[Federal Register Volume 61, Number 70 (Wednesday, April 10, 1996)]
[Notices]
[Pages 15985-16005]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-8786]
-----------------------------------------------------------------------
NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from March 16, 1996, through March 29, 1996. The
last biweekly notice was published on March 27, 1996 (61 FR 13521).
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.
[[Page 15986]]
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 May 10, 1996, the licensee may file a request for a hearing with
respect to issuance of the amendment to the subject facility operating
license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a 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.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: September 30, 1994, as supplemented
September 18, 1995, January 19 and March 15, 1996
Description of amendment request: Currently, the steam generators
(SGs) in place in the Catawba units are Westinghouse Model ``D'' type
preheat SGs. The tube degradation levels in the SGs at Catawba Unit 1
have affected the reliability of the unit. Therefore, these generators
are scheduled to be replaced with feedring SGs designed by Babcock
[[Page 15987]]
& Wilcox International. The design differences and analysis changes to
support the feedring SGs result in the need to change the Technical
Specifications (TS) in the following areas: (a) revise low-low SG water
level for the reactor trip setpoint in TS Table 2.2-1 and for auxiliary
feedwater actuation in TS Table 3.3-4, (b) revise high-high SG water
level setpoint for turbine trip and feedwater isolation in TS Table
3.3-4, (c) delete reference to SG tube repair methods which will no
longer be applicable after the replacement of the SGs and clarify
initial surveillances, (d) revise reactor coolant system volume, (e)
update Topical Report revision numbers in the Administrative Controls
Section 6.9 of the TS, and (f) change the nominal average temperature
in TS Table 2.2-1 for the reactor trip system setpoints to reflect the
value incorporated into the safety analyses for the replacement SGs.
The change made in the September 30, 1994, submittal, to reduce the
steam line safety valve lift settings in TS Table 3.7-2, was withdrawn
in the September 18, 1995, submittal. The January 19, 1996, submittal
proposed changes to reflect the NRC's approved revisions to Topical
Reports DPC-NE-3000 and DPC-NE-3002. The March 15, 1996, submittal
provided additional information in response to NRC staff requests and
also updated and clarified the involved TS pages including changes made
to these TS pages by license amendments issued on other topics since
the original application dated September 30, 1994.
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:
Operation of Catawba Unit 1 in accordance with the proposed
changes to the Technical Specifications will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The low-low steam generator water
level reactor trip setpoint, the high-high steam generator water
level setpoint for turbine trip and feedwater isolation, and the
low-low steam generator water level setpoint for auxiliary feedwater
initiation are changing to support operation with the replacement
steam generators. These setpoints were chosen both to optimize plant
operation, and ensure that all applicable acceptance criteria are
met for licensing basis safety analysis. These setpoints do not
contribute to the initiation of any accident evaluated in the
Catawba FSAR [Final Safety Analysis Report] and have no adverse
impact on system operation, therefore it can be concluded that these
changes will not significantly increase the probability or
consequences of an accident evaluated in the FSAR.
The increase in Reactor Coolant System volume due to the
replacement steam generators will not increase the probability or
consequences of an accident previously evaluated. The increase in
volume has no effect on the probability of occurrence of any
accident evaluated in the FSAR. The mass and energy release inside
containment due to postulated loss of coolant accidents inside
containment has been analyzed to ensure that the peak containment
pressure limit is not exceeded. All Chapter 15 reanalysis which was
required due to the replacement steam generators assumed the new
Reactor Coolant System volume. Since the results of these analyses
show the applicable acceptance criteria continue to be met, it can
be concluded that the consequences of an accident previously
evaluated are not significantly increased due to this change.
Operation of Catawba Unit 1 in accordance with the proposed
changes to the Technical Specification will not create the
possibility of a new or different accident from any accident
previously evaluated. The proposed changes to revise the low-low
steam generator water level reactor trip setpoint, high-high steam
generator water level setpoint for turbine trip and feedwater
isolation, and low-low steam generator water level setpoint for
auxiliary feedwater initiation ensure that the appropriate
acceptance criteria for FSAR Chapter 15 transients which rely on
these functions are met for operation with the replacement steam
generators. ... The increase in Reactor Coolant System volume is
taken into account in the analysis of the mass and energy release
due to a postulated loss of coolant inside containment, and Chapter
15 events which have been reanalyzed due to replacement of the steam
generators. As discussed above, the proposed changes will not
introduce the possibility of a new or different accident from any
previously evaluated, they will ensure that transients that take
credit for these functions and dose analyses meet applicable
acceptance criteria for operation with the replacement steam
generators.
Operation of Catawba Unit 1 in accordance with the proposed
changes to the Technical Specifications will not involve a
significant reduction in a margin of safety. The proposed changes
were made to ensure that transients that rely on low-low steam
generator water level reactor trip setpoint, high-high steam
generator water level setpoint for turbine trip and feedwater
isolation, and low-low steam generator water level setpoint for
auxiliary feedwater actuation meet applicable acceptance criteria.
... The proposed change in the Reactor Coolant System volume will
not involve a significant reduction in a margin of safety. The
increased volume affects the mass and energy release due to a
postulated loss of coolant accident inside containment and the other
Chapter 15 events which were reanalyzed due to replacement of the
steam generators. This event has been analyzed and the results are
within current acceptable limits. As discussed above, the acceptance
criteria for FSAR transients which are affected by these proposed
changes continue to be met, therefore there is no significant
reduction in the margin of safety.
Changes to the steam generator surveillance requirements will
simply delete inspection requirements which are no longer applicable
after installation of the replacement steam generators. References
to F* criteria, interim plugging criteria, and sleeving are deleted
since these repair criteria were approved for use on the current
steam generators. Since these changes only delete criteria which
will no longer be applicable and cannot be used, no significant
hazards considerations are involved.
The changes to Technical Specification 6.9.1.9 are
administrative in nature. These changes are being made to reflect
the most recent revisions of DPC-NE-3002 and DPC-NE-3000, which
includes changes associated with the replacement steam generators.
These topical report revisions [have been] reviewed and approved for
use regarding McGuire and Catawba Nuclear Stations. Since these
changes are administrative in nature, no significant hazards
considerations are involved.
The proposed change to Technical Specifications [average coolant
temperature in Table 2.2-1] does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Changing the value for [the average coolant temperature]
in Notes 1 and 3 of Table 2.2-1 will update the value to agree with
[the average coolant temperature] assumed in the applicable safety
analyses for replacement of the steam generators. Acceptable results
were obtained for all required reanalyses. The probability of an
accident will not be significantly affected by operation with the
new [average coolant temperature] value, because all equipment will
be operated within acceptable design limits. The consequences of
previously evaluated accidents which are affected by this change
have been evaluated, and have been determined to be within
acceptable limits.
This proposed change [to TS Table 2.2-1] will not create the
possibility of a new or different kind of accident from any
previously evaluated. This change does not change the physical
configuration of the plant, and all analyses which are affected by
replacement of the steam generators have been determined to have
acceptable results assuming this value for [average coolant
temperature].
This proposed change to the Technical Specifications [Table 2.2-
1] will not involve a significant reduction in the margin of safety.
All safety analyses which were affected by replacement of the steam
generators assumed this value for [average coolant temperature] and
the results were determined to be within previously acceptable
limits.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
[[Page 15988]]
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: January 12, 1996, as supplemented March
4, 1996
Description of amendment request: This request was previously
published in the Federal Register on January 31, 1996 (61 FR 3498). It
is being renoticed to provide clarification to the scope of the
original request. Compliance with 10 CFR Part 50, Appendix J, provides
assurance that the primary containment, including those systems and
components that penetrate the primary containment, do not exceed the
allowable leakage rate values specified in the Technical Specifications
(TS) and Bases. The allowable leakage rate is determined so that the
leakage assumed in the safety analyses is not exceeded.
