[Federal Register Volume 60, Number 2 (Wednesday, January 4, 1995)]
[Notices]
[Pages 493-513]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-100104]
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[[Page 494]]
NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from December 12, 1994, through December 21,
1994. The last biweekly notice was published on December 21, 1994 (59
FR 65809).
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 Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, 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 February 3, 1994, 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 [[Page 495]] contention will
not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, 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.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: November 30, 1994
Description of amendment requests: The proposed amendment would
relocate Table 3.3-2, ``Reactor Protective Instrumentation Response
Times,'' and Table 3.3-5, ``Engineered Safety Features Response
Times,'' of Technical Specifications (TS) 3/4.3.1 and 3/4.3.2,
respectively, to the Palo Verde Updated Final Safety Analysis Report
(UFSAR) in accordance with the guidance provided in Generic Letter (GL)
93-08. In addition, the proposed amendment would make administrative
changes to two previous TS amendment requests to reflect the deletion
of Tables 3.3-2 and 3.3-5. The amendment would also delete an obsolete
footnote on page 3/4 3-17 of the Palo Verde Unit 2's TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1 -- Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed change relocates two tables of instrument response
time limits from the TS to the UFSAR. The changes are in accordance
with the guidance provided by the NRC in Generic Letter 93-08. The
changes are administrative in nature and do not involve any
modifications to plant equipment or affect plant operation.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Standard 2 - Does the proposed change create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed change relocates two tables of instrument response
time limits from the TS to the UFSAR. The changes are in accordance
with the guidance provided by the NRC in Generic Letter 93-08. The
changes are administrative in nature, do not involve any
modifications to plant equipment and cause no change in the method
by which any safety-related system performs its function. Therefore,
the proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Standard 3 - Does the proposed change involve a significant
reduction in a margin of safety?
The proposed change relocates two tables of instrument response
time limits from the TS to the UFSAR. The changes are in accordance
with the guidance provided by the NRC in Generic Letter 93-08. The
changes are administrative in nature, do not change or alter
regulatory requirements and do not affect the safety analysis. Plant
procedures contain response time testing acceptance criteria that
reflect the reactor trip and ESFAS [engineered safety feature
actuation system] response time limits in the tables being relocated
from the TS into the UFSAR. Therefore, the proposed change does not
involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: December 7, 1994
Description of amendment requests: The proposed amendment would
change Table 4.3-1 of Technical Specification 3/4.3.1 to allow
verification of the shape annealing matrix elements used in the Core
Protection Calculators. This would provide the option to use generic
shape annealing matrix elements in the Core Protection Calculators.
Presently, cycle-specific shape annealing elements are determined
during startup testing after each core reload. Use of a generic shape
annealing matrix would eliminate approximately 2 to 3 hours of critical
path work during startup after a refueling outage.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis [[Page 496]] about the issue of no significant
hazards consideration, which is presented below:
Standard 1 -- Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
The proposed Technical Specification amendment does not involve
a significant increase in the probability or consequences of an
accident previously evaluated. The Technical Specification amendment
provides the option to use generic shape annealing matrix elements
in the Core Protection Calculators. The design basis of the Core
Protection Calculators is to provide the DNBR [departure from
nucleate boiling ratio] and linear heat rate trip functions for the
Reactor Protection System so that the Specified Acceptable Fuel
Design Limits on DNBR and fuel centerline melt are not exceeded
during normal operation or Anticipated Operational Occurrences, and
assist the Engineered Safety Features Actuation System in limiting
the consequences of postulated accidents. The generic shape
annealing matrix elements will be validated during startup testing
and will meet the same acceptance criteria as the cycle specific
shape annealing matrix elements. If the generic shape annealing
matrix elements are not valid, cycle specific shape annealing matrix
elements would be used in the Core Protection Calculators. This
change will not affect the Core Protection Calculators capability to
protect the plant by tripping the reactor, based on a conservative
calculation of minimum DNBR and peak linear heat rate, to ensure
that the Specified Acceptable Fuel Design Limits are not violated in
the event of an Anticipated Operational Occurrence. Therefore, the
generic shape annealing matrix elements will not affect the safety
analysis, since there is no change to the design basis of the Core
Protection Calculator System.
Standard 2 -- Does the proposed change create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed Technical Specification amendment does not create
the possibility of a new or different kind of accident from any
accident previously evaluated. Since the generic shape annealing
matrix elements will still have to meet the same acceptance criteria
as the cycle specific shape annealing matrix elements, the Core
Protection Calculators will still generate axial power shapes that
fall within the required uncertainties. The Core Protection
Calculators will still trip the reactor, based on a conservative
calculation of minimum DNBR and peak linear heat rate, to ensure
that the Specified Acceptable Fuel Design Limits are not violated in
the event of an Anticipated Operational Occurrence.
Standard 3 -- Does the proposed change involve a significant
reduction in a margin of safety?
The proposed Technical Specification amendment will not involve
a significant reduction in a margin of safety. There is no reduction
in the margin of safety, since the generic shape annealing matrix
elements will still have to meet the same acceptance criteria as the
cycle specific shape annealing matrix elements. Therefore, this
change will not affect the design basis of the Core Protection
Calculators. The Core Protection Calculators will still provide a
reactor trip based on a conservative calculation of minimum DNBR and
peak linear heat rate.
The NRC staff has reviewed the licensees' analysis and, based on
that 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: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit No. 2, Maricopa County,
Arizona
Date of amendment request: November 30, 1994
Description of amendment request: The proposed amendment would
change the pressurizer code safety valve lift setting from 2500 psia to
2475 psia. The lift setting is being changed to permit Unit 2 to
operate with up to 1500 plugged tubes in each steam generator.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1 -- Does the proposed change involve a significant
increase in the probability or consequence of an accident previously
evaluated?
The proposed Technical Specification amendment does not involve
a significant increase in the probability or consequences of an
accident previously evaluated. Chapters 6 and 15 of the [Palo Verde
Nuclear Generating Station] PVNGS [Updated Final Safety Analysis
Report] UFSAR have been reviewed to address the impact of these
changes (1500 plugged tubes and a pressurizer code safety valve
nominal lift setpoint of 2475 psia) on accident consequences. For
most of the events that were previously analyzed in the UFSAR, the
proposed change does not have a significant affect or adversely
impact the accident analysis. For RCS [reactor coolant system]
pressure peaking events, Loss of Condenser Vacuum (LOCV) and
Feedwater Line Breaks (FLB), a new analysis was performed to justify
the acceptability of the changes.
For the LOCV event (anticipated operational occurrence), the
reanalysis determined that the peak RCS pressure, assuming 1500
plugged tubes and a pressurizer code safety valve nominal lift
setpoint of 2475 psia, is 2728 psia. The maximum reactor coolant
system (RCS) pressure reached for this event as described in UFSAR
Section 15.2.3 is 2742 psia. Therefore, this change is bounded by
the reference cycle (UFSAR analysis) and remains below the 110%
(2750 psia) design pressure limit.
Several FLB scenarios are analyzed in support of PVNGS Unit 2
operation. The scenario with the highest system pressures is the
large FLB with a loss of alternating current (LOAC). For the large
FLB with a LOAC event (limiting fault event), assuming 1500 plugged
tubes and a pressurizer code safety valve nominal lift setpoint of
2475 psia, is 2813 psia. The maximum RCS pressure reached for this
event as described in UFSAR Section 15.2.8 is 2843 psia. The
analysis shows that the RCS peak pressure for the large FLB with a
LOAC (very low probability) event remains below the required value
of 120% (3000 psia) of design pressure. Therefore, the analyses and
reviews of the RCS pressure peaking events determined that the UFSAR
design pressure limit is still bounding with this change. That is,
the RCS design pressure limit will not be exceeded. Also, safety
valves are accident mitigating devices and do not contribute to the
probability of an event.
Standard 2 -- Does the proposed change create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The proposed Technical Specification amendment does not create
the possibility of a new or different kind of accident from any
accident previously evaluated. The analyses and reviews show that
the current licensing basis remains valid for this change (UFSAR
design pressure limit is still bounding with this change). Safety
valves are accident mitigating devices and do not contribute to the
possibility of an accident. The pressurizer code safety valves are
not manually or remotely operated, but are designed to automatically
open to provide overpressure protection for pressure peaking events.
The change in the pressurizer code safety valve setpoint to 2475
psia does not significantly increase the probability of a
pressurizer code safety valve opening, since the pressure is still
well above the Technical Specification Table 2.2-1 reactor trip
setpoint of 2383 psia for high pressurizer pressure.
Standard 3 -- Does the proposed change involve a significant
reduction in a margin of safety?
The proposed Technical Specification amendment does not involve
a significant reduction in a margin of safety. The analyses and
reviews show that the limits in the licensing and design basis are
still valid with this change. The analyses show that the RCS peak
pressure remains below the 110% (2750 psia) design pressure limit
for the LOCV event and remains below the required value of 120%
(3000 psia) of design pressure RCS peak pressure for the large FLB
with a LOAC (very low probability) event. The analyses
[[Page 497]] and reviews of the RCS pressure peaking events
determined that the UFSAR design pressure limit is still bounding
with this change. Therefore, the proposed Technical Specification
amendment maintains the margin of safety to the design pressure
limit.
The NRC staff has reviewed the licensees' analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County,North Carolina
Date of amendments request: November 16, 1994Description of
amendments request: The proposed revision to the Technical
Specifications (TS) would change the Technical Specification 3/4.6.2 to
remove the specific instrumentation requirements for monitoring of the
suppression chamber average water temperature. Also, the proposed
revision would change the TS Bases 3/4.6.2 to indicate the methods that
are acceptable for determining suppression chamber average water
temperature.Proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change maintains the same number of monitored
locations from which an average suppression chamber water
temperature can be derived, while making available additional valid
RTD [resistance temperature detector] inputs from what was the
redundant channel. No safety-related equipment, safety function or
plant operation will be altered as a result of the proposed change.
The SPTMS [suppression chamber temperature monitoring system] is
neither an accident initiator nor does it provide any automatic
accident mitigation function. The change does not affect the design,
materials, or construction standards applicable to the suppression
chamber average water temperature monitoring instrumentation.
2. The proposed amendment would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The fundamental function and objective of the system is not
affected by the proposed change. As stated above, no safety-related
equipment, safety function or plant operations will be altered as a
result of the proposed change. The change does not affect the
design, materials, or construction standards applicable to the
suppression chamber average water temperature instrumentation.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change allows the substitution of a qualified RTD
already installed at a monitored location to insure the suppression
chamber average water temperature remains valid. It does not involve
any changes to the plant design or operation, therefore, no margins
of safety, as defined by the plant's accident analyses, are
impacted. Deletion of the defined instrument channels will not
affect the ability to verify the suppression chamber ``average''
water temperature is being maintained below the maximum average
temperatures required by the specification. This will insure the
suppression chamber is Operable and able to perform its intended
safety function.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: December 12, 1994
Description of amendment request: The requested change would revise
the containment spray (CS) nozzle surveillance interval from 5 to 10
years.
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 requested change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The requested change extends the surveillance interval for
performance of qualitative flow testing of the CS nozzles. A
revision to this surveillance interval can in no way increase the
probability of any accident previously evaluated.
Containment spray nozzle testing is not intended to track
degradation of equipment by monitoring or trending performance.
Rather, this surveillance constitutes a test of the passive design
of the spray nozzles, i.e., it merely demonstrates whether the
nozzles are or are not blocked or clogged. Based upon industry and
plant-specific operating experience, a single failure rendering a
significant number of nozzles inoperable as a result of blockage is
considered highly unlikely. Since the reliability or functioning of
the spray nozzles will not be affected by the revised surveillance
interval, the consequences of any accident previously evaluated will
not be increased. The requested change does not affect the physical
design or operation of the plant, does not alter assumptions
contained within the Updated Final Safety Analysis Report, and will
not affect other Technical Specifications that preserve safety
analysis assumptions. Therefore, operation of the facility in
accordance with the requested change will not involve a significant
increase in the consequences of any accident previously evaluated.
