Generating Station, Ocean County, New Jersey
Date of application for amendment: October 31, 1996.
Brief description of amendment: The amendment deletes Table 3.5.2
which lists automatic primary containment isolation valves. In
addition, the amendment clarifies the applicability of an action
statement that applies to several limiting conditions for operation in
Section 3.5, and deletes closure time requirements for several
automatic isolation valves in Section 4.5.F.
Date of Issuance: August 13, 1998.
Effective date: August 13, 1998, to be implemented within 60 days.
Amendment No.: 196.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 18, 1996 (61
FR 66707).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated August 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit
1, DeWitt County, Illinois
Date of application for amendment: October 22, 1995.
Brief description of amendment: The amendment changes Technical
Specification 5.2.2.e, ``Unit Staff,'' by revising the requirements for
controls on the working hours of unit staff who perform safety related
functions.
Date of issuance: August 13, 1998.
Effective date: August 13, 1998.
Amendment No.: 115.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 20, 1995 (60
FR 65681).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Vespasian Warner Public
[[Page 45534]]
Library, 120 West Johnson Street, Clinton, IL 61727.
Northeast Nuclear Energy Company, et al., Docket No. 50-423, Millstone
Nuclear Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: January 22, 1998, as
supplemented July 17, 1998.
Brief description of amendment: The amendment revises the Millstone
Unit 3 licensing basis to accept the existing use of epoxy coatings on
safety related components. The revised licensing basis will be
incorporated into Chapter 9 of the Final Safety Analysis Report.
Date of issuance: August 7, 1998.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 162.
Facility Operating License No. NPF-49: Amendment revised the Final
Safety Analysis Report and the Facility Operating License.
Date of initial notice in Federal Register: February 25, 1998 (63
FR 9606).
The July 17, 1998, letter provided clarifying information that did
not change the scope of the January 22, 1998, 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 August 7, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano,
49 Rope Ferry Road, Waterford, Connecticut.
Northern States Power Company, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: September 24, 1996, as
supplemented October 17, 1996, January 3, January 20, and November 10,
1997, and January 9, June 8, and July 20, 1998.
Brief description of amendments: The amendments revise the
Technical Specifications (TSs) for the Prairie Island Nuclear
Generating Plant Units 1 and 2 to allow use of an alternate steam
generator tube repair criteria (elevated F-star or EF*) in the
tubesheet region when used with the repair method of additional roll
expansion. The amendments incorporate revised acceptance criteria for
tubes with degradation in the tubesheet region and enable the licensee
to avoid unnecessary plugging and sleeving of steam generator tubes.
Date of issuance: August 13, 1998.
Effective date: August 13, 1998, with full implementation within 30
days.
Amendment Nos.: 137 and 128.
Facility Operating License Nos. DPR-42 and DPR-60. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 4, 1996 (61 FR
64388).
The licensee's submittals dated January 3, January 20, and November
10, 1997, and January 9, June 8, and July 20, 1998, provided additional
clarifying information within the scope of the original Federal
Register notice and did not affect the staff's initial no significant
hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 13, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: February 6, 1998.
Brief description of amendment: The amendment revises the Reactor
Protection System Normal Supply Electrical Protection Assembly
Undervoltage Trip Setpoint.
Date of issuance: July 29, 1998.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 245.
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 22, 1998 (63 FR
19976).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 29, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
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: November 14, 1997.
Brief description of amendments: The amendments add technical
specification (TS) surveillance requirements for the service water
accumulator vessels. Specifically, surveillance requirements are
provided for vessel level, pressure and temperature, and discharge
valve response time. The surveillance requirements are included in TS
3/4.6.1.1 and 3/4.6.2.3, and the applicable Bases sections are expanded
to provide supporting information.
Date of issuance: August 6, 1998.
Effective date: August 6, 1998.
Amendment Nos.: 213 and 193.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 28, 1998 (63 FR
4432).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 6, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, NJ 08079.
Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: May 6, 1998.
Brief description of amendment: The requested changes would replace
the two percent penalty addressed in Surveillance Requirement
3.2.1.2(a) with a burnup-dependent factor to be specified in the WBN
Core Operating Limits Report and makes associated changes to the
administrative controls in Technical Specification 5.9.5 and the BASES.
Date of issuance: August 10, 1998.
Effective date: August 10, 1998.
Amendment No.: 11.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: June 17, 1998 (63 FR
33109).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 10, 1998.
No significant hazards consideration comments received: None
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402.
[[Page 45535]]
Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 50-
339, North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of application for amendments: November 5, 1997.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) Sections 3.9.7, 4.9.7.1, 4.9.7.2, and 3/
4.9.7 for Unit 1, and 3.9.7, 4.9.7.1, 4.9.7.2, and 3/4 .9.7 for Unit 2,
allowing the movement of the spent fuel pit gate over the irradiated
fuel.
Date of issuance: August 3, 1998.
Effective date: August 3, 1998.
Amendment Nos.: 213 and 194.
Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised
the Technical Specifications.
Date of initial notice in Federal Register: December 17, 1997 (62
FR 66146).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 3, 1998.
No significant hazards consideration comments received: No.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Dated at Rockville, Maryland, this 19th day of August 1998.
For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of
Nuclear Reactor Regulation.
[FR Doc. 98-22766 Filed 8-25-98; 8:45 am]
BILLING CODE 7590-01-P