On September 12, 1995, the NRC approved issuance of a revision to
10 CFR Part 50, Appendix J, which was subsequently published in the
Federal Register on September 26, 1995, and became effective on October
26, 1995. The revision added Option B ``Performance-Based
Requirements'' to Appendix J to allow licensees to voluntarily replace
the prescriptive testing requirements of Appendix J with testing
requirements based on both overall and individual component leakage
rate performance.
Regulatory Guide 1.163, ``Performance-Based Containment Leak Test
Program,'' was developed as a method acceptable to the staff for
implementing Option B. Accordingly, the licensee has submitted, in its
application dated January 12, 1996, proposed changes to the TS to
implement 10 CFR Part 50, Appendix J, Option B, by referring to
Regulatory Guide (RG) 1.163, ``Performance-Based Containment Leakage-
Test Program.'' Although the licensee's proposal indicated that it was
consistent with RG 1.163, it did not include the clarifying changes to
the TS that would require the visual examination of containment systems
to be consistent with the guidance of RG 1.163. The licensee submitted
a supplement, dated March 4, 1996, to its January 12, 1996 proposal,
which proposes such changes to TS Surveillance Requirements 4.6.1.6 and
4.6.1.7 and associated Bases.
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.
Containment leak rate testing is not an initiator of any
accident; the proposed change does not affect reactor operations or
accident analysis, and has no significant radiological consequences.
Therefore, this proposed change will not involve an increase in the
probability or consequences of any previously-evaluated accident.
2. The proposed change will not create the possibility of any
new not previously evaluated.
The proposed change does not affect normal plant operations or
configuration, nor does it affect leak rate test methods. The test
history at Catawba (no ILRT [integral leak rate test] failures)
provides continued assurance of the leak tightness of the
containment structure.
3. There is no significant reduction in a margin of safety.
The proposed changes are based on NRC-accepted provisions, and
maintain necessary levels of reliability of containment integrity.
The performanced-based approach to leakage rate testing recognizes
that historically good results of containment testing provide
appropriate assurance of future containment integrity; this supports
the conclusion that the impact on the health and safety of the
public as a result of extended test intervals is negligible.
Based on the above, no significant hazards consideration is
created by the proposed change.
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: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley
Power Station, Unit 2, Shippingport, Pennsylvania
Date of amendment request: March 11, 1996
Description of amendment request: The proposed amendment would
increase the alarm setpoints of the in-containment high range area and
containment purge radiation monitors. These alarm setpoints are
specified in Table 3.3-6 of Technical Specification 3.3.3.1. The
proposed amendment would also include several editorial changes.
The proposed change to the in-containment high range area radiation
monitor alarm setpoint would make the setpoint consistent with the
Beaver Valley Power Station Emergency Action Levels (EALs) approved by
the NRC in August 1994. These EALs use the in-containment high
radiation area monitors as indication of fission product barrier
challenges or failures.
The containment purge radiation monitors are provided to: (1)
analyze the ventilation effluent from the reactor containment building,
(2) detect abnormal releases and isolate the release if the setpoint is
reached or exceeded, and (3) alert refueling personnel of the need to
evacuate affected areas so as to maintain occupational exposures as low
as reasonably achievable. The proposed increase in this setpoint value
provides alarm and isolation based on offsite dose considerations and
will provide greater operational flexibility since inadvertent
engineered safety feature actuations due to evacuation alarms caused by
minor (greater than three times background) increases in radiation
levels will be minimized.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed monitor alarm setpoint changes and editorial
changes are administrative in nature. Should the in-containment high
range area monitors fail to annunciate or give a false alarm, there
would be no effect on any other plant equipment or systems. These
monitors are safety related; however, they do not initiate any
safety function, nor do they interface with any other safety related
system. The monitors' alarm as a visual (lighted icon) and audible
alarm in the control room. The operator is then responsible for
taking any corrective actions necessary, based on the alarm and
Emergency Action Level (EAL) guidelines. The in-containment high
range area monitors do not provide for any automatic actions of
other equipment or systems when an alarm condition occurs.
The containment purge monitors are also safety related with the
ability for an operator to input a radiation level value for high
alarm levels during Mode 6, which upon actuation, create both a
visual (lighted icon) and
[[Page 15989]]
audible alarm in the control room. At the high alarm level, each
monitor automatically sends a signal to close the purge supply and
exhaust isolation dampers in the containment building. A change in
the value of the alarm setpoint has no effect on the performance of
the containment purge and exhaust system. The high alarm and
subsequent automatic termination of a radioactive release will now
be based on offsite dose considerations. There is no credible
failure of the monitors associated with a change of the alarm
setpoint value.
The operating and design parameters of the subject radiation
monitors will not change. The proposed change affects only
theradiation level at which an alarm condition is created and does
not affect any accident assumptions. The in-containment high range
area monitors' alarm setpoint change will not affect the
radiological consequences of an accident. However, since the
containment purge monitors revised setpoint is based on offsite
doses consequences and is a higher value than the current setpoint
of three times the background radiation level, the postulated
offsite radiological consequences of a fuel handling accident inside
containment would be increased. An analysis of a fuel handling
accident inside containment with the purge and exhaust system
discharging through the Supplementary Leak Collection and Release
System (SLCRS) filter trains was performed and a summary of this
analysis is to be added to Chapter 15 of the Updated Final Safety
Analysis Report (UFSAR). The analysis which determined the
containment purge monitors' setpoint postulated offsite doses that
are less than a small fraction (less than twenty-five percent) of
the 10 CFR Part 100 guidelines. The fuel handling accident inside
containment calculation demonstrated control room operator doses
that comply with General Design Criteria (GDC) 19. Therefore, the
increased radiological consequences of the change in the alarm
setpoint are acceptable. The analysis assumed no isolation, so
isolation actuated by the monitor alarm will reduce doses further.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed radiation monitor alarm revisions cannot initiate a
new type of accident. The referenced radiation monitors' alarms
cannot initiate a new type of accident, since even a failure of the
monitor itself cannot serve as the initiating event of an accident.
Operator action is not made solely on a radiation monitor alarm;
other plant condition indicators are also evaluated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The in-containment high range area monitors have no capability
to mitigate the consequences of an accident and do not interface
with any safety related system. These monitors are safety related
channels which provide indication to the operator of the integrity
of the fission product barriers incontainment. This indication,
combined with other indications of plant conditions may direct an
operator to take action to mitigate the consequences of an accident.
The alarm setpoint itself does not perform any specific safety
related function and the trip value is not referenced in the UFSAR,
nor does any site design basis document take credit for this
setpoint. Safety limits and limiting safety system settings are not
affected by this proposed change. The site will continue to meet the
requirements of 10 CFR Part 100 which limits offsite dose following
a postulated fission product release.
The containment purge monitors' revised setpoint is based on
offsite dose consequences and is a higher value than the current
setpoint of three times the background radiation level. Thus the
postulated offsite radiological consequences of a fuel handling
accident inside containment are increased which reduces the current
margin of safety. An analysis of a fuel handling accident inside
containment with the purge and exhaust system discharging through
the SLCRS filter trains was performed and a summary of this analysis
will be added to Chapter 15 of the UFSAR. The analysis postulated
offsite doses to be less than twenty-five percent of the 10 CFR Part
100 guidelines and control room operator doses that comply with GDC
19. The analysis shows that the increased radiological consequences
of the change in the alarm setpoint are acceptable. Further, the
analysis assumed that no isolation would occur; therefore, isolation
actuated by the monitors' alarm will reduce the postulated doses.