2. The requested change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The requested change extends the surveillance interval for
performance of qualitative flow testing of the CS nozzles. This
change in the spray nozzle surveillance interval will not change or
affect the physical plant or the modes of plant operation defined
within the facility Operating License. This change does not involve
the addition or modification of plant equipment, nor does it alter
the design or operation of plant systems. Therefore, operation of
the facility in accordance with the requested change will not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. The requested change does not involve a significant reduction
in the margin of safety.
The requested change extends the surveillance for performance of
qualitative flow testing of the CS nozzles. This revised
surveillance interval will not change or otherwise influence the
degree of operability assumed for the CS system within the plant
safety analyses. As demonstrated by plant-specific and industry
experience, an operational failure of the containment spray nozzles
is considered highly unlikely. Since prior testing has demonstrated
proper functioning of the CS spray nozzles, and operational single-
failures are considered highly unlikely, a reduction in testing
frequency should not affect the ability of the CS system to mitigate
the affects of a large loss-of-coolant or steam release accident.
[[Page 498]] Therefore, operation of the facility in accordance with
the requested change will not result in a significant reduction in
the margin or 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: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: October 24, 1994
Description of amendment request: The proposed amendments would
restructure the primary containment integrity and primary containment
leakage technical specifications (TS) to reduce the repetition of those
requirements contained in NRC regulations such as Appendix J to 10 CFR
50. The amendments also support proposed exemptions from Appendix J
requirements related to the scheduling of containment integrated leak
rate tests (CILRT). In addition to the restructuring and scheduling
changes, the proposed amendments incorporate (1) the relocation of the
list of primary containment isolation valves in accordance with Generic
Letter 91-08, ``Removal of Component Lists from Technical
Specifications,'' and (2) a revision of the interval for functional
testing of hydrogen recombiners from 6 months to 18 months in
accordance with Generic Letter 93-05, ``Line-Item Technical
Specifications Improvements to Reduce Surveillance Requirements for
Testing During Power Operation.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated because of the
following:
a. The relocation of Technical Specification 3/4.6.1.2, Primary
Containment Leakage, and Surveillance Requirements 4.6.1.1.a,
4.6.4.3, and 4.6.6.1.d to specification 3/4.6.1.1, Primary
Containment Integrity, as Surveillance Requirement 4.6.1.1.b
continues to assure that Primary Containment leakage is maintained
within the analyzed limit assumed for accident analysis by testing
in accordance with 10 CFR part 50, Appendix J as modified by
approved exemptions.
The requirement to be less than 0.75 La for as-left Type A
test and less than 0.60 La for Type B and C tests prior to
first unit startup following testing performed in accordance with 10
CFR part 50, Appendix J, as modified by approved exemptions,
provides margin for degradation between tests and thus primary
containment integrity is maintained during the time period between
required leakage testing. The current Limiting Condition for
Operation 3.6.1.2 in conjunction with Surveillance Requirements
4.6.1.2 basically require the same leakage limits as proposed
Surveillance Requirement 4.6.1.1.b. The Limiting Condition for
Operation (LCO) is required to be less than 1.0 La and is
applicable during a fuel cycle for the Type A test. The LCO for Type
B and C combined leakage total is currently required to be less than
0.60 La. The proposed Surveillance Requirement maintains the
following:
1.The current LCO for Overall Containment leakage (as determined
by a Type A test) and for the Type B and C combined leakage during
the cycle by requiring overall containment leakage to be less than
1.0 La and Type B and C leakage total less than 0.60 La.
2. The associated limits specified in the current Action
Statements are maintained by verifying Overall Containment leakage
to be less than 0.75 La and Type B and C leakage total less
than 0.60 La prior to startup from an outage in which the
applicable leakage testing is conducted.
Therefore, there is no change to the consequences of an accident
previously evaluated, because maintaining leakage within the
analyzed limit assumed for accident analysis does not change either
the onsite or offsite dose consequences resulting from an accident.
In addition to this, containment leakage is not an accident
initiator, so there is no effect on the probability of accident
initiators. Thus there is no increase in the probability of an
accident previously analyzed.
b. Relocation of Technical Specification table of Primary
Containment Isolation Valves, Table 3.6.3-1, to the LaSalle UFSAR is
an administrative change to remove the component list of Primary
Containment Isolation Valves, Table 3.6.3-1, from the Technical
Specifications. The Limiting Condition for Operation (LCO), 3.6.3,
is being revised to define which components the LCO applies to. The
wording of the revised LCO encompasses all of the components listed
in the current Technical Specification Table 3.6.3. Removal of this
component list does not change the probability of any accident
initiators or change any other relevant initial assumptions. Also,
there is no change to the consequences of an accident previously
evaluated, because removing this list from Technical Specifications
does not change either the onsite or offsite dose consequences
resulting from the event. The component list will be controlled by
an Administrative Procedure and can only be changed by the 10 CFR
50.59 change process with review and approval per the Onsite Review
and Investigative Function. Therefore, there is no increase in
either the probability or consequences of an accident previously
evaluated.
c. The change in the functional test interval for the Drywell
and Suppression Chamber Hydrogen Recombiner systems from ``once per
6 months'' to ``once per 18 months'' was determined by the NRC in
NUREG 1366 and Generic Letter 93-05 to be acceptable by evaluation
of the industry Licensing Event Reports (LERs) to assess the
reliability of hydrogen recombiners. The conclusion was that the
interval should be changed, because of the redundancy and apparent
high reliability. A review of LaSalle LERs has shown only one LER
that involved the operability of the hydrogen recombiner system and
that was due to a Part 21 issue regarding circuit breaker
environmental qualification. The breakers were replaced with
qualified breakers. Therefore, the LaSalle Hydrogen Recombiner
reliability is consistent with or better than that found by the NRC
in determining this surveillance interval extension based on all
LERs. Also, redundancy is the same as that assumed by the NRC;
because, LaSalle has two hydrogen recombiner subsystems that are
shared by Unit 1 and Unit 2. Both hydrogen recombiners subsystems
are required to be Operable for either or both units in Operational
Conditions 1 and 2. Based on LaSalle operating experience, the
hydrogen recombiner subsystems are expected to continue to be
demonstrated operable when the functional test is performed at an 18
month frequency.
Therefore, there is minimal or no change to the consequences of
an accident previously evaluated, because at least one of the
hydrogen recombiner subsystems is expected to be available to meet
its design function to reduce the potential for hydrogen explosion
or hydrogen burn in the primary containment. By preserving the
integrity of the primary containment, there is no change to either
the onsite or offsite dose consequences resulting from an accident.
In addition to this, control of hydrogen concentration by use of a
hydrogen recombiner subsystem is not an accident initiator, so there
is no effect on the probability of accident initiators. Thus there
is no significant increase in the probability of an accident
previously analyzed.
d. The first exemption request is from the requirements of
paragraph III.A.6(b) of Appendix J to allow LaSalle County Station
Unit Two to return to or resume a Type A test schedule of three
times in ten years (40 plus or minus 10 months). Due to consecutive
failures, 10 CFR 50 Appendix J requires that Type A tests be
performed every refueling outage on Unit Two until two consecutive
Type A tests are satisfactory. 10 CFR Part 50 has an exemption
process and is specified in 10 CFR Part 50.12(a), which states:
``The Commission may, upon application by any interested person
or upon its own [[Page 499]] initiative, grant exemptions from the
requirements of the regulations of this part,...''
The exemption process requires showing that the granting of the
exemption is authorized by law, will not present an undue risk to
the public health and safety, and is consistent with the common
defense and security. Also, special circumstances are required to be
present for the granting of an exemption. One of the special
circumstances that would apply in this instance is 10 CFR part
50.12(a)(2)(ii) which states:
``Application of the regulation in the particular circumstances
would not serve the underlying purpose of the rule or is not
necessary to achieve the underlying purpose of the rule''.
This requires that it be shown that unacceptable containment
leakage will be identified and corrected, by alternative methods.
The alternative method is specifically Type B and C tests, which
will identify any local penetration leakage. This is acceptable,
because Type C test failures have been the cause for failures of as-
found Type A tests in the LaSalle Unit 2 first, third, and fourth
refueling outages.
Exceeding the allowable leakage rate during the performance of
the Type A test is indicative of either a passive or a structural
component that is leaking or that there is an inadequacy in the
Local Leak Rate Test (Type B and C tests) program. When the failure
of a Type A test is due to a passive or structural component, the
only test for adequate repair would be the Type A test. For a Local
Leak Rate Test program inadequacy, the Type A test would serve as a
means of verification of the results of the test program. The Type A
tests have not found new significant Type B or C tested local
penetration leakage that has not been identified by Type B or C
testing alone. Therefore, the LaSalle Local Leak Rate Test program
is adequate to find and correct Type B and C containment penetration
leakage.
When it is determined that Type A tests failed as a direct
result of as-found Type B and C minimum path leakage penalty
additions and not due to a non Type B or C tested components or
structures, then performance of the Type A test more frequently as
required by 10 CFR Part 50, Appendix J, due only to Type B and C
test failures is redundant to the performance of Type B and C tests.
Therefore, Type B or C tested penetration leakage that can be
determined by Type B or C tests is evaluated and corrected, as
applicable, to maintain overall containment leakage within limits,
without an additional Type A test.
Primary Containment leakage which includes the minimum path
Primary Containment Isolation Valve leakage is an assumption in any
analyzed accident which could involve an offsite radioactive
release. Because performance of Type B and C tests will find and
allow correction/repair of leaking valves/penetrations, verification
of as-found and as-left local leakage assures that Primary
Containment leakage will be within the analyzed limit assumed for
accident analysis.
Therefore, for this one-time exemption for LaSalle Unit 2, there
is little or no increase in the consequences of an accident
previously evaluated involving the dose previously calculated either
onsite or offsite at the site boundary due to any analyzed accident.
In addition to this, containment leakage is not an accident
initiator, so there is no effect on the probability of accident
initiators. Thus there is no significant increase in the probability
of an accident previously analyzed.
e. The request for a partial exemption from paragraph III.D of
Appendix J to 10 CFR 50 involves a deletion of the requirement to
perform the third Type A test for each 10-year service period during
the shutdown for the 10-year plant inservice inspections. There is
no significant benefit in coupling these two surveillances (i.e.,
the Type A test and the 10-year ISI program). Each of the two
surveillances is independent of the other and provides assurance of
different plant characteristics. The Type A test assures the
required leak-tightness for the reactor containment building be less
than Appendix J acceptance criteria. This demonstrates compliance
with the guidelines of 10 CFR Part 100 based on the assumptions used
in the UFSAR which conform to NRC Safety Guide 4. The 10-year ISI
program provides assurance of the integrity of the plant structures,
systems, and components in compliance with 10 CFR 50.55(a). There is
no safety-related concern necessitating their coupling to the same
refueling outage. As a result, this change cannot increase the
consequences (i.e., offsite dose) of any accident previously
evaluated. Furthermore, since the decoupling of the test schedules
has no affect on the test's effectiveness, decoupling their
schedules will not increase the probability of an accident.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated because:
a. Technical Specification 3/4.6.1.2, Primary Containment
Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3, and
4.6.6.1.d are being relocated to specification 3.4.6.1.1, Primary
Containment Integrity, as Surveillance Requirement 4.6.1.1.b. The
proposed Surveillance Requirement 4.6.1.1.b assures that Primary
Containment leakage is maintained within the analyzed limit assumed
for accident analysis by testing in accordance with 10 CFR part 50,
Appendix J as modified by approved exemptions. Primary containment
leakage is an assumption in accident analyses, and is maintained by
both the current specifications and the proposed specification. The
leakage does not cause an accident and no new failure modes are
created. Therefore this request for exemption does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
b. This is an administrative change to control the list of
Primary Containment Isolation Valves outside the LaSalle Unit 1 and
Unit 2 Technical Specifications. The administrative controls
provided to control this component list assure that the design and
operation of the plant will continue to be in accordance with the
UFSAR, Facility License and the associated Technical Specifications.