Therefore, use of the proposed technical specification would 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: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Florida Power and Light Company, Docket Nos. 50-250 and 50-251,
Turkey Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: March 5, 1996
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) as follows:
(1) TS Surveillance Requirement (SR) 4.4.3.3: Delete the
requirement for testing the switching capability for pressurizer heater
power supplies on an 18-month interval.
(2) TS SR 4.5.2.d: Change the containment sump inspection
requirements from each containment entry to once daily if a containment
entry has been made and upon the final entry prior to establishing
CONTAINMENT INTEGRITY.
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) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed amendments conform to the uidance given in
Enclosure 1 of the NRC Generic Letter 93-05. The overall functional
capabilities of the pressurizer heater system and the Emergency Core
Cooling System (ECCS) will not be modified by the proposed changes.
These amendments will not involve a significant increase in the
probability or consequences of an accident previously evaluated for
the following reasons:
(1) Deleting the requirement to test the switching capabilities
of the pressurizer heater emergency power supplies will reduce an
unnecessary testing requirement since the pressurizer heaters are
already connected to the emergency bus.
(2) Increasing the interval of containment sump inspections to
once daily if containment has been entered and upon final entry will
reduce unnecessary personnel exposure from performance of
containment sump inspections for each containment entry.
[The staff notes that although statement (2) is correct, it does
not provide a reason why the amendments will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. The staff finds that once daily
inspection of the containment adequately ensures that the
containment sump remains free of debris.]
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The use of the proposed changes to the TS can not create the
possibility of a new or different kind of accident from any accident
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined
[[Page 15990]]
in the facility operating license. No new failure mode is introduced
due to the surveillance changes and inspection requirements, since
the proposed changes do not involve the addition or modification of
equipment nor do they alter the design or operation of affected
plant systems.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The operating limits and functional capabilities of the affected
systems are unchanged by the proposed amendments. The proposed
changes to the TS which establish new or clarify old surveillance
and inspection requirements [are] consistent with the NRC Generic
Letter 93-05 line-item improvement guidance [and] do not
significantly reduce any of the margins of safety even though the
number of surveillances is decreased. These requested amendments are
justified by the following reasoning from NUREG-1366:
(1) The surveillance or inspection results in radiation exposure
to plant personnel which is not justified by the safety significance
of the surveillances as in the case of the containment sump
inspection requirements.
(2) The surveillance places an unnecessary burden on plant
personnel because the time required is not justified by the safety
significance of the surveillance as in the emergency power switching
requirements for the pressurizer heater system.
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: Florida International
University, University Park, Miami, Florida 33199
Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis &
Bockius, 1800 M Street, NW., Washington, DC 20036
NRC Project Director: Eugene V. Imbro
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: February 22, 1996 (AEP:NRC:1243)
Description of amendment requests: The proposed amendments would
revise the technical specifications to reference NRC Regulatory Guide
1.9, Revision 3 rather than NRC Regulatory Guide 1.108, Revision 1
criteria for the determination of a valid diesel generator test.
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:
Per 10 CFR 50.92, proposed changes do not involve a significant
hazards consideration if the changes do not:
1. involve a significant increase in the probability [or]
consequences of an accident previously evaluated,
2. create the possibility of a new or different kind of accident
from any accident previously evaluated, or
3. involve a significant reduction in a margin of safety
Criterion 1
This amendment request does not involve a significant increase
in the probability or consequences of an accident previously
evaluated because the proposed change to the T/S [technical
specifications] does not affect the assumptions, parameters, or
results of any UFSAR [updated final safety analysis report] accident
analysis.
The proposed amendment does not modify any existing equipment,
and the proposed acceptance criteria for diesel generator testing
will conform to NRC guidance. Based on these considerations, it is
concluded that the changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Criterion 2
The proposed changes do not involve physical changes to the
plant or changes in plant operating configuration. The proposed
changes update guidance for diesel generator testing. Thus, it is
concluded that the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
Criterion 3
The proposed changes update guidance for the testing of diesel
generators. The guidance is endorsed by the NRC in Regulatory Guide
1.9, and compliance with this guidance will ensure the operability
of the diesel generators. Thus, there is no 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 requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: Mark Reinhart, Acting
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: December 7, 1995
Description of amendment request: The proposed change will remove
the requirement that primary containment always be purged or vented
through the standby gas treatment (SBGT) system and adds requirements
that would limit the use of SBGT for purging and venting. The proposed
amendment also makes editorial changes and revises the associated Bases
section.
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:
NNECO has reviewed the proposed change in accordance with
10CFR50.92 and concluded that the change does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed change will allow primary containment to be purged
or vented without the use of the SBGT system. This change only
modifies the alignment of the atmospheric control system for purging
or venting containment. The change does not affect any primary
system, nor does it affect the ability of the containment isolation
valves to close. As such, the proposed change can not affect the
probability of occurrence of an accident previously analyzed. This
change increases the possibility that some initial post-accident
containment atmosphere could be released directly to the atmosphere
at the top of the 375 foot stack prior to the closure of the
containment isolation valves. However, this condition is bounded by
the original radiological release analysis. This is balanced by the
increased likelihood that post-accident reactor building atmosphere
(from the time that the containment isolation valves close) is
processed by the SBGT system.
The proposed technical specification also establishes strict
controls for the use of the SBGT system for purging and venting
containment atmosphere. This includes disabling the automatic
initiation of the train not in use and relying on a dedicated
operator to initiate the remaining train, should a DBA [design basis
accident] occur. Since SBGT system operation does not affect the
initiation of any postulated accident, disabling the automatic
initiation and relying upon operator action to start the remaining
train can not affect the probability of an accident previously
evaluated. The failure of the train to start within one minute
following the DBA could increase the consequences of
[[Page 15991]]
an analyzed accident. To ensure timely initiation, NNECO has
implemented a procedure for purging or venting through the SBGT
system which establishes a dedicated operator whose function at the
onset of a DBA is to isolate the train in use (the train expected to
be damaged by the pressure spike), verify the open AC [atmospheric
control] valves go closed, and then start the second train. This
procedure has been validated to ensure that these actions can be
completed within one minute.
Although not expected, a delay in operator action to initiate
the SBGT has been evaluated for impact upon the radiological
consequences. The evaluation shows that the offsite doses remain
well within the 10CFR100 limit even if the operator actions are not
completed until three minutes after the DBA occurs.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed change allows removal of the SBGT system from the
release path for normal containment purge and venting. The change
does not affect the frequency or requirement for venting. Nor does
the proposed LCO [limiting condition for operation] affect the
processes of venting or purging primary containment; the same
penetrations and containment isolation valves will continue to be
used. All purging and venting functions can still be performed when
required by existing specifications and plant procedures. The
proposed change does not diminish the capability of any isolation
valve for performing its isolation function.
Therefore, the proposed change can not create a new or different
kind of accident.
3. Involve a significant reduction in the margin of safety.
The affect of this change has been analyzed against the criteria
of 10CFR100 and 10CFR20. The potential release which may occur as a
result of a postulated DBA while purging or venting directly to the
stack will not exceed the limits of 10CFR100. Likewise, the
technical specifications and administrative controls established for
purging or venting through the SBGT minimize the potential for an
unfiltered release should a DBA occur during that evolution.
Further, the amount of time that a SBGT train is aligned to primary
containment is expected to be substantially reduced from that
required by the existing Technical Specification. Decreasing the
amount of time that SBGT is aligned to primary containment decreases
the possibility that a DBA would occur while in such an alignment.
Finally, the potential increase in dose which could occur as a
result of normal purge and vent activities will be controlled such
that it remains below acceptable limits.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360.
Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford,
CT 06141-0270.
NRC Project Director: Phillip F. McKee
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: January 17, 1996
Description of amendment requests: The proposed amendment would
revise selected technical specifications (TS) in accordance with the
NRC's Final Policy Statement on TS Improvements for Nuclear Power
Reactors and relocate the TS to the Diablo Canyon Power Plant Equipment
Control Guidelines. The proposed change would also create TS 6.8.4.j,
``Explosive Gas and Storage Tank Radioactivity Monitoring Program.''
Some of the TS would be relocated and maintained in accordance with
this program. Specifically, the following TS would be relocated: TS
3.1.2.1, ``Boration Systems Flow Path - Shutdown,'' TS 3.1.2.3,
``Charging Pumps - Shutdown,'' TS 3.1.2.4, ``Charging Pumps -
Operating,'' TS 3.1.2.5, ``Borated Water Sources - Shutdown,'' TS
3.1.2.6, ``Borated Water Sources - Operating,'' TS 3.3.3.2, ``Movable
Incore Detectors,'' TS 3.3.3.4, ``Meteorological Instrumentation,'' TS
3.3.3.10, ``Explosive Gas Effluent Monitoring Instrumentation,'' TS
3.9.3, ``Decay Time,'' TS 3.9.5, ``Communications,'' TS 3.9.6,
``Manipulator Crane,'' TS 3.9.7, ``Crane Travel - Fuel Handling
Building,'' TS 3.9.10.2, ``Water Level - Reactor Vessel - Control
Rods,'' TS 3.9.13, ``Spent Fuel Shipping Cask Movement,'' TS 3.10.1,
``Special Test Exceptions - Shutdown Margin,'' TS 3.10.4, ``Position
Indication System - Shutdown,'' TS 3.11.1.4, ``Liquid Holdup Tanks,''
TS 3.11.2.5, ``Explosive Gas Mixture,'' and TS 3.11.2.6, ``Gas Storage
Tanks.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes simplify the Technical Specifications (TS),
meet regulatory requirements for relocated TS, and implement the
recommendations of the Commission's Final Policy Statement on TS
Improvements and revised 10 CFR 50.36. Future changes to these
requirements will be controlled by 10 CFR 50.59. The proposed
changes are administrative in nature and do not involve any
modifications to any plant equipment or affect plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes are administrative in nature, do not
involve any physical alterations to any plant equipment, and cause
no change in the method by which any safety-related system performs
its function. Also, no changes to the operation of the plant or
equipment are involved.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes involve relocating TS requirements to a
licensee-controlled document. The requirements to be relocated were
identified by applying the criteria endorsed in the Commission's
Final Policy Statement, which is included in the new revision of 10
CFR 50.36, and are consistent with NUREG-1431, Rev. 1. Thus, the
proposed changes do not alter the basic regulatory requirements and
do not affect any safety analysis.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: William H. Bateman
[[Page 15992]]
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: February 12, 1996
Description of amendment request: The amendments would revise the
Susquehanna Units 1 and 2 Technical Specifications establish and
reference a Primary Containment Leakage Rate Testing Program in order
to implement 10 CFR 50, Appendix J, Option B in accordance with the
guidelines contained in Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Test Program'', dated September 1995.
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. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed license amendments do not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed license amendments revise the
Technical Specifications to reflect the adoption of a performance-
based containment leakage-testing program. The Nuclear Regulatory
Commission has approved the use of a performance-based option for
containment leakage testing programs when it amended 10 CFR Part 50,
Appendix J (60 FR 49495).
To adopt of (sic) the revised regulations, licensees are
required to incorporate into their Technical Specifications, by
general reference, the NRC regulatory guide or other plant specific
implementing document. A new Administrative Controls Specification
is being added to the Susquehanna SES Technical Specifications that
requires the establishment and maintenance of a Primary Containment
Leakage Rate Testing Program. As stated in the Technical
Specification, this Primary Containment Leakage Rate Testing Program
will conform with NRC Regulatory Guide 1.163, ``Performance-Based
Containment Leak-Rate Testing Program'', dated September 1995. The
Primary Containment Leakage Rate Testing Program establishes
requirements intended to ensure on-going containment integrity,
including the performance of a periodic general visual inspection of
the containment to detect early indications of structural
deterioration.
The effect of increasing containment leakage rate testing
intervals has been evaluated by the Nuclear Energy Institute using
the methodology described in NUREG-1493 and historical
representative industry leakage rate testing data. The results of
this evaluation, as published in NEI 94-01, Revision 0, are that the
increased risk corresponding to the extended test interval is small
(less than 0.1 percent of total risk) and compares well to the
guidance of the NRC's safety goal. The primary containment leak rate
data and component performance history at Susquehanna SES are
consistent with the conclusions reached in NUREG-1493 and NEI 94-01.
Therefore, adoption of performance-based verification of leakage
rates for isolation valves, containment penetrations, and the
overall containment boundary will provide an equivalent level of
safety and does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
No safety-related equipment, safety function, or plant
operations will be altered as a result of the proposed license
amendment.
The safety objective for the primary containment is stated in 10
CFR 50, Appendix A, ``General Design Criteria for Nuclear Power
Plants.'' The safety function of the primary containment will be met
since the containment will continue to provide ``an essentially leak
tight barrier against the uncontrolled release of radioactivity to
the environment...'' for postulated accidents. Therefore, the
proposed license amendments will not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
As stated above, the Nuclear Regulatory Commission has approved
the use of a performance-based option for containment leakage
testing programs when it amended 10 CFR Part 50, Appendix J (60 FR
49495). The new Primary Containment Leakage Rate Testing Program
will conform with NRC Regulatory Guide 1.163, Revision 0, dated
September 1995, ``Performance-Based Containment Leak-Rate Testing
Program'' by requiring that leakage testing intervals be established
based on the criteria in Section 11.0 of NEI 94-01, Revision 0.
As discussed in Part 1 above, the effect of increasing
containment leakage rate testing intervals has been evaluated by the
Nuclear Energy Institute using the methodology described in NUREG-
1493 and historical representative industry leakage rate testing
data. The results of this evaluation, as published in NEI 94-01,
Revision 0, are that the increased safety risk corresponding to the
extended test intervals is small (less than 0.1 percent of total
risk) and compares well to the guidance of the NRC's safety goal. In
addition, as demonstrated by risk analyses contained in NUREG-1482,
relaxation of the integrated leak rate test frequency does not
significantly increase the probability or consequences of a
previously evaluated accident. Integrated leakage rate tests have
been demonstrated to be of limited value in detecting significant
leakages from penetrations and isolation valves. The primary
containment leak rate data and component performance history at
Susquehanna SES are consistent with the conclusions reached in
NUREG-1493 and NEI 94-01. Therefore, the proposed license amendments
adopting a performance-based approach for verification of leakage
rates for isolation valves, containment penetrations, and the
containment overall will continue to meet the regulatory goal of
providing an essentially leak-tight containment boundary, will
provide an equivalent level of safety, and do not involve a
significant reduction in a margin of safety.
The revised Technical Specifications will continue to maintain
the allowable leak rate (La) as the Type A test performance
criterion. In addition, a requirement to perform a periodic general
visual inspection of the containment is part of the performance-
based leakage testing program.
The revised Technical Specifications will continue to maintain
the allowable leak rate (La) as the Type B and C tests' performance
criterion. As supported by the findings of NUREG-1493, the
percentage of leakages detected only by integrated leak rate tests
is small (only a few percent) and Type B and C leakage tests are
capable of detecting more than 97 percent of containment leakages
and virtually all such leakages are identified by local leak rate
tests (LLRTs) of containment isolation valves.