Therefore, the possibility of a new or different kind of accident
from any previously evaluated is not created.
c. The change in the functional test interval for the Drywell
and Suppression Chamber Hydrogen Recombiner systems from ``once per
6 months'' to ``once per 18 months'' is based on good equipment
performance on a 6 month frequency. The expected outcome of the 18
month surveillances, based on the low failure rate at a six month
frequency, is to show the hydrogen recombiner subsystems Operable.
This system is for mitigating the consequences of an accident that
causes generation of hydrogen and oxygen in the primary containment.
No new failure modes are created by this change in surveillance
frequency. Therefore, the possibility of a new or different kind of
accident from any previously evaluated is not created.
d. The first exemption is from the requirements of paragraph
III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to
return to or resume a Type A test schedule of three times in ten
years (40 plus or minus 10 months). Containment leakage testing,
including both Type B and C testing and Type A testing as specified
in the LaSalle County Station Safety Analysis Report were evaluated
in Section 6.2.6 of Safety Evaluation Report, NUREG-0519, and found
to be acceptable. Since Type B and C testing will find and verify
correction of penetration leakage when Type B and C test as-found
penalties are specifically what caused the failure of the as-found
Type A tests, then Type B and C testing will provide adequate
assurance of the continued integrity of the Primary Containment
without increasing the frequency of Type A tests. As a result, the
Primary Containment will continue [to] be maintained as designed and
previously evaluated.
Based on this, the requirement of two acceptable as-found Type A
tests prior to returning to the Appendix J paragraph III.D frequency
of three times in ten years (40 plus or minus 10 months) is not
necessary to assure that the primary containment remains within the
analyzed leakage limits. Containment leakage is an assumption for
the dose consequences of accident analyses, and not an accident
initiator. Also, no new failure modes are created by this exemption.
Therefore this Amendment does not create the possibility of a new or
different kind of accident.
e. The request for a partial exemption from paragraph III.D of
Appendix J to 10 CFR 50 involves a deletion of the requirement to
perform the third Type A test for each 10-year service period during
the shutdown for the 10-year plant inservice inspections. The
proposed exemption does not involve any change to the plant design
or operation. As discussed above, this change cannot increase the
consequences of any accident previously evaluated. As a result, no
new failure modes are created. Therefore, this proposed change
cannot create the possibility of any new or different kind of
accident from any accident previously evaluated.
3) Involve a significant reduction in the margin of safety
because:
a. Technical Specification 3/4.6.1.2, Primary Containment
Leakage, and Surveillance Requirements 4.6.1.1.a, 4.6.4.3,
[[Page 500]] and 4.6.6.1.d are being relocated to specification 3/
4.6.1.1, Primary Containment Integrity, as proposed Surveillance
Requirement 4.6.1.1.b. The proposed Surveillance Requirement
4.6.1.1.b continues to assure that Primary Containment leakage is
maintained within the analyzed limit assumed for accident analysis
by testing in accordance with 10 CFR part 50, Appendix J as modified
by approved exemptions.
As stated in 1)a. above, the proposed Surveillance Requirement
4.6.1.1.b maintains the acceptance criteria and limits for continued
operation of the current specification for primary containment
leakage. Therefore, the margin of safety is not reduced by this
change. Also, the proposed addition of a definition for the maximum
allowable primary containment leakage rate assures that the margin
of safety is maintained.
The leakage limits for MSIVs and hydrostatically tested valves
are maintained by relocating the current surveillance requirements
to specification 3/4.6.3, with the acceptance criteria of the
current specification retained. Thus preserving the current margin
of safety by maintaining the leakage rates as assumed in the
accident analyses.
b. The Limiting Condition for Operation for Technical
Specification 3.6.3, Primary Containment Isolation Valves, is
revised by this Technical Specification change to specifically
define the components to which the LCO applies. Therefore, removal
of Technical Specification Table 3.6.3-1, which lists the specific
components to which the LCO applies does not change the scope or
applicability of the specification. The component list will be
controlled administratively with any changes to the list made in
accordance with the 10 CFR 50.59 change process. Therefore, this is
an administrative change only and there is no reduction in the
margin of safety.
c. The change in the functional test interval for the Drywell
and Suppression Chamber Hydrogen Recombiner systems from ``once per
6 months'' to ``once per 18 months'' is based on good equipment
performance on a 6 month frequency. The expected outcome of the 18
month surveillances, based on the low failure rate at a six month
frequency, is to show the hydrogen recombiner subsystems Operable.
The change in frequency has no affect on the hydrogen or oxygen
generation assumptions or the recombination rate of the hydrogen
recombiner subsystems. Therefore, the margin of safety is not
reduced or changed by this surveillance interval change.
d. The first exemption is from the requirements of paragraph
III.A.6(b) of Appendix J to allow LaSalle County Station Unit Two to
return to or resume a Type A test schedule of three times in ten
years (40 plus or minus 10 months). The limit of total leakage
determined from Type B and C tests will remain the same, providing a
margin of 40 percent to the maximum allowable containment leakage
rate (La) at the design basis accident pressure specified in
proposed Technical Specification definition of La. This 40
percent is as specified by 10 CFR Part 50, Appendix J. In addition
to this, administrative guidelines have been set for each
penetration/valve, so that any abnormal leakage will be corrected by
adjustment or repair as needed. Any postponement of repairs is based
on a technical evaluation and then only if the total Type B and Type
C leakage is maintained at less than 0.60 La. Repairs will be
required to restore the leakage rate to less than the administrative
limit at the next refueling outage.
This request for exemption is based the fact that Type B and C
testing minimum path leakage rate penalties are the direct cause of
the failure of as-found Type A tests. The leakage through Type B and
C tested penetrations is best measured and corrected via a local
leak test. Therefore, verification of an adequate margin of safety
is assured by conducting Type B and C tests, and not another
increased frequency Type A test.
e. The request for a partial exemption from paragraph III.D of
Appendix J to 10 CFR 50 involves a deletion of the requirement to
perform the third Type A test for each 10-year service period during
the shutdown for the 10-year plant inservice inspections. The
proposed exemption does not change the acceptance criteria that must
be met for inservice inspections, does not relax the condition of
containment that must be met prior to plant restart, and does not
change the requirements that must be met between plant refueling
outages. Therefore, the proposed change does not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Local Public Document Room location: Public Library of Illinois
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of amendment request: November 21, 1994
Description of amendment request: The proposed amendment would
revise the Technical Specifications to allow a one-time extension of
the allowed outage time for an inoperable reserve source of offsite
power.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability of occurrence or consequences of an
accident previously evaluated.
The proposed changes will extend the allowed outage time for the
Reserve source of off-site power, on a one time basis, to allow the
installation of high speed protective relays on the unit system
auxiliary transformers which will increase the level of protection
from ground faults on the low voltage (secondary) side of the
transformers. Operation of Zion, Units 1 and 2, in accordance with
the proposed requirements will not affect the initiators or
precursors of any accident previously evaluated. Operation in
accordance with the proposed requirements will not increase the
likelihood that a transient initiating event will occur because
transients are initiated by equipment malfunction and/or
catastrophic system failure. As a result, the probability of
occurrence of accidents previously evaluated is not significantly
increased.
During the [system auxiliary transformer] SAT outage, power to
the shut down unit will be provided by backfeeding off-site power
through the unit main power transformers and the UAT to supply the
unit non-essential 4-KV service buses. Emergency on-site power will
be available to the shut down unit from at least one unit specific
[emergency diesel generator] EDG when fuel is in the reactor core.
This will ensure that at least one train of Residual Heat Removal
(RHR) will have an emergency source of AC power at all times. RHR
Train A is powered by ESF bus 149(249) which can be energized by the
1B(2B) EDG during a loss of off-site power. RHR Train B is powered
by bus 148(248) which can be energized by the 1A(2A) EDG. Because
the 'O' EDG must be operable for the operating unit, it will also be
available to energize the Division 7 ESF bus on the shut down unit.
The 'O' EDG can supply buses 147 and 247 simultaneously if the need
should arise during an emergency.
Power to the operating unit (opposite unit) will be provided by
the SAT and the UAT in the normal at-power configuration. Emergency
on-site will be provided by the two unit specific EDGs (A and B) and
the common 'O' EDG. In accordance with the proposed requirements,
the Reserve source of off-site power will not be removed from
service unless all three EDGs are operable and the normal source of
off-site power is operable. Administrative controls will be in place
to limit activities in the switchyard that could impact the
reliability of the remaining source of off-site power to the unit.
The Zion PRA was used to compare the impact of extending the
action time versus the impact of manual reactor shutdown on core
damage probability. The PRA result concluded that the risk of
continuing to operate the operating unit for an additional 11 days
with the shutdown unit's SAT out of service is not significantly
greater than the risk of manually shutting down the operating unit
at the expiration of the current 72 hour action statement and is not
significant when compared to the total core damage probability in a
year.
The revised surveillance requirements will provide additional
assurance that redundant sources of power are maintained operable
while the reserve source of off-site power is
[[Page 501]] unavailable. The ability to safely shut down the
operating unit and mitigate the consequences of all accidents
previously evaluated will be maintained. The reserve source of off-
site power is not relied upon in any design basis accident.
Therefore, based on the previous discussion, the proposed changes do
not involve a significant increase in consequences of any accident
previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any previously analyzed.
The proposed changes to the Technical Specifications do not
involve the addition of any new or different types of safety-related
equipment, nor does it involve the operation of equipment required
for safe operation of the facility in a manner different from those
addressed in the safety analysis. No safety related equipment or
function will be altered as a result of the proposed changes. Also,
the procedures governing normal plant operation and recovery from an
accident are not changed by the proposed Technical Specification
changes. The proposed changes will extend the allowed outage time
for the Reserve source of off-site power, on a one-time basis, to
allow the installation of high speed protective relays on the unit
system auxiliary transformers which will increase the level of
protection from ground faults on the low voltage (secondary) side of
the transformers. The addition of the high speed relaying has been
evaluated pursuant to 10 CFR 50.59, and no unreviewed safety
questions were identified.
Requirements will be modified to require additional assurance
that the remaining off-site source of AC power and the on-site
source of emergency (emergency diesel generators) are OPERABLE.
Since no new failure modes or mechanisms are added by the proposed
changes, the possibility of a new or different kind of accident is
not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed changes will extend the allowed outage time for the
reserve source of off-site power, on a one-time basis, to allow for
installation of high speed protective relays on the unit system
auxiliary transformers which will increase the level of protection
from ground faults on the low voltage (secondary) side of the
transformers.
During the SAT outage, power to the operating unit (opposite
unit) will be provided by the unit SAT and the UAT in the normal
configuration. Emergency on-site power will be provided by the two
unit specific EDGs (A and B) and the common 'O' diesel generator.
Because the accident analyses take no credit for offsite power
availability, this temporary degradation will not impact the
analysis results.
No safety system setpoints are changed by this proposal. There
is no impact on any physical design margins, and no analytical
results are affected by this change. The revised surveillance
requirements will provide additional assurance that redundant
sources of power are maintained operable while the Reserve source of
off-site power is unavailable.
Based on the above discussion, the ability to safely shut down
the operating unit and mitigate the consequences of all accidents
previously evaluated will be maintained. Therefore, the margin of
safety is not significantly affected.
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: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085
Attorney for licensee: Michael I. Miller, Esquire; Sidley and
Austin, One First National Plaza, Chicago, Illinois 60690
NRC Project Director: Robert A. Capra
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of amendment request: October 5, 1994
Description of amendment request: The proposed amendment would (1)
revise primary coolant system (PCS) pressure-temperature (P-T) limits,
power-operated relief valve (PORV) setting limits, and primary coolant
pump starting limits to accommodate reactor vessel fluence for an
additional 4 effective full power years (up to 2.192 x 1019nvt).