Thus, the proposed license amendments do not involve a
significant reduction in a margin of safety and will continue to
support the regulatory goal of ensuring an essentially leak-tight
containment boundary.
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
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: February 23, 1996
Description of amendment request: The proposed amendment would
change the Technical Specification (TS) Surveillance Requirement
4.6.2.1d concerning drywell-to suppression chamber bypass testing.
Currently, Susquehanna TSs require the performance of a bypass test at
40 plus or minus 10-month intervals. The proposed TS change would
request that the bypass test interval be revised to correspond with the
interval for Primary Containment Integrated Leak Rate
[[Page 15993]]
Testing (ILRT) under 10 CFR Part 50, Appendix J, Option B.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to allow bypass testing at the [Integrated
Leak Rate Testing] interval involves no physical or operational
changes to the Susquehanna SES. Reviews of bypass leakage test
results at Susquehanna and other similarly designed plants confirm
that minimal suppression pool bypass leakage has occurred. Based on
this data, the risk of suppression pool bypass leakage from non
vacuum breaker sources is no greater than that of other primary
containment passive structures which are tested at the ILRT
frequency. Leak testing of the drywell-to-suppression chamber vacuum
breakers will continue to be performed on a refueling and inspection
outage frequency to ensure that their contribution to the leakage
area is acceptable. In addition, inspection of the diaphragm slab
within the testing interval provides additional assurance that any
degradation to the structure will be detected and resolved.
Therefore, the pressure suppression capability of the containment is
not reduced from the existing design, and there will be no
significant increase in the probability or consequences of an
accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change to allow bypass testing at the ILRT interval
involves no physical or operational changes to the Susquehanna SES.
The surveillance change does not impact the LOCA response of the
units, or impact the design basis of the units in any way.
Therefore, the possibility of a new or different kind of accident
will not be created.
III. This change does not result in a significant reduction in a
margin of safety.
The drywell-to-suppression chamber bypass leak test data
obtained during previous testing at Susquehanna SES and other
similarly designed plants demonstrates conformance, by a large
margin, to the Technical Specification and design leakage
requirements. The test data and safety analysis provided here
indicate that there is negligible risk that the bypass leakage will
change adversely in future years. Furthermore, the proposed
performance based test methodology is judged to be acceptable based
on the small risk of bypass leakage through paths other than those
containing the suppression pool vacuum breakers. Testing of the
bypass leak pathway containing the vacuum breakers will be used to
verify acceptable bypass leakage during those outages when the
bypass leak test is not performed. In addition, periodic visual
inspection of the diaphragm slab within the bypass test interval
provides additional assurance that any degradation to the structure
will be detected and resolved.
Testing of the bypass leakage pathways containing vacuum
breakers, with stringent acceptance criteria, combined with the
other negligible potential leakage areas, and periodic inspection of
the diaphragm slab, provide an acceptable level of assurance that
the bypass leakage will be minimized. The proposed performance based
approach to bypass testing and inspection ensures that adverse
conditions can be detected and corrected such that the existing
level of confidence that the primary containment will function as
required during a LOCA is maintained. Therefore, the proposed
Technical Specification changes do 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: 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
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: February 29, 1996
Description of amendment request: The proposed amendment relocates
Technical Specification 3/4.9.6, ``Refueling Platform,'' to the
Technical Requirements Manual, which is controlled under the
requirements of 10 CFR 50.59.
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. Involves a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change relocates the provisions of the Refueling
Platform that are contained in the Technical Specifications and
places them in the Technical Requirements Manual. Review and
approval of those portions of the Refueling Platform requirements
contained in the Technical Requirements Manual and revisions thereto
will be the responsibility of the Plant Operations Review Committee
just as it was their responsibility to review changes to the
refueling platform Limiting Condition for Operation and Surveillance
Requirements when they were part of the Technical Specifications.
Requiring review by the Plant Operations Review Committee reinforces
the importance of the Technical Requirements manual and the
requirements controlled by it and assures a multidisciplined review.
Approved Technical Requirements or changes thereto are provided to
the Susquehanna Review Committee for information. No design basis
accidents are affected by the change, nor are safety systems
adversely affected by the change. Therefore, these changes will not
result in any change to current Technical Specification
requirements, but will reduce the level of regulatory control
associated with the identified requirements. The level regulatory
control has no impact on the probability or the consequences of an
accident previously evaluated, therefore, the proposed change will
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change relocates the provisions of the Refueling
Platform that are contained in the Technical Specifications and
places them in the Technical Requirements Manual. This change will
not involve any physical changes to the Refueling Platform and its
associated instrumentation nor any changes in the manner in which
this equipment is operated, maintained, tested or inspected. Future
changes to these relocated requirements or surveillances will be
evaluated in accordance with the requirements of 10CFR50.59.
Therefore, this change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The margin of safety is not reduced. The relocated requirements
do not meet any of the four criteria in the NRC Policy Statement
used for defining the scope of Technical Specifications. In
addition, the relocated requirements and surveillances for the
refuel platform and associated instrumentation remain the same as
stated in the existing Technical Specifications. Future changes to
these relocated requirements or surveillances will be evaluated in
accordance with the requirements of 10CFR50.59. Review and approval
of those portions of the Refueling Platform requirements contained
in the Technical Requirements Manual and the revisions thereto will
be the responsibility of the Plant Operations Review Committee just
as it was their responsibility to review changes to the refueling
platform Limiting Condition for Operation and Surveillance
Requirements when they were part of the Technical Specifications.
Approved Technical Requirements or changes thereto are provided to
the Susquehanna Review Committee for information. Therefore, no
[[Page 15994]]
significant reduction in a margin to safety is proposed.
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
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: February 29, 1996
Description of amendment request: The proposed amendment removes
the Rod Block Monitor (RBM) requirements from the Technical
Specifications, thereby reducing the number of rod movements during
power maneuvers.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change removes the Rod Block Monitor requirements
from Technical Specifications based on no credit being taken for the
RBM in the reload licensing analysis. The RBM was originally
designed to prevent fuel damage during the Rod Withdrawal Error
[RWE] event by automatically stopping control rod motion before any
fuel design limits are exceeded. However, due to control rod drift
events in which the RBM can not (sic) stop control rod motion, the
RWE is analyzed without taking credit for the RBM. The results of
this analysis are operating limits that prevent fuel damage from a
RWE in which control rod motion is not stopped by the RBM.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
This proposed change of removing the RBM requirements from
Technical Specifications does not change the currently approved
approach for performing the reload licensing analysis for either
Unit. To date all reload analyses have been performed considering
the rod drift event as a moderate frequency event and no credit
being taken for the RBM. Since no credit is taken, removal of these
requirements from Technical Specifications does not impact the
current approach for performing reload analysis. Therefore, this
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Continued compliance to the governing General Design Criteria
[GDC] for the RWE analysis assumes an appropriate margin of safety.
GDC 10 is met when the specified acceptable fuel design limits
(SAFDLs) are not exceeded for the RWE. The first SAFDL requires that
a MCPR [Minimum Critical Power Ratio] Operating Limit be determined
such that the reduction of MCPR margin due to an RWE does not
violate the MCPR Safety Limit. The second SAFDL requires that the
uniform cladding strain does not exceed 1% during an RWE. PP&L's
[Pennsylvania Power and Light Company] licensing analysis of the
RWE, without taking credit for the RBM, determines a MCPR Operating
limit such that the reduction of MCPR margin due to an RWE does not
violate the MCPR Safety Limit and validates that the maximum uniform
cladding strain is less than 1%. Therefore, the applicable SAFDLs
for the RWE are satisfied and the GDC requirements met.
GDC 20 is met when the reactivity control system is
automatically actuated to prevent exceeding the SAFDLs. PP&L's
licensing analysis of the RWE, without taking credit for the RBM,
conservatively determines a MCPR Operating Limit and validates that
the maximum uniform cladding strain is less than 1%. Therefore,
actuation of the RBM is not necessary to prevent exceeding the
applicable SAFDLs for the RWE.
GDC 25 is met when a single malfunction in the reactivity
control system will not cause the SAFDLs to be exceeded. The current
RWE licensing analysis assumes a control rod drift event without any
credit for the RBM. With respect to the reactivity control system,
the assumptions of a control rod drift event and no actuation of the
RBM are more conservative than the assumptions in the original SSES
Safety Evaluation. Therefore, the requirements from GDC 25 are still
met. Therefore, no significant reduction in the safety margin
exists.
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
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of amendment request: February 9, 1996, as supplemented March
15, 1996. This notice supersedes the notice published on February 28,
1996 (61 FR 7557) in its entirety.
Description of amendment request: The proposed amendment would
revise the Administrative Controls Section 5.6.6 of the Ginna Technical
Specifications (TSs) to incorporate a reference to the methodology for
determining pressure/temperature (P/T) and low-temperature overpressure
protection (LTOP) limits. The proposed amendment would follow guidance
given in Generic Letter 96-03 for relocating LTOP and the reactor
coolant system (RCS) P/T limits to the RCS Pressure and Temperature
Limits Report (PTLR). The proposed amendment will allow the licensee to
perform future LTOP and RCS P/T evaluations, using NRC-approved
methodology, without requiring changes to the TS.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant increase in the probability
or consequences of an accident previously evaluated. The proposed
changes only require that future RCS P/T and LTOP limits be
developed using NRC approved methodology as specified within the
Administrative Controls section and do not involve any technical
changes. As such, these changes are administrative in nature and do
not impact initiators or analyzed events or assumed mitigation of
accident or transient events. Therefore, these changes do not
involve a significant increase in the probability or consequences of
an accident previously analyzed.
2. Operation of Ginna Station in accordance with the proposed
changes does not create the possibility of a new or different kind
of accident from any accident previously evaluated. The proposed
changes do not involve a physical alteration of the plant (i.e., no
new or different type of equipment will be installed) or changes in
the methods governing normal plant operation. The proposed changes
will not impose any new or different requirements. Thus, this change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
[[Page 15995]]
3. Operation of Ginna Station in accordance with the proposed
changes does not involve a significant reduction in a margin of
safety. The proposed changes will not reduce a margin of plant
safety because the changes do not impact any safety analysis
assumptions other than requiring future evaluations of RCS P/T and
LTOP limits to be performed in accordance with NRC approved
methodology. These changes are administrative in nature. As such, no
question of safety is involved, and the change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Rochester Public Library, 115
South Avenue, Rochester, New York 14610
Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400
L Street, NW., Washington, DC 20005 NRC Acting Project Director: Susan
Frant Shankman
South Carolina Electric & Gas Company (SCE&G), South Carolina
Public Service Authority, Docket No. 50-395, Virgil C. Summer
Nuclear Station, Unit No. 1, Fairfield County, South Carolina
Date of amendment request: March 19, 1996
Description of amendment request: The licensee is proposing to
change the Technical Specification (TS) 3/4.2.4, QUADRANT POWER TILT
RATIO (QPTR), the Bases for QPTR, and TS 3/4.3.1, REACTOR TRIP SYSTEM
INSTRUMENTATION, Table 3.3-1, ``Table Notation, Action Statement 2.c.''
The licensee is requesting the changes in order to use the guidance in
the improved Westinghouse Standardized Technical Specifications, NUREG
1431, Rev. 1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The probability or consequences of an accident previously
evaluated in the FSAR is not significantly increased.
The QPTR limits ensure that FNdelta-H and FQ(z)
remain below their limiting values by preventing an undetected
change in the gross radial power distribution. In MODE 1, the
FNdelta-H and FQ(z) limits must be maintained to
preclude core power distributions from exceeding design limits
assumed in the safety analyses. The QPTR satisfies Criterion 2 of
the NRC Policy Statement.
The QPTR limit of 1.02, at which corrective action is required,
provides a margin of protection for both the departure from nucleate
boiling ratio and linear heat generation rate contributing to
excessive power peaks resulting from X-Y plane power tilts. A
limiting QPTR of 1.02 can be tolerated before the margin for
uncertainty in FQ(z) and FNdelta-H is possibly
challenged. With the QPTR exceeding its limit, a power level
reduction of 3% from RATED THERMAL POWER for each 1% by which the
QPTR exceeds 1.00 is a conservative tradeoff of total core power
with peak linear power.
The Power Range Neutron Flux trip setpoint reduction is not
required since incore flux measurements are not expected to change
concurrent with the loss of a Power Range Channel. These setpoints,
which were previously reduced in order to account for uncertainties,
will now be monitored and corrected, if necessary, per TS 3.2.4.
Any change in the QPTR would be detected by requiring a check of
the QPTR once per 12 hours. If the QPTR indicates an increase,
THERMAL POWER has to be reduced accordingly. A 12 hour completion
time is sufficient because any additional change in QPTR would be
relatively slow.
The improvement of TS 3/4.2.4 to reflect the improved STS in no
way impacts the accident analysis of the FSAR. Therefore, the
probability or consequences of a previously evaluated accident has
not been increased.
2. The possibility of an accident or a malfunction of a
different type than any previously evaluated is not created.
The proposed amendment request does not necessitate physical
alteration of the plant nor changes in parameters governing normal
plant operation. Therefore, the change does not create the
possibility of a new or different kind of accident or malfunction.
3. The margin of safety has not been significantly reduced.
This proposed amendment request precludes core power
distributions that may lead to violation of the following fuel
design criteria:
a. During a large break loss of coolant accident, the peak
cladding temperature must not exceed 2200 deg.
b. During a loss of forced reactor coolant flow accident, there
must be at least 95% probability at the 95% confidence level (the
95/95 departure from nucleate boiling (DNB) criterion) that the hot
fuel rod in the core does not experience a DNB condition;
c. During an ejected rod accident, the energy deposition to the
fuel must not exceed 280 cal/gm; and
d. The control rods must be capable of shutting down the reactor
with a minimum required shutdown margin with the highest worth
control rod stuck fully withdrawn.
The improvement of TS 3/4.2.4 ensures that the gross radial
power distribution remains consistent with the design values used in
the safety analyses.
The core peaking factors and the quadrant tilt must be evaluated
because they are the factors that best characterize the core power
distribution. This reevaluation is required to ensure that the
reactor core conditions are consistent with the assumptions in the
safety analyses. Therefore, the margin of safety has not decreased.
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: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Attorney for licensee: Randolph R. Mahan, South Carolina Electric &
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
NRC Project Director: Frederick J. Hebdon
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: November 2, 1995
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 3.8.1, ``AC Sources - Operating,'' of the
improved TS, to (1) extend the offsite circuit allowed outage time
(AOT) from ``72 hours AND 6 days from discovery of failure to meet
LCO'' to ``72 hours AND 10 days from discovery of failure to meet LCO''
and (2) extend the emergency diesel generator (EDG) AOT from ``72 hours
AND 6 days from discovery of failure to meet LCO'' to ``7 days AND 10
days from discovery of failure to meet LCO.'' Additionally, the
licensee proposes to further extend the EDG AOT to ``10 days AND 10
days from discovery of failure to meet LCO'' on a once-per-refueling
cycle frequency for maintenance purposes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Emergency Diesel Generators (EDGs) are backup alternating
current power sources designed to power essential safety systems in
the event of a loss of offsite power. EDGs are not accident
initiators in any accident previously evaluated. Therefore, this
change does not involve an increase in the probability of an
accident previously evaluated.