The existing P-T limit curves are calculated for a fluence of 1.8 x
1019 could be reached as early as March 1, 1995; (2) require the
high pressure safety injection (HPSI) pumps to be ``rendered incapable
of injection into the PCS'' when the PCS is below 300 deg.F, rather
than the existing requirement to render both HPSI pumps ``inoperable''
when the PCS is below 260 deg.F. This change supports the assumption in
the P-T limit analyses that HPSI injection would not occur below
300 deg.F; and (3) establish a more restrictive limit on pressurizer
heatup rate to achieve consistency between design assumptions and
technical specification (TS) limits. The limit in the existing TS is
less restrictive than used in design calculations. Neither the design
heatup rate nor the TS heatup rate limit is achievable with installed
equipment.
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 following evaluation supports the finding that operation of
the facility in accordance with the proposed Technical
Specifications would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The revision of the Primary Coolant Pump [PCP] starting limits, PCS
P-T curves, and PORV setting limits would not cause any changes to the
capability or operation of plant systems that would affect the
probability of occurrence or consequences of an accident. These
revisions simply update the existing requirements to account for
additional reactor vessel fluence.
The reduction of the allowable pressurizer heatup rate would
have no effect on operation of the plant. The current limit is
physically unobtainable with installed equipment. The proposed
change better aligns the Technical Specification limits with the
design analysis. The change in the pressurizer heatup rate limit
will not increase the probability or consequences of an accident.
Requiring the HPSI pumps to be operable when above 325 deg.F,
rather than when above 300 deg.F does not affect the probability or
consequences of any accident previously evaluated. Neither the
existing 300 deg.F requirement nor the proposed 325 deg.F
requirement has an analytical base. This requirement was recently
changed from 325 deg.F to 300 deg.F simply for uniformity. With the
revised P-T limit analysis requirement to assure that inadvertent
HPSI injection will not occur below 300 deg.F, it is necessary to
revert to the former limit of 325 deg.F to provide time to
transition between these two contrasting HPSI pump requirements.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
The revised specifications, PCP starting limits, PCS P-T limits,
pressurizer heatup rate, PORV setting limits, and HPSI pump
restrictions, all are directly related to, and intended to prevent,
a previously analyzed event, failure of the Reactor Coolant Pressure
Boundary. Revision of these limits would not create the possibility
of a new or different kind of accident.
3. Involve a significant reduction in a margin of safety.
The revised PCP starting limits, PCS P-T limits, and PORV
setting limits are calculated using a similar methodology as the
limits which they replace. Therefore they provide the same margin of
safety.
The revised pressurizer heatup rate reduces the currently
allowable limit which is in the direction of increased margin of
safety. Since there is no equipment installed which would cause
either the existing or the proposed limit to be reached, there will
be no change on the operation of the plant equipment. Therefore
reducing the limit on the pressurizer heatup rate will not involve a
significant reduction in the margin of safety.
Requiring the HPSI pumps to be operable when above 325 deg.F,
rather than when above 300 deg.F does not involve a significant
reduction in any margin of safety. Neither the existing 300 deg.F
requirement nor the proposed [[Page 502]] 325 deg.F requirement has
an analytical base. This requirement was recently changed from
325 deg.F to 300 deg.F simply for uniformity. With the revised P-T
limit analysis requirement to assure that inadvertent HPSI injection
will not occur below 300 deg.F, it is necessary to revert to the
former limit of 325 deg.F to provide time to transition between
these two contrasting HPSI pump requirements.
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: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
NRC Project Director: John N. Hannon
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear
One,Unit No. 1, Pope County, Arkansas
Date of amendment request: November 8, 1994
Description of amendment request: The proposed amendment revises
technical specifications (TSs) associated with requirements for
performing the containment integrated leak rate test (ILRT). The
proposed change describes the ILRT test frequency by referencing the
test frequency requirements included in 10 CFR Part 50, Appendix J. The
existing specifications paraphrase the Appendix J requirements, but
include differences that result in interpretation problems.
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:
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change revises Technical Specification 4.4.1.1.4 to
reference the testing frequency requirements of 10 CFR 50, Appendix
J, and to state that NRC approved exemptions to the applicable
regulatory requirements are permitted. The current requirements of
TS 4.4.1.1.4 paraphrase the requirements of Section III.D.1.(a) of
Appendix J. The proposed administrative revision simply deletes the
paraphrased language and directly references Appendix J. No new
requirements are added, nor are any existing requirements deleted.
An approved exemption to Section III.D.1.(a) of Appendix J would not
necessarily affect the requirements of TS 4.4.1.1.4, unless the
proposed clarification phrase permitting the use of approved
exemptions is added. Any specific changes to the requirements of
Section III.D.1(a) will require a submittal from Entergy Operations
under 10CFR50.12 and subsequent review and approval by the NRC prior
to implementation. The proposed change is stated generically to
avoid the need for further TS changes if different exemptions are
approved in the future.
The proposed change, in itself, does not affect reactor
operations or accident analysis and has no radiological
consequences. The change provides clarification so that TS changes
will not be necessary in the future to correspond to applicable NRC
approved exemptions from the requirements of Appendix J. Therefore,
this change does not involve a significant increase in the
probability or consequences of any accident previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
different Kind of Accident from any Previously Evaluated.
The proposed change provides clarification to a specification
which paraphrases a codified requirement. Since the proposed
amendment would not change the design, configuration or method of
operation of the plant, it would not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed change is administrative and clarifies the
relationship between the requirements of TS 4.4.1.1.4, Appendix J,
and any approved exemptions to Appendix J. It does not, in itself,
change a safety limit, an LCO, or a surveillance requirement on
equipment required to operate the plant. The NRC will directly
approve change proposed exemption to III.D.1.(a) of Appendix J prior
to implementation. Therefore, this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
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 amendment request: December 2, 1994
Description of amendment request: The proposed amendments would
replace Appendix B, ``Environmental Technical Specifications'' with an
Environmental Protection Plan (Nonradiological) and revise the
Operating Licenses to reflect these changes. The proposed changes are
administrative in nature, altering only the format and location of
programmatic controls and procedural details relative to
nonradiological environmental monitoring.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1) The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes to the Environmental Technical
Specifications (ETS) are administrative in nature, altering only the
format and location of programmatic controls and procedural details
relative to nonradiological environmental values. The proposed
Environmental Protection Plan (EPP) (Nonradiological) contains the
programmatic controls now residing in the ETS, with appropriate
plant procedures serving as implementing documents. The proposed
changes to the operating licenses are also administrative in nature
and change the Appendix B reference from ETS to EPP. Compliance with
applicable regulatory requirements will be maintained. In addition,
the proposed changes do not alter the conditions or assumptions in
any of the accident analyses. Therefore, these proposed changes do
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2) The proposed amendments do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes to the ETS do not involve any change to the
configuration or method of operation of any plant equipment. These
proposed changes are administrative in nature and consist of
replacing the ETS with an EPP. The proposed changes to the operating
licenses are also administrative in nature and change the Appendix B
reference from ETS to EPP. Accordingly, no new failure modes have
been identified for any plant system or component important to
safety nor has any new limiting single failure been identified as a
result of the proposed changes. Therefore, the proposed changes do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3) The proposed amendments do not result in a significant
reduction in the margin of safety.
The proposed changes to the ETS relate primarily to matters
involving recordkeeping, reporting, and administrative procedures or
requirements. No significant change in the type or quantity of any
effluent release will result from this action. These changes replace
[[Page 503]] the ETS with an EPP. The proposed EPP contains the
programmatic controls now residing in the ETS, with appropriate
plant procedures serving as implementing documents to ensure
compliance with applicable regulatory requirements. The proposed
changes to the operating licenses are also administrative in nature
and change the Appendix B reference from ETS to EPP. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: November 21, 1994
Description of amendment request: The proposed amendment would
eliminate the Main Steam Isolation Valve (MSIV) - Leakage Control
System (LCS) including the primary containment isolation valves
associated with the MSIV - LCS, along with increasing the allowable
MSIV leakage rates.
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:
I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to TS Section 3.6.1.2 do not involve a
change to structures, components, or systems that would affect the
probability of an accident previously evaluated. The TS limits for
MSIVs are increased from 46 scf per hour for all four main steam
lines to less than or equal to 100 scf per hour for any one MSIV and
a combined maximum pathway leakage rate of less than or equal to 300
scf per hour for all four main steam lines. The consequences of an
accident are affected as discussed in this section.
The proposed changes to TS Section 3.6.1.4 eliminate the Main
Steam Isolation Valves (MSIVs) Leakage Control System (LCS)
requirements from the TS. As described in Section 6.7 of the FSAR,
the LCS is manually initiated in about 20 minutes following a design
basis Loss of Coolant Accident (LOCA). Since the LCS is operated
only after an accident has occurred, these proposed changes have no
effect on the probability of an accident.
Since MSIV leakage and operation of the LCS are included in the
radiological analysis for the design basis LOCA as described in
Section 15.6.5 of the FSAR, the proposed changes do not affect the
precursors of other analyzed accidents. Analysis of the effects of
the proposed changes do, however, result in acceptable radiological
consequences for the design basis LOCA previously evaluated in
Section 15.6.5 of the FSAR.
SSES, Units 1 and 2 have an inherent MSIV leakage treatment
capability as discussed below. We propose to use the drain lines
associated with the main steam lines and main turbine condenser as
an alternative to the guidance in Regulatory Guide 1.96, ``Design of
Main Steam Isolation Valve Leakage Control System For Boiling Water
Nuclear Power Plants'', Revision 0, May 1975, for MSIV leakage
treatment. If approved, we will incorporate this alternate method in
the appropriate operational procedures and Emergency Operating
Procedures.
The Boiling Water Reactor Owners' Group (BWROG) has evaluated
the availability of main steam system piping and main condenser
alternate pathways for processing MSIV leakage, and has determined
that the probability of a near coincident LOCA and a seismic event
is much smaller than for other plant safety risks. Accordingly, this
alternate MSIV leakage treatment pathway is available during and
after a LOCA. Nevertheless, the BWROG has also determined that main
steam piping and main condenser design are extremely rugged, and the
design requirements applied to SSES Unit 1 and Unit 2 main steam
system piping and main condenser contain substantial margin, based
on the original design requirements. Therefore, the alternate
treatment method has been evaluated for its capability to mitigate
the consequences of a LOCA, and has been evaluated to assure its
availability considering a seismic event.
In order to determine the capability of the main steam piping
and main condenser alternate treatment pathway, the BWROG has
reviewed earthquake experience data on the performance of non-
seismically designed piping and condensers during past earthquakes.
The data is summarized in General Electric (GE) Report, ``BWROG
Report for Increasing MSIV Leakage Rate Limits and Elimination of
Leakage Control Systems,'' NEDC 31858P, Revision 2, submitted to the
NRC by BWROG letter dated October 4, 1993. This study concluded that
the possibility of a failure that could cause a loss of steam or
condensate in Boiling Water Reactor (BWR) main steam piping or
condensers in the event of a design basis (i.e., safe shutdown)
earthquake is highly unlikely, and that such a failure would also be
contrary to a large body of historical earthquake experience data,
and thus unprecedented.
A verification has been performed of the seismic adequacy of the
Unit 1 and Unit 2 main steam piping and main condenser consistent
with the guidelines discussed in Section 6.7 of NEDC-31858P,
Revision 2, to provide reasonable assurance of the structural
integrity of these components. An evaluation, including the walkdown
report outliers, ``MSIV Leakage Alternate Treatment Method Seismic
Evaluation,'' for Unit 1 and Unit 2, is attached. The results of the
evaluation clearly demonstrate that the MSIV Leakage Alternate
Treatment Method meets the intent of 10CFR100 Appendix A, with
regards to seismic qualification. Except for the requirement to
establish a proper flow path from the MSIVs to the condenser, the
proposed method is passive and does not require any additional logic
control and interlocks. The method proposed for MSIV leakage
treatment is consistent with the philosophy of protection by
multiple barriers used in containment design for limiting fission
product release to the environment.