The EDGs provide backup power to components that mitigate the
consequences of accidents. The proposed changes to the Allowed
Outage Times (AOTs) do not affect
[[Page 15996]]
any of the assumptions used in the deterministic safety analysis.
To fully evaluate the effect of the EDG AOT extension,
Probabilistic Safety Analysis (PSA) methods were utilized. The
results of these analyses show no significant increase in the core
damage frequency. As a result, there would be no significant
increase in the consequences of accidents previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change does not alter the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the Limiting Conditions for
Operation or their Bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes, and these evaluations determined that the
changes are either risk neutral or risk beneficial.
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 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: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P.O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: November 6, 1995
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 3.5.1, ``Safety Injection Tanks (SITs),''
of the improved TS to extend, in general, the allowed outage time (AOT)
for a single inoperable SIT from 1 hour to 24 hours. Additionally, the
licensee proposes to extend the SIT AOT from 1 hour to 72 hours if a
single SIT becomes inoperable due to malfunctioning SIT water level
and/or nitrogen cover pressure instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Safety Injection Tanks (SITs) are passive components in the
Emergency Core Cooling System (ECCS). The SITs are not accident
initiators in any accident previously evaluated.
Therefore, this change does not involve an increase in the
probability of an accident previously evaluated.
The SITs are designed to mitigate the consequences of Loss of
Coolant Accidents (LOCAs). The proposed changes do not affect any of
the assumptions used in deterministic LOCA analysis. Therefore, the
consequences of accidents previously evaluated do not change.
To fully evaluate the SIT Allowed Outage Time (AOT) extension,
Probabilistic Safety Analysis (PSA) methods were utilized. The
results of these analyses show no significant increase in core
damage frequency. As a result, there would be no significant
increase in the consequences of an accident previously evaluated.
The proposed change pertaining to SIT inoperability based solely
on instrumentation malfunction does not involve a significant
increase in the consequences of an accident as evaluated and
endorsed by the Nuclear Regulatory Commission (NRC) in NUREG-1366,
``Improvements to Technical Specifications Surveillance
Requirements.''
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
This proposed change does not change the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not affect the limiting conditions for
operation or their bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes. These evaluations demonstrate that the
changes are either risk neutral or risk beneficial.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: November 8, 1995
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 3.5.2, ``ECCS - Operating,'' in the
improved TS to extend the allowed outage time from 72 hours to 7 days
for a single low pressure safety injection (LPSI) train.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The Low Pressure Safety Injection (LPSI) system is a part of the
Emergency Core Cooling System (ECCS). Inoperable LPSI components are
not considered to be accident initiators. Therefore, this change
does not involve an increase in the probability of an accident
previously evaluated.
The LPSI system is primarily designed to mitigate the
consequences of a large Loss of Coolant Accident (LOCA). This
proposed change does not affect any of the assumptions used in the
deterministic LOCA analysis. Therefore, the consequences of
accidents previously evaluated do not change.
To fully evaluate the LPSI Allowed Outage Time (AOT) extension,
Probabilistic Safety Analysis (PSA) methods were utilized. The
results of these analyses show no significant increase in core
damage frequency. As a result, there would be no significant
increase in the consequences of an accident previously evaluated.
Therefore, this change does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
[[Page 15997]]
This proposed change does not change the design, configuration,
or method of operation of the plant. Therefore, this change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change does not affect the limiting conditions for
operation or their bases that are used in the deterministic analyses
to establish the margin of safety. PSA evaluations were used to
evaluate these changes.
Therefore, this change does not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: December 6, 1995
Description of amendment requests: The licensee proposes to revise
Technical Specification (TS) 4.3, ``Fuel Storage,'' of the improved TS,
to allow fuel assemblies having a maximum U-235 enrichment of 4.8
weight percent to be stored in both the spent fuel racks and the new
fuel racks.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
There is no increase in the probability of an accident because
the physical characteristics of a fuel assembly are not changed when
fuel enrichment is increased. No changes will be made to any safety
related equipment or systems. Fuel assembly movement will continue
to be controlled by approved fuel handling procedures.
Fuel cycle designs will continue to be analyzed with Nuclear
Regulatory Commission (NRC)-approved codes and methods to ensure the
design bases for San Onofre Units 2 and 3 are satisfied.
The double contingency principle of American National Standards
Institute/American Nuclear Society (ANSI/ANS) Standard 8.1-1983 can
be applied to any postulated accident in the Spent Fuel Pool (SFP)
which could cause reactivity to increase. In conjunction with
administrative controls for heavy loads and impact zones, a boron
concentration of 1850 parts per million (PPM) (the current Technical
Specification (TS) limit) is sufficient to maintain k-eff less than
or equal to 0.95 for all normal and postulated accident conditions.
Regarding the new fuel storage racks, there is no postulated
accident which could cause reactivity to increase above 0.95 for all
moderator densities from 0.0 to 1.0 grams/cubic centimeter (gms/cc).
The radiological consequence analyses performed in the Updated
Final Safety Analysis Report (UFSAR) include the development of
source terms which bound discharge fuel burnups to 60,000 megawatt
days per ton (MWD/T). Increasing the San Onofre Units 2 and 3
enrichment to 4.8 weight percent (w/o) does not result in discharge
fuel assembly burnups greater than 60,000 MWD/T. Thus, the
consequences of the fuel handling accident are unchanged from the
current UFSAR bases.
Therefore, this proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve any physical changes to the
plant or any changes to the method in which the plant is operated.
They do not affect the performance or qualification of safety
related equipment. Fuel handling accidents were previously
considered. Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
3. The proposed change does not involve a significant reduction
in a margin of safety.
For the SFP, the NRC acceptance criteria is k-eff less than or
equal to 0.95 under all normal and accident conditions and including
uncertainties. For the new fuel storage racks, k-eff must remain
less than 0.95 if completely flooded with unborated water, and must
remain below 0.98 in an optimum moderation event. Analyses have been
performed which demonstrate that these acceptance criteria will
continue to be met when the enrichment is increased to 4.8 w/o.
The current UFSAR design bases SFP decay heat loads bound the
proposed enrichment increase due to the reduced fuel batch size.
Radiological effects of fuel handling accidents are unchanged by
this enrichment increase.
The proposed design of the higher enriched fuel will result in a
slight weight increase. However, the seismic event is bounded by the
analyses performed for the rerack project.
Therefore, there will not be a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of amendment requests: January 4, 1996
Description of amendment requests: The licensee proposes to delete
License Conditions 2.C(26) and 2.C(27). These license conditions
require the licensee to implement and maintain a plan for scheduling
all capital modifications based on an NRC approved Integrated
Implementation Schedule Program Plan.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change deletes an administrative means of tracking
and scheduling NRC required plant modifications and license
commitments. It does not affect the plant configuration nor NRC
mandated schedules for implementation of modifications. Because the
deletion of the license condition does not affect the plant
configuration, no accident analyses are affected; therefore, the
proposed change does not increase the probability or consequences of
any previously evaluated accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change will not alter the configuration of the
plant or its operation; therefore, the proposed change does not
create a new or different kind of accident from any previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change is administrative and does not affect any
accident analyses or
[[Page 15998]]
involve any modification to the plant configuration; therefore, the
proposed change does not involve a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Attorney for licensee: T. E. Oubre, Esquire, Southern California
Edison Company, P. O. Box 800, Rosemead, California 91770
NRC Project Director: William H. Bateman
Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear
Plant, Unit 1, Rhea County, Tennessee
Date of amendment request: February 28, 1996
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to extend the ice weighing and
flow channel inspection surveillance frequencies from 9 to 18 months.