A plant-specific radiological analysis has been performed in
accordance with NEDC-31858P, Revision 2, to assess the effects of
the proposed increase to the allowable MSIV leakage rate in terms of
control room and off-site doses following a postulated design basis
LOCA. This analysis utilizes the hold-up volumes of the main steam
piping and condenser as an alternate method for treating the MSIV
leakage. As discussed earlier, there is reasonable assurance that
the main steam piping and condenser remain intact following a design
basis earthquake. The radiological analysis uses standard
conservative assumptions for the radiological source term consistent
with Regulatory Guide (RG) 1.3, Assumptions Used for Evaluating the
Potential Radiological Consequences of a Loss-Of-Coolant Accident
for Boiling Water Reactor, Revision 2, dated April 1974.
The analysis results demonstrate that dose contributions from
the proposed MSIV leakage rate limit of 100 scfh per steam line, not
to exceed a total of 300 scfh for all four main steam lines, and
from the proposed deletion of the LCS, result in an insignificant
increase to the LOCA doses previously evaluated against the
regulatory limits for the off-site doses and control room doses
contained in 10CFR100 and 10CFR50, Appendix A, General Design
Criterion (GDC) 19, respectively. The off-site and control room
doses resulting from a LOCA are discussed in Section 15.6.5 of the
FSAR. The off-site and control room doses resulting from a LOCA
associated with the proposed changes are the sum of LOCA doses
evaluated in the power uprate revision to the design basis DBA-LOCA
calculation (EC-RADN-1009) and the additional doses calculated using
the alternate MSIV leakage treatment method. Enclosure 3 [of
application dated November 21, 1994] summarizes the off-site and
control room doses and compares the alternate treatment method doses
to the original MSIV-LCS treatment method doses.
The 30-day whole body doses at the Low Population Zone (LPZ) did
not change and remained at .37 rem for the alternate treatment
method. The 30-day control room whole body doses increased slightly
from .38 [[Page 504]] rem to .76 rem for the alternate treatment
method. The increase in control room dose is not significant since
the revised doses are well below the regulatory limits, i.e., .76
rem calculated versus the limit of 5 rem in the control room. The
two-hour whole body dose at the Exclusion Area Boundary (EAB)
decreased slightly from 2.47 rem to 2.217 rem.
The 30-day thyroid dose at the LPZ increased from 30.4 rem for
the MSIV-LCS treatment method to 41.74 rem for the alternate
treatment method. This increase is not significant since the revised
dose of 41.74 rem is well within the regulatory limit of 300 rem.
The two-hour thyroid dose at the EAB decreased slightly from 127.8
rem to 125.61 rem. The 30-day control room thyroid dose increased
from 14.19 rem for the MSIV-LCS treatment method to 18.55 rem for
the alternate treatment method. The increased control room thyroid
dose is not significant since the revised dose remains well below
the regulatory limit of 30 rem.
The 30-day control room beta dose increased insignificantly from
12 rem for the MSIV-LCS treatment method to 12.17 rem for the
alternate treatment method, remaining a small fraction relative to
the limit of 75 rem.
In summary, the proposed changes discussed above do not result
in a significant increase in the radiological consequences of a LOCA
when the same assumptions and methods specified in the FSAR are
used, recognizing that radiological consequences calculated in the
FSAR and for these proposed changes are significantly higher than
those using more realistic assumptions and methods. Nevertheless,
the calculated off-site and control room doses resulting from a LOCA
remain well below the regulatory limits.
The proposed change to TS Table 3.6.3-1 deletes the LCS valves
from the list of primary containment isolation valves. This proposed
change is consistent with the proposed deletion of the LCS. The LCS
lines that are connected to the main steam piping are welded and/or
capped closed to assure primary containment integrity is maintained.
The welding and post weld examination procedures will be in
accordance with American Society of Mechanical Engineers (ASME)
Code, Section III requirements. These welds and/or caps will be
periodically tested as part of the Containment Integrated Leak Rate
Test (CILRT). This proposed change does not involve an increase in
the probability of equipment malfunction previously evaluated in the
FSAR. In fact, this proposed change reduces the probability of
equipment malfunction since, upon implementation of these proposed
changes, the plant will be operated with less primary containment
isolation valves subjected to postulated failure. This proposed
change has no effect on the consequences of an accident since the
LCS lines will be welded and/or cap closed, thus assuring that the
containment integrity, isolation and leak test capability are not
compromised.
The proposed change to TS Table 3.8.4.2.1-1 deletes the LCS
motor operated valves from the list of ``Motor Operated Valves
Thermal Overload Protection - Continuous.'' The proposed change has
no effect on the probability or consequences of an accident since
the valves are eliminated and not performing a safety function.
Therefore, as discussed above, the proposed changes do not
involve a significant increase in the probability or consequences
from any accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
As stated in Section I, the proposed changes do not involve a
change to structures, components, or systems that would affect the
probability of an accident previously evaluated , nor would these
changes create any new or different kind of accident from any
previously evaluated. The proposed changes will introduce and take
credit for a new level of operational performance for existing plant
systems and components to mitigate the consequences of the accident.
The effect on this equipment has been evaluated and found to provide
an acceptable level of reliability resulting in the required level
of protection. This conclusion is based on the evaluation performed
in NEDC 31858P, Revision 2, and the plant specific seismic
evaluation provided in the Enclosure 2 [of application dated
November 21, 1994], ``MSIV Leakage Alternate Treatment Method
Seismic Evaluation.'' The Leakage Control System has been installed
to direct any leakage past the MSIVs during the LOCA; acting after
the accident has occurred. The resulting consequences of the
evaluated accidents have been affected as discussed in Section I
resulting in no significant increase in the probability or
consequences of said accident. Therefore, reliance on different
equipment than previously assumed to mitigate the consequences of an
accident does not create the possibility of a new or different kind
of accident from any accident previously evaluated.
The BWROG evaluated MSIV performance and concluded that MSIV
leakage rates up to 200 scfh per valve will not inhibit the
capability and isolation performance of the MSIVs to effectively
isolate the primary containment. Implementation of the proposed
changes does not result in modifications which could adversely
impact the operability of the MSIVs. The LOCA has been analyzed
using the main steam piping and main condenser as a treatment method
to process MSIV leakage at the proposed maximum rate of 100 scfh per
main steam line, not to exceed 300 scfh total for all four main
steam lines. Therefore, the proposed TS Section 3.6.1.2 change to
increase the allowed MSIV leakage rate does not create any new or
different kind of accident from any accident previously evaluated.
The proposed TS Section 3.6.1.4 change to eliminate the LCS does
not create the possibility of a new or different kind of accident
from any accident previously evaluated because the removal of the
LCS does not affect any of the remaining SSES Unit 1 and Unit 2
systems, and the LOCA has been re-analyzed using the proposed
alternate method to process MSIV leakage. The associated proposed
change to delete the LCS isolation valves from TS Table 3.6.3-1 and
Table 3.8.4.2.1-1 does not create the possibility of a new or
different kind of accident. The affected main steam piping will be
welded and/or capped closed to assure that the primary containment
integrity, isolation, and leak testing capability are not
compromised. The affected LCS motor operated valves will be
eliminated so their thermal overloads will not need to be bypassed.
Therefore, as discussed above, the proposed changes do not
create the possibility for any new or different kind of accident
from any accident previously evaluated.
III. Involve a significant reduction in a margin of safety.
The proposed change to TS Section 3.6.1.2 to increase the MSIV
allowable leakage does not involve a significant reduction in the
margin of safety. As discussed in the current Bases for TS Section
3/4.6.1.2, the allowable leak rate limit specified for the MSIVs is
used to quantify a maximum amount of leakage assumed to bypass
primary containment in the LOCA radiological analysis. Accordingly,
results of the re-analysis supporting these proposed changes are
evaluated against the dose limits contained in 10CFR100 for the off-
site doses, and 10CFR50, Appendix A, GDC 19, for the control room
doses. As discussed above, sufficient margin relative to the
regulatory limits is maintained even when assumptions and methods
(e.g., RG 1.3) that are considered highly conservative relative to
more realistic assumptions and methods are used in the analysis.
Results of the radiological analysis demonstrate that the
proposed changes do not involve a significant reduction in the
margin of safety. Whole body doses, in terms of margin of safety,
are insignificantly reduced by .38 rem in the control room. The
margin of safety remains constant for the LPZ whole body dose or
actually increases by .253 rem for the EAB whole body dose. The
margin of safety for thyroid dose category is reduced by 11.34 rem
at the LPZ and 4.36 rem in the control room. The margin of safety is
found to increase for the EAB thyroid dose by 2.19 rem. The margin
of safety for beta dose is insignificantly reduced by .17 rem in the
control room. The reductions in the margin of safety are not
significant since the revised calculated doses are highly
conservative yet remain well below the regulatory limits, and
therefore, a substantial margin to the regulatory limits is
maintained.
The proposed change to eliminate the LCS from TS Section 3.6.1.4
does not reduce the margin of safety, in fact, the overall margin of
safety is increased. The function of the LCS for MSIV leakage
treatment will be replaced by alternate main steam drain lines and
condenser equipment. This treatment method is effective in reducing
the dose consequences of MSIV leakage over an expanded operating
range compared to the capability of the LCS and will, thereby,
resolve the safety concern that the LCS will not function at MSIV
leakage rates higher than the LCS design capacity. Except for the
requirement to establish a proper flow path from the MSIVs to the
condenser, the proposed method is passive and does not require any
new logic control and interlocks. This proposed method is consistent
with the [[Page 505]] philosophy of protection by multiple barriers
used in containment design for limiting fission product release to
the environment. Furthermore, as previously identified, based on the
evaluations discussed in NEDC-31858P, Revision 2, and the seismic
evaluation provided in the Enclosure 2 [of application dated
November 21, 1994] report, ``MSIV Leakage Alternate Treatment Method
Seismic Evaluation,'' the design of the MSIV leakage alternate drain
pathway, meets the intent of the 10CFR100, Appendix A requirement
for seismic qualification. Therefore, the proposed method is highly
reliable and effective for MSIV leakage treatment.
The revised calculated LOCA doses remain within the regulatory
limits for the off-site and the control room. Therefore, the
proposed method maintains a margin of safety for mitigating the
radiological consequences of MSIV leakage for the proposed TS
leakage rate limit of 100 scfh per main steam line, not to exceed a
total of 300 scfh for all four main steam lines.
The proposed change to delete LCS isolation valves from TS Table
3.6.3-1 and Table 3.8.4.2.1-1 does not reduce the margin of safety.
Welded and/or capped closure of the LCS lines assures that the
primary containment integrity and leak testing capability are not
compromised. These welds and/or caps will be periodically leak
tested as part of the CILRT. The LCS motor operated valves will be
eliminated so their thermal overloads will not need to be bypassed.
Therefore, the proposed deletion of the LCS isolation valves 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,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
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 amendment request: November 18, 1994
Description of amendment request: The proposed change would revise
the Reactivity Control System Technical Specification Limiting
Conditions for Operation for boration flow paths and charging pumps by
reducing the number of operable charging pumps required for boron
addition in Mode 4 from two to one.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will not involve a significant increase in the probability or
consequences of an accident or malfunction of equipment important to
safety previously evaluated.
The Emergency Core Cooling System (ECCS) requirements assume
that only one charging pump will be available below 350 deg.F
without single failure considerations on the bases of the stable
reactivity condition of the reactor and limited core cooling
requirements. Therefore, the Mode 4 Applicability has been deleted
from LCOs 3.2.1.2 and 3.2.1.4, and was added to LCOs 3.2.1.1 and
3.2.1.3 consistent with the requirements of LCO 3.5.3.
The current Bases for the Unit 2 Technical Specification for
boration system flow paths via the charging pumps supports the use
of a similar LCO for Salem Unit 1.
The limitation for a maximum of one centrifugal charging pump to
be operable when the RCS temperature is less than or equal to
312 deg.F has been added to LCO 3.1.2.3 for clarity and is
consistent with the Cold Overpressure Protection (POPS) analysis and
the requirements of Technical Specification 3.5.3.
The requirements for Boric Acid Transfer Pump operability are
adequately addressed in Technical Specifications 3.1.2.1 and 3.1.2.2
which specify the boron injection flow paths to be operable and the
components required to perform this function. This includes the
availability of the transfer pumps to meet this Technical
Specification requirement.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated in the UFSAR.
2. Will not create the possibility of a new or different kind of
accident from any previously evaluated.
As discussed in response to Question 1 above, the proposed
amendment to the number of charging pumps required to be operable in
Mode 4 is consistent with the current Technical Specification
requirements for the ECCS LCO and the POPS. The current bases for
the Unit 2 Technical Specification for boration system flow paths
via the charging pumps supports the use of a similar LCO for Salem
Unit 1. The requirements for Boric Acid Transfer Pump operability
for Unit 1 are adequately addressed in Technical Specifications
3.1.2.1 and 3.1.2.2 which specify the boron injection flow paths to
be operable and the components required to be available to perform
this function including the transfer pumps. Therefore, the proposed
amendment does not create the possibility of a new or different kind
of accident from any previously evaluated.
3. Will not involve a significant reduction in a margin of
safety. The proposed amendment to the number of charging pumps
required to be operable in Mode 4 will not result in any changes to
the assumptions or conditions for the current ECCS analysis and POPS
analysis. The current bases for the Unit 2 Technical Specification
for boration system flow paths via the charging pumps supports the
use of a similar LCO for Salem Unit 1 (i.e., the Bases are
essentially the same). The requirements for Boric Acid Transfer Pump
operability for Unit 1 are adequately addressed in Technical
Specifications 3.1.2.1 and 3.1.2.2 which specify the boron injection
flow paths to be operable and the components required to be
available to perform this function including the transfer pumps.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama
Date of amendments request: December 19, 1994Description of
amendments request: The proposed change to Table 3.7-3 of the Technical
Specifications includes the revision to the main steam safety valve
(MSSV) setpoint tolerance from plus or minus 1 percent to plus or minus
3 percent and modifies the bases to 3/4.7.1.1 to increase the relieving
capacity of the MSSVs to at least 12,984,660 pounds per hour which
corresponds to approximately 112 percent of total secondary steam flow
at 100 percent rated thermal power. In addition, modifications to Table
3.7-1 are proposed to reduce the allowable power range neutron flux
high setpoints for multiple inoperable steam generator safety valves.
The proposed amendment includes an editorial correction to Bases 3/
4.7.1.2 to indicate required auxiliary feedwater flow at ``1133 psia''
rather than ``1133 psig.''
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 license amendment does not involve a significant
increase in the [[Page 506]] probability or consequences of an
accident previously evaluated.
These proposed changes to the Farley Technical Specifications do
not result in a condition where the design, material and
construction standards of the MSSVs that were applicable prior to
the proposed change are altered. The valves will continue to
function as designed. All applicable safety analyses have been
reviewed, evaluated or reanalyzed and all applicable safety criteria
continue to be met. No accident sequences are altered because of the
proposed amendment. The radiological consequences for the Steam
Generator Tube Rupture were reanalyzed and 10 CFR 100 criteria
continue to be met. All other FSAR radiological analyses remain
bounding. Analyses have been performed to justify the proposed high
nuclear flux setpoint changes. All acceptance criteria for these
analyses continue to be met. Therefore, the proposed amendment does
not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed license amendment does not create the
possibility of a new or different accident from any accident
previously evaluated.
The MSSVs continue to have the required pressure relieving
capacity to ensure that system design pressure remains below 110% of
shell design pressure. The proposed changes are not accident
initiators nor do they create any new accident scenarios or any new
limiting single failures. The ability of the MSSVs to respond to an
accident condition is not impaired by the proposed changes. The
proposed high nuclear flux setpoints for multiple valves out of
service ensure all applicable safety criteria for accident analyses
are met. No new accident scenarios are created by these proposed
changes. Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed license amendment does not involve a significant
reduction in the margin of safety.
Acceptance criteria for accident analysis continue to be met.
Radiological consequences for the affected Chapter 15 analysis
remain within 10 CFR 100 acceptance criteria. No safety limits or
safety system setpoint requires modification due to the proposed
changes. The current secondary side over-pressure limit of 100% of
steam generator shell design pressure is not violated. Analysis for
the high nuclear flux setpoints have verified that there is no
reduction in margin for the events analyzed. Therefore, there is not
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Attorney for licensee: M. Stanford Blanton, Esq., Balch and
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham,
Alabama 35201
NRC Project Director: William H. Bateman
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: September 9, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3/4.8.1 and its associated Bases to
improve emergency diesel generator reliability and availability.
Several surveillance requirements would be revised or eliminated, and
guidance provided in Regulatory Guide 1.9, Revision 3, and Generic
Letter 93-05 would be incorporated.
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 proposed changes do not involve a significant hazards
consideration because operation of Callaway Plant with these
changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
Emergency diesel generator operability and reliability will continue
to be assured while minimizing the number of required emergency
diesel generator starts. Also, emergency diesel generator
reliability will be enhanced by minimizing service test conditions
which can lead to premature failures.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
The performance capability of the emergency diesel generator will
not be affected. Emergency diesel generator reliability and
availability will be improved by the implementation of the proposed
changes. There is no actual impact on accident analysis.
3. Involve a Significant Reduction in the Margin of Safety.
These proposed changes do not involve a change in the
operational limits or physical design of the emergency power system.
The performance capability of the emergency diesel generator will
not be affected. Emergency diesel generator reliability and
availability will be improved by the implementation of the proposed
changes. No margin of safety is reduced.
Based on the above discussions, it has been determined that the
requested technical specification revision does not involve a
significant increase in the probability or consequences of an
accident or other adverse condition over previous evaluations; or
create the possibility of a new or different kind of accident or
condition over previous evaluations; or involve a significant
reduction in a margin of safety. The requested license amendment
does not involve a significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: Leif J. Norrholm
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: September 9, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3.8.2.1 and 3.8.2.2, 125-volt D.C.
busses for battery bank and chargers and provides for the installation
of swing chargers during the next refueling outage. Technical
Specifications 3.8.3.1 and 3.8.3.2 would be revised to address the 120-
volt A.C. Vital Busses.
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 proposed changes to the Technical Specifications do not
involve a significant hazards consideration because operation of
Callaway Plant in accordance with these changes would not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
These proposed Technical Specification changes do not involve
any hardware changes nor do they affect the probability of any event
initiators. There will be no change to normal plant operating
parameters or accident mitigation capabilities. There will be no
increase in the consequences of any accident or equipment
malfunction. [[Page 507]]
2) Create the possibility for accident or malfunction of
equipment of a different type than previously evaluated in the FSAR.
The proposed Technical Specification changes do not involve any
design changes nor are there any changes to the method by which any
safety-related plant system performs its safety function. The normal
manner of plant operation is unaffected. No new accident scenarios,
transient precursors, failure mechanisms, or limiting single
failures are introduced as a result of these changes.
Involve a significant reduction in the margin of safety.
There will be no affect [SIC] on the manner in which safety
limits or limiting safety system settings are determined, nor will
there be any effect in those plant systems necessary to assure the
accomplishment of protection functions. There will be no impact on
DNBR limits, FQ, F-delta-H, LOCA PCT, peak local power density
or any other margin of safety.
Based on the information presented above, the proposed amendment
does not involve a significant increase in the probability or
consequences of an accident previously evaluated, create the
possibility of a new or different kind of accident from any
previously evaluated, or involve a significant reduction in a margin
of safety. Therefore, it is concluded that the proposed changes meet
the requirements of 10 CFR 50.92(c) and does [SIC] not involve a
significant hazards consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: Leif J. Norrholm
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: August 27, 1993
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to be consistent with recent
revisions to 10 CFR Part 20 and 10 CFR 50.36a. Administrative changes
are also proposed.
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 will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The changes as proposed consist of revisions to the
Technical Specifications to meet new regulatory requirements as
contained in 10CFR20 and 10CFR50.36a, and other related changes of
an administrative nature. There is no change in the types and
amounts of effluents released, nor will there be any increase in
individual or cumulative occupational radiation exposures. None of
the changes proposed will affect any plant hardware, plant design,
safety limit settings, or plant system operation, and therefore do
not modify or add any initiating parameters that would significantly
increase the probability or consequences of any previously analyzed
accident.
2. The proposed change will not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The changes as proposed do not physically alter the plant
nor do they change the operation of the plant.
3. The proposed change does not involve a significant reduction
in the margin of safety. The changes will not increase the amount or
types of effluents that may be released offsite, nor do they
significantly increase individual or cumulative occupational
radiation exposures. These changes will not alter any of the
requirements or responsibilities for protection of the public and/or
employees against radiation hazards.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, Massachusetts 02110-2624
NRC Project Director: Walter R. Butler
Vermont Yankee Nuclear Power Corporation, Docket No. 50-271,
Vermont Yankee Nuclear Power Station, Vernon, Vermont
Date of amendment request: March 31, 1994
Description of amendment request: The proposed amendment would
modify the requirements for avoidance and protection from thermal
hydraulic instabilities to be consistent with the Boiling Water Reactor
(BWR) Owner's Group long-term solution Option 1-D described in the
Licensing Topical Report, ``BWR Owner's Group Long-Term Stability
Solutions Licensing Methodology, NEDO-31960 June 1991'' and NEDO-31960,
Supplement 1, dated March 1992. NEDO-31960 and NEDO-31960, Supplement
1, were accepted by the NRC staff in a letter to L.A. England (BWR
Owner's Group) dated July 12, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The implementation of BWR Owner's Group long
term stability solution Option 1-D at Vermont Yankee does not modify
the assumptions contained in the existing accident analysis. The use
of an exclusion region and the operator actions required to avoid
and minimize operation inside the region do not increase the
possibility of an accident. Conditions of operation outside of the
exclusion region are within the analytical envelope of the existing
safety analysis. The operator action requirement to exit the
exclusion region upon entry minimizes the possibility of an
oscillation occurring. The actions to drive control rods and/or to
increase recirculation flow to exit the region are maneuvers within
the envelope of normal plant evolutions. The flow biased scram has
been analyzed and will provide automatic fuel protection in the
event of an instability. Thus, each proposed operating requirement
provides defense in depth for protection from an instability event
while maintaining the existing assumptions of the accident analysis.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from an accident previously
evaluated. As stated in 1), the proposed operating requirements
either mandate operation within the envelope of existing plant
operating conditions of force specific operating maneuvers within
those carried out in normal operation. Since operation of the plant
with all of the proposed requirements are within the existing
operating basis, an unanalyzed accident will not be created through
implementation of the proposed change.
3. The proposed amendment will not involve a significant
reduction in a margin of safety. Each of the proposed requirements
for plant thermal hydraulic stability provides a means for fuel
protection. The combination of avoiding possible unstable conditions
and the automatic flow biased reactor scram provides an in depth
means for fuel protection. Therefore, the individual or combination
of means to avoid and suppress an instability supplements 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
amendment request involves no significant hazards consideration.
[[Page 508]]
Local Public Document Room location: Brooks Memorial Library, 224
Main Street, Brattleboro, Vermont 05301
Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray,
One International Place, Boston, Massachusetts 02110-2624
NRC Project Director: Walter R. Butler
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: November 29, 1994
Description of amendment request: Virginia Electric and Power
Company plans to insert fuel assemblies containing fuel rods, guide
thimble tubes, instrumentation tubes, and mid-span grids fabricated
with Westinghouse Electric Corporation's (Westinghouse's) advanced
zirconium alloy material, ZIRLO, into the Surry Units 1 and 2 reactors,
beginning with Cycle 14 at each unit. In the current fuel design, these
components are fabricated from Zircaloy-4.
Because the Technical Specifications define the fuel rod cladding
material as Zircaloy-4, implementation of this material change requires
changes to the Technical Specifications. Technical Specification
5.3.A.1 is being modified to allow the use of either Zircaloy-4 or
ZIRLO fuel rod cladding, and an additional reference for the
calculation of the heat flux hot channel factor for loss-of-coolant-
accident evaluations of fuel with ZIRLO cladding is being defined in
Technical Specification 6.2. The use of the ZIRLO fabricated guide
thimble tubes, instrumentation tubes, and mid-span grids does not
require changes to the Technical Specifications.
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:
Specifically, operation of Surry Power Station in accordance
with the Technical Specifications changes will not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated. The Surry fuel
assemblies containing fuel rods, guide thimble tubes,
instrumentation tubes and mid-span grids fabricated with ZIRLO alloy
meet the same fuel assembly and fuel rod design bases as the current
fuel assemblies fabricated with Zircaloy-4 components. In addition,
the 10 CFR 50.46 criteria will be applied to the fuel rods, guide
thimble tubes, instrumentation tubes and mid-span grids fabricated
with ZIRLO alloy. The use of these fuel assemblies will not result
in a change to the Surry Units 1 and 2 reload design and safety
analysis limits. The ZIRLO alloy is similar in chemical composition
to Zircaloy-4, and also has physical and mechanical properties
similar to those of Zircaloy-4. Thus the cladding integrity is
maintained and the structural integrity of the fuel assembly is not
affected. The ZIRLO clad fuel rods improve corrosion resistance and
dimensional stability. Since the dose predictions in the safety
analyses are not sensitive to the fuel rod cladding material changes
as specified in this report, the radiological consequences of
accidents previously evaluated in the safety analyses remain valid.
Therefore, neither the probability of occurrence nor the
consequences of any accident previously evaluated is significantly
increased.
2. Create the possibility of a new or different kind of accident
from any accident previously identified, since the Surry Units 1 and
2 fuel assemblies containing fuel rods, guide thimble tubes,
instrumentation tubes and mid-span grids fabricated with ZIRLO alloy
will satisfy the same design bases used for previous fuel regions
containing Zircaloy-4 components. Since the original design criteria
are being met, the fuel rods, guide thimble tubes, instrumentation
tubes and mid-span grids fabricated with ZIRLO alloy will not be
initiators for any new accident. Applicable design and performance
criteria will continue to be met and no single failure mechanisms
have been created. In addition, the use of these fuel assemblies
does not involve any alteration to plant equipment or procedures
which would introduce any new or unique operational modes or
accident precursors. Therefore, the possibility for a new or
different kind of accident from any accident previously evaluated is
not created.
3. Involve a significant reduction in a margin of safety. The
Surry Units 1 and 2 fuel assemblies containing fuel rods, guide
thimble tubes, instrumentation tubes and mid-span grids fabricated
with ZIRLO alloy do not change the Surry Units 1 and 2 reload design
and safety analysis limits. The use of fuel assemblies containing
fuel rods, guide thimble tubes, instrumentation tubes and mid-span
grids fabricated with ZIRLO alloy will take into consideration the
normal core operating conditions allowed in the Technical
Specifications. For each cycle reload core these fuel assemblies
will be specifically evaluated using approved reload design methods
and approved fuel rod design models and methods. This will include
consideration of the core physics analysis peaking factors and core
average linear heat rate effects. Analyses or evaluations will be
performed each cycle to confirm that the 10 CFR 50.46 criteria will
be met for the use of fuel with fuel rods, guide thimble tubes,
instrumentation tubes and mid-span grids fabricated with ZIRLO
alloy. Therefore, the margin of safety as defined in the Bases to
the Surry Units 1 and 2 Technical Specifications is not
significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Mohan C. Thadani, Acting
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: December 2, 1994
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS)
3.2 by eliminating the requirements for the charging pumps, high
concentration boric acid in the boric acid storage tanks (BASTs), the
boric acid transfer pumps, and boric acid heat tracing. Changes to TS
3.3 and Table TS 3.5.3 are also being proposed to add requirements
associated with the emergency core cooling system (ECCS) accumulators,
remove the requirements associated with the boric acid storage tanks,
and to increase the minimum required boron concentration in the
refueling water storage tank (RWST). Additionally, the surveillance
requirements involving the BASTs, associated valves and heat tracing
located in Table TS 4.1-1, Table TS 4.1-2 and Section 4.5 would be
eliminated. Supporting analysis for the limiting design basis accident
conditions have been performed using the proposed minimum RWST boron
concentration of 2400 ppm.
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:
Significant Hazards Determination for Proposed Changes to
Technical Specification (TS) 3.2 and Table TS 3.5-3.
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated;
Neither the charging pumps, the high concentration boric acid,
the BASTs, the boric acid transfer pumps nor the boric acid heat
tracing system are accident initiators. Therefore, a change to these
systems will not significantly increase the probability of an
accident previously evaluated. The effect of a reduction in initial
safety injection boron concentration on the accident analysis was
evaluated. The limiting accidents were the Large-Break Loss-of-
Coolant Accident [[Page 509]] (LOCA) and the Steam Line Break (SLB)
event. A decrease in the initial safety injection boron
concentration from 20,000 ppm to 2400 ppm will not adversely affect
the Large or Small-Break Loss-of-Coolant Accident analysis because
the evaluation models used in analyzing these accidents do not take
credit for the high concentration boric acid stored in the BASTs.
However, the evaluation models did take credit for boron in
maintaining the long term post LOCA reactor core sub-critical. An
analysis was performed which concluded that the inventory contained
in the BASTs would not be required provided the minimum RWST boron
concentration was increased to 2400 ppm. The SLB event is the other
design basis event that could be affected by the proposed
elimination of the high boron concentration BASTs as a source of
safety injection fluid. Analyses have been performed which conclude
that the BASTs are not required and that a minimum RWST boron
concentration of only 1950 ppm is sufficient to provide adequate
protection for the SLB event although 2400 ppm will be maintained to
address post-LOCA subcriticality thus providing further safety
margin. The results of these analyses indicate that the departure
from nucleate boiling (DNB) design basis continues to be met. (A
minimum Departure from Nucleate Boiling Ratio (DNBR) of 1.45 can be
maintained throughout the event.) Finally, the containment pressure
and temperature remains within the acceptable containment design
limits. Since these criteria have been satisfied, there will be no
adverse effect on the health and safety of the public and the
consequences of any accident previously evaluated have not
significantly increased.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated;
Neither the charging pumps, the removal of the BASTs from
initial SI pump injection, nor the elimination of both the boric
acid transfer pumps and the boric acid heat tracing system as
safety-related components would create the possibility of a new or
different kind of accident from any accident previously evaluated.
Furthermore, the reactivity control function of the boron in the
CVCS and SI systems is not being changed. Therefore, the proposed
changes will not adversely affect the health and safety of the
public or create the possibility of a new or different kind of
accident from any accident previously evaluated.
3) Involve a significant reduction in the margin of safety.
The reduction in the initial concentration of boron injected
into the reactor coolant system for accident mitigation has been
analyzed. These analyses conclude that all applicable criteria for a
LOCA are satisfied. A decrease in the initial safety injection boron
concentration from 20,000 ppm to 2400 ppm will not adversely effect
the Large-or Small-Break Loss-of-Coolant Accident analysis because
the evaluation models used in analyzing these accidents do not take
credit for the high concentration boric acid stored in the BASTs.
However, in order to maintain the long term post LOCA reactor core
sub-critical, a minimum RWST boron concentration of 2400 ppm is
required. To meet this requirement, the RWST boron concentration is
being raised to 2400 ppm. All criteria of 10 CFR 50.46 can be
achieved for both the Large or Small-Break LOCA with no BASTs and
2400 ppm boron in the RWST. Since all criteria of 10 CFR 50.46 are
satisfied, there is no adverse effect on the health and safety of
the public and there is not a significant reduction in the margin of
safety for these casualties.
Since both the core response and the containment response can be
limiting in the SLB event, both were considered in the boron
concentration reduction analysis. This analysis concludes that a
minimum RWST boron concentration of 1950 ppm is sufficient to
provide adequate protection for the SLB event, although a 2400 ppm
boron solution will be maintained to provide protection for the post
LOCA concerns. Since the containment pressure and temperature
remains within the acceptable containment design limits, and a
minimum DNBR of 1.45 can be maintained throughout the event, there
is not a significant reduction in the margin of safety for this
event and therefore there is no adverse effect on the health and
safety of the public.
These proposed changes involve the conversion of the TS to Word
Perfect format now being used at WPSC. Minor typographical errors
and format inconsistencies were corrected. These proposed changes
are administrative in nature; accordingly, these proposed changes do
not involve a significant hazards consideration.
Additionally, the proposed changes are similar to example
C.2.e.(i) in 51 FR 7751. Example C.2.e.(i) states that changes which
are purely administrative in nature; i.e., to achieve consistency
throughout the Technical Specifications, correct an error, or a
change in nomenclature, are not likely to involve a significant
hazard.
Significant Hazards Determination for Proposed Changes to Table
TS 4.1-1, ``Minimum Frequencies for Checks, Calibrations and Test of
Instrument Channels'' and Table TS 4.1-2 ``Minimum Frequencies for
Sampling Tests''
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will 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 the margin of safety.
The above listed surveillance requirements insure BAST
operability. The BASTs will no longer be relied upon as a source of
boron for safety injection, and will serve no safety related
function. Whether the BASTs are operable or not will have no effect
on plant safety. Therefore, elimination of the surveillance
requirements which insure BAST operability is possible without any
adverse effect on the health and safety of the public and presents
no significant hazards.
Significant Hazards Determination for Proposed Changes to
Technical Specification TS 3.3 and Section 4.5.
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Neither the RWST, the boron solution contained within the RWST
nor valves SI-3, SI-4A/B are accident initiators. Therefore, a
change to these systems will not significantly increase the
probability of an accident previously evaluated. The effect of a
reduction in initial Safety Injection boron concentration on the
accident analysis was evaluated. The limiting accidents were the
Large-Break Loss-of-Coolant Accident (LOCA) and the Steam Line Break
(SLB) event. A decrease in the initial safety injection boron
concentration from 20,000 ppm to 2400 ppm will not adversely effect
the Large or Small-Break Loss-of-Coolant Accident analysis because
the evaluation models used in analyzing these accidents do not take
credit for the high concentration boric acid stored in the BASTs.
However, the evaluation models did take credit for boron in
maintaining the long term post LOCA reactor core sub-critical. An
analysis was performed which concluded that the BASTs could be
eliminated provided the minimum RWST boron concentration was
increased to 2400 ppm. The SLB event is the other design basis event
that could be affected by the proposed elimination of the high
concentration BASTs as a safety-related source for reactivity
control injection fluid. However, analyses have been performed which
conclude that a minimum RWST boron concentration of only 1950 ppm is
sufficient to provide adequate protection for the SLB event although
2400 ppm will be maintained to address post-LOCA subcriticality thus
providing further safety margin. The results of these analyses
indicate that the departure from nucleate boiling (DNB) design basis
continues to be met. (A minimum Departure from Nucleate Boiling
Ratio (DNBR) of 1.45 can be maintained throughout the event.)
Furthermore, maintaining the suction of the SI pumps to the RWST
with valves SI-4A or SI-4B open with power removed places the system
in a normal SI sequence and eliminates the requirement to switch
suction from the BASTs to the RWST. This eliminates a potential
failure mechanism and increases the overall reliability of the ECCS
system. Finally, the containment pressure and temperature remains
within the acceptable containment design limits.
Since these criteria have been satisfied, there will be no
adverse effect on the health and safety of the public and the
consequences of any accident previously evaluated have not
significantly increased.
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This change to the Technical Specifications allows use of 2400
ppm boron for safety injection. SI pump suction would be directly
from the RWST. This eliminates [[Page 510]] the necessity of
shifting suction from the BASTs to the RWST, reducing the complexity
of the operation. Since the pumps remain connected to the RWST
throughout the injection phase, there is no possibility of a new or
different kind of accident from any accident previously evaluated.
Neither the reduction in initial boron concentration for safety
injection, nor the increase in the boron concentration in the RWST
would create the possibility of a new or different kind of accident
from any accident previously evaluated.
Lastly, the reactivity control function of the boron in the CVCS
and SI systems is not being changed. Therefore, the proposed changes
will not adversely affect the health and safety of the public or
create the possibility of a new or different kind of accident from
any accident previously evaluated.
3) Involve a significant reduction in the margin of safety.
The change in concentration of boron injected into the primary
system for accident mitigation has been analyzed. These analyses
conclude that all applicable criteria for a LOCA are satisfied. A
change in safety injection boron concentration to 2400 ppm will not
adversely affect the Large or Small-Break LOCA analysis because the
evaluation model codes used in analyzing these accidents did not
take credit for boron. However, a minimum RWST boron concentration
of 2400 ppm is required to maintain long term post LOCA reactor core
sub-criticality. To meet this requirement, the RWST minimum boron
concentration is being raised to 2400 ppm. All criteria of 10 CFR
50.46 can be achieved for both the Large or Small-Break LOCA with
2400 ppm boron in the RWST. Since all criteria of 10 CFR 50.46 are
satisfied, there is no adverse effect on the health and safety of
the public and there is not a significant reduction in the margin of
safety for these casualties.
Since both the core response and the containment response can be
limiting in the SLB event, both were considered in the boron
concentration reduction analysis. Although a minimum RWST boron
concentration of 1950 ppm is sufficient to provide adequate
protection for the SLB event, a 2400 ppm boron solution will be
maintained to provide protection for the post large break LOCA
concerns. Since the containment pressure remains below the design
pressure, and a minimum DNBR of 1.45 can be maintained throughout
the event, there is not a significant reduction in the margin of
safety for this event.
These proposed changes involve the conversion of the TS to Word
Perfect format now being used at WPSC. Minor typographical errors
and format inconsistencies were corrected. These proposed changes
are administrative in nature; accordingly, these proposed changes do
not involve a significant hazards consideration.
Additionally, the proposed changes are similar to example
C.2.e.(i) in 51 FR 7751. Example C.2.e.(i) states that changes which
are purely administrative in nature; i.e., to achieve consistency
throughout the Technical Specifications, correct an error, or a
change in nomenclature, are not likely to involve a significant
hazard.
Significant Hazards Determination for Proposed Changes to
Technical Specification (TS) Section 4.5 ``Emergency Core Cooling
System and Containment Air Cooling System Tests.''
The proposed changes were reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed changes will 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 the margin of safety.
The above listed surveillance requirements insure BAST
operability. The BASTs will no longer be relied upon as a source of
boron for safety injection, and will serve no safety related
function. Whether the BASTs are operable or not will have no effect
on plant safety. Therefore, elimination of the surveillance
requirements which insure BAST operability is possible without any
adverse effect on the health and safety of the public and presents
no significant hazards.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497.
NRC Project Director: Leif J. Norrholm
Previously Published Notices Of Consideration Of Issuance Of
Amendments ToFacility 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.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear
One,Unit No. 2, Pope County, Arkansas
Date of amendment request: November 29, 1994
Brief description of amendment request: The proposed amendment
would delete requirements to perform the full complement of steam
generator surveillances as outlined in the technical specifications
(TSs) when the steam generators are subjected to special inspections
that are in addition to inspections that are required by the TSs.
Date of individual notice in the Federal Register: December 5, 1994
(59 FR 62416)
Expiration date of individual notice: January 4, 1995
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
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
[[Page 511]] Assessment as indicated. All of these items are available
for public inspection at the Commission's Public Document Room, the
Gelman Building, 2120 L Street, NW., Washington, DC, and at the local
public document rooms for the particular facilities involved.
Arizona Public Service Company, et al., Docket No. STN 50-528, Palo
Verde Nuclear Generating Station, Unit No. 1, Maricopa County,
Arizona
Date of application for amendment: November 22, 1994
Brief description of amendment: The amendment adds a note to
Technical Specification Table 3.7-2. The note allows continuous
operation of Unit 1 during Cycle 5 at 100-percent maximum steady state
power with one main steam safety valve inoperable per steam generator.
This note applies only during the current fuel cycle (Cycle 5) for Unit
1.
Date of issuance: December 19, 1994
Effective date: December 19, 1994
Amendment No.: 87
Facility Operating License No. NPF-41: The amendment revised the
Technical Specifications.
Public comments requested as to proposed no significant hazards
consideration: Yes (59 FR 61907, dated December 2, 1994). 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 December 19, 1994, but stated that, if the Commission makes
a final no significant hazards consideration determination, any such
hearing would take place after issuance of the amendment. The
Commission's related evaluation of the amendment, finding of exigent
circumstances, and final determination of significant hazards
consideration is contained in a Safety Evaluation dated December 19,
1994.
Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail
Station 9068, Phoenix, Arizona 85072-3999
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: October 7, 1994
Brief description of amendment: The amendment revises the
introduction to TS Section 6.9.3.3 to require the approved revision
number for the referenced analytical methods to be listed in the Core
Operating Limits Report. The methodology referenced in 6.9.3.3.b.f (XN-
NF-82-49(A)) has been updated to clarify that all supplements are
included. New methodologies ANF-89-151(A) and EMF-92-081(A) will be
added to TS Section 6.9.3.3.b.
Date of issuance: December 12, 1994
Effective date: December 12, 1994
Amendment No.: 154
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55868)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 12, 1994.No significant hazards
consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College, Hartsville, South Carolina 29550
Consolidated Edison Company of New York, Docket No. 50-247, Indian
PointNuclear Generating Unit No. 2, Westchester County, New York
Date of application for amendment: February 18, 1994, as
supplemented by letters dated June 3, 1994, November 1, 1994, December
2, 1994, December 14, 1994 and December 16, 1994.
Brief description of amendment: The amendment revises surveillance
intervals for the Vapor Containment Sump Discharge Flow and Temperature
Channel, the Loss of Power Undervoltage and Degraded Voltage Relays,
and the Control Rod Protection System Trip to accommodate a 24-month
refueling cycle. In addition it changes the trip setpoint for the
Control Rod Protection System Trip. These revisions are being made in
accordance with the guidance provided by Generic Letter 91-04,
``Changes in Technical Specification Surveillance Intervals to
Accommodate a 24-Month Fuel Cycle.''
Date of issuance: December 20, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 179
Facility Operating License No. DPR-26: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 28, 1994 (59 FR
22003) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: July 19, 1994
Brief description of amendments: These amendments change Technical
Specification 3.1.5 for each unit for the standby liquid control system
(SLCS) to remove the operability requirement for the SLCS while the
plant is in Operational Condition 5 (refueling) with any control rod
withdrawn, and to delete the 18-month system surveillance requirement
(Surveillance Requirement 4.1.5.d.3).
Date of issuance: December 20, 1994
Effective date: December 20, 1994
Amendment Nos.: 136 and 106
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42344)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 20, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: July 22, 1994
Brief description of amendments: The amendment removes the
surveillance frequency details regarding 10 CFR Part 50, Appendix J,
Types B and C testing from the Technical Specifications.
Date of issuance: December 19, 1994
]Effective date: December 19, 1994
Amendment Nos. 83 and 44
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47180) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 19, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
[[Page 512]] High Street, Pottstown, Pennsylvania 19464.
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: October 7, 1994
Brief description of amendment: The amendment revises Technical
Specification (TS) 4.6E.4 and the associated Bases to establish that
the manual cycling of reactor coolant system safety/relief valves
(SRVs) during plant startups is to be accomplished within 12 hours
after steam pressure and flow are adequate to perform the testing. TS
4.6E.4 had previously required that this testing be performed within 12
hours of continuous power operation at a reactor steam dome pressure of
at least 940 psig. The amendment also makes several editorial changes
to clarify the intent of TSs involving SRV testing and performance
requirements.
Date of issuance: December 16, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 219
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55889)The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 16, 1994.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: September 9,
1994
Brief description of amendments: The amendments revise the
Technical Specification surveillance requirements regarding visual
inspection of snubbers and are consistent with the guidance provided in
Generic Letter 90-09, ``Alternative Requirements for Snubber Visual
Inspection Intervals and Corrective Actions.''
Date of issuance: December 12, 1994
Effective date: December 12, 1994
Amendment Nos. 161 and 142
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55889)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 12, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
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: March 28,
1994, as supplemented June 1, 1994, and August 24, 1994
Brief description of amendments: The amendments revise the
sustained degraded voltage relay trip setpoint and the allowable value
due to changes in the switchyard configuration.
Date of issuance: December 14, 1994
Effective date: December 14, 1994
Amendment Nos. 162 and 143
Facility Operating License Nos. DPR-70 and DPR-75. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29633) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 14, 1994.No significant
hazards consideration comments received: No
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: October 7, 1993 (TS 313)
Brief description of amendments: The changes include the addition
of the high range primary containment radiation monitors and recorders
and the wide range gaseous effluent radiation recorder and monitor,
which were installed at the Browns Ferry facility in response to NUREG
0737 ``Clarification of TMI Action Plan Requirements'' and GL 83-36,
into the technical Specifications (TS) for Units 1 and 3. Similar
changes to the Unit 2 TS were issued previously (Amendment Nos. 125 and
171). The amendment also clarifies that the high range primary
containment radiation recorders and monitors are both part of the
instrument loop. The amendment contains administrative typographical
changes which provide consistency for the TS tables and footnotes for
Units 1 and 3.
Date of issuance: December 21, 1994
Effective Date: December 21, 1994
Amendment Nos.: 214, 230, 187
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67863)The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 21, 1994.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: November 11, 1994, as supplemented by
letter dated November 16, 1994.
Brief description of amendments: The proposed amendment would
modify Comanche Peak Steam Electric Station Technical Specification
Table 4.8-1, ``Diesel Generator Test Schedule,'' by excluding two valid
failures of the Unit 2 Train B diesel generator from contributing
towards an accelerated test schedule.
Date of issuance: December 9, 1994
Effective date: December 9, 1994
Amendment Nos.: Unit 1 - Amendment No. 33; Unit 2 - Amendment No.
19
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.Public comments requested as to
proposed no significant hazards consideration: Yes (59 FR 69399, dated
November 23, 1994). 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 December 23,
1994, but stated 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 final determination of no significant
hazards consideration is contained in a Safety Evaluation dated
December 9, 1994.
Local Public Document Room location: University of Texas at
Arlington library, Government [[Page 513]] Publications/Maps, 702
College, P.O. Box 19497, Arlington, Texas 76019.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: March 29, 1994
Brief description of amendments: These amendments revise Point
Beach Nuclear Plant Technical Specification 15.3.2, ``Chemical and
Volume Control System,'' by eliminating the necessity for high
concentration boric acid and removing the operability requirements for
the associated heat tracing. The basis for Section 15.3.2 and
applicable surveillances in Table 15.4.1-2 are also revised to support
the above changes.
Date of issuance: December 12, 1994
Effective date: Date of issuance, to be implemented within 45 days.
Amendment Nos.: 158 & 162
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: July 20, 1994 (59 FR
37091) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 12, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301,
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks,
Manitowoc County, Wisconsin
Date of application for amendments: September 12, 1994
Brief description of amendments: These amendments revise Point
Beach Nuclear Plant Technical Specification (TS) 15.3.3, ``Emergency
Core Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan
Coolers, and Containment Spray,'' by incorporating allowed outage times
similar to those contained in NUREG-1431, Revision 0, ``Westinghouse
Owner's Group Improved Standard Technical Specifications,'' and by
clarifying the operability requirements for the service water pumps.
The changes also clarify the completion times for placing a unit in hot
or cold shutdown, if a limiting condition for operation cannot be met.
Date of issuance: December 21, 1994
Effective date: Date of issuance, to be implemented within 45 days
Amendment Nos.: 159 & 163
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 24, 1994 (59 FR
53844)The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 21, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 27th day of December 1994.
For The Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of
Nuclear Reactor Regulation.
[Doc. 95-5 Filed 1-3-95; 8:45 am]
BILLING CODE 7590-01-F