Concurrently, the required total ice bed weight would be increased from
2,360,875 to 2,403,800 lbs. to account for the anticipated additional
ice sublimation during the longer interval between weighing and
inspection.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's analysis is
presented below.
1. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The ice condenser system is provided to absorb thermal energy
release following a LOCA or high energy line break (HELB) and to limit
the peak pressure inside containment. The containment analysis for
Watts Bar is based on a minimum of 1093 lbs of ice per ice basket
evenly distributed throughout the ice condenser, and the subcompartment
analysis is based on 85 percent of the available flow area (flow
channels) being open uniformly throughout the ice condenser. For the
predicted sublimation rate of up to 12 percent for 18 months, an
average ice basket weight of 1093 lbs at the end of the 18 month period
would still be available. An evaluation of the operating history of the
other operating ice condenser plants shows that after 18 months 85
percent of the flow channels will still be available.
Thus the ice condenser will perform its design functions with the
revised minimum ice weight and inspection interval. There will be no
design change or other operational changes. Accordingly, the proposed
changes to the technical specifications do not affect the probability
or consequences of an accident.
2.
The changes do not create the possibility of a new or different
kind of accident from any previously analyzed.
As stated above, the proposed changes do not involve modifications
to the ice condenser or other plant systems. Hence there is no
possibility of a new or different kind of accident since no new design
is involved.
3. The changes do not involve a significant reduction in a margin
of safety.
Plant safety margins are established through limiting conditions of
operation, limiting safety system settings, and safety limits specified
in the TS. None of these will be changed.
Based on this analysis, 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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-390 Watts Bar Nuclear
Plant,Unit 1, Rhea County, Tennessee
Date of amendment request: February 28, 1996
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) surveillance frequency for
Westinghouse type AR relays, used as solid state protection system
slave relays or auxiliary relays, from quarterly to a refueling outage
frequency.
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) Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
This change to the Technical Specifications does not result in a
condition where the design, material, and construction standards
that were applicable prior to the change are altered. The same ESFAS
instrumentation is being used and the same ESFAS system reliability
is expected. The proposed change will not modify any system
interface or function and could not increase the likelihood of an
accident since these events are independent of this change. The
proposed activity will not change, degrade or prevent the
performance of any accident mitigation systems or alter any
assumptions previously made in evaluating the radiological
consequences of an accident described in the safety analysis report.
Therefore, the proposed amendment does not result in any increase in
the probability or consequences of an accident previously evaluated.
(2) Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any previously evaluated.
This change does not alter the performance of the ESFAS
mitigation systems assumed in the plant safety analysis. Changing
the interval for periodically verifying ESFAS slave relays (assuring
equipment operability) will not create any new accident initiators
or scenarios. Implementation of the proposed amendment does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendment would not involve a significant reduction in a margin of
safety.
This change does not affect the total ESFAS system response
assumed in the safety analysis. The periodic slave relay functional
verification is relaxed because of the demonstrated high reliability
of the relay and its insensitivity to any short term wear or aging
effects. Implementation of the proposed amendment does not result in
a reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
thisreview, 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: Chattanooga-Hamilton County
Library, 1001 Broad Street, Chattanooga, TN 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
[[Page 15999]]
NRC Project Director: Frederick J. Hebdon
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: March 12, 1996
Brief description of amendments: The proposed amendments would
revise Technical Specification (TS) 3/4.6.1.1, ``Containment
Integrity,'' 3/4.6.1.2, ``Containment Leakage,'' 3/4.6.1.3,
``Containment Air Locks,'' and 3/4.6.1.6, ``Containment Structural
Integrity,'' and add new TS 6.8.3g, ``Containment Leakage Rate Testing
Program,'' to implement the new performance-based leakage rate testing
program as permitted by 10 CFR Part 50, Appendix J.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes to the TS and the addition of specification
6.8.3g to implement the new performance based Containment Leakage
Rate Testing Program, have no effect on plant operation. The
proposed changes only provide mechanisms within the TS for
implementing a performance based methodology for determining the
frequency of leak rate testing which has been approved by the
Commission. The test type and test method used for testing would not
be changed. The test acceptance criteria would not be changed and
containment leakage will continue to be maintained within the
required limits.
Directly referencing the Containment Leakage Rate Testing
Program for containment [integrated leak rate test] ILRT and [local
leak rate test] LLRT requirements does not involve any modification
to plant equipment or affect the operation or design basis of the
containment. Leakage rate testing is not a precursor to or an
initiating event for any accident.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
The proposed changes only allow for the implementation of Option
B testing frequencies and do not involve any modifications to any
plant equipment or affect the operation or design basis of the
containment. The proposed changes do not affect the response of the
containment during a design basis accident.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
The proposed changes do not adversely affect a Safety Limit,
Limiting Condition for Operation (LCO) or plant operations. These
changes only implement the allowed Option B testing frequencies that
have been determined by the Commission not to involve a safety
concern. The testing method, acceptance criteria and bases are not
changed and still provide assurance that the containment will
provide its intended function.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, N.W., Washington, DC 20036
NRC Project Director: William D. Beckner
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: March 21, 1996
Description of amendment request: The proposed changes clarify the
requirements for testing the charcoal adsorbent in the auxiliary
ventilation and control room air filtration systems as outlined in
Technical Specifications 4.12 and 4.20, respectively.
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 probability or consequences of an accident previously
evaluated is not significantly increased.
The charcoal testing clarifications and explict reference to the
testing currently conducted do not affect system operation or
performance, nor do they affect the probability of any event
initiators. The changes do not affect any Engineered Safety Features
actuation setpoints or accident mitigation capabilities. Therefore,
the proposed changes do not significantly increase the consequences
of an accident or malfunction of equipment important to safety
previously evaluated in the UFSAR [Updated Final Safety Analysis
Report].
2. The possibility of an accident or a malfunction of a different
type than any previously evaluated is not created.
The clarification to the charcoal sample testing protocol does
not affect the method of operation of the system. The proposed
changes clarify and explicitly identify the testing methodology for
the charcoal samples. No new or different accident scenarios,
transient precursors, failure mechanisms, or limiting single
failures are introduced as a result of these changes. Therefore, the
possibility of a new or different kind of accident other than those
already evaluated is not created by this change.
3. The margin of safety has not been significantly reduced.
The charcoal adsorber sample laboratory testing accurately
demonstrates the required performance of the adsorbers following a
design basis LOCA [loss-of-coolant accident] or Fuel Handling Accident.
Changing the Technical Specifications to clarify the actual test
methodology and explicitly [referencing] the charcoal testing actually
performed does not affect system performance or operation. Therefore,
these changes do not result in a significant reduction in any margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Eugene V. Imbro
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: February 19, 1996
Description of amendment request: The proposed amendment would
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS) Section 4.2 and its associated basis by allowing the application
of a voltage-based repair limit for the steam generator tube support
plate intersections experiencing
[[Page 16000]]
outside diameter stress corrosion cracking. The proposed repair
criteria are based on guidance provided in Generic Letter 95-05,
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes
affected by Outside Diameter Stress Corrosion Cracking,'' dated August
14, 1995, and on associated industry guidance.
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 change was reviewed in accordance with the
provisions of 10 CFR 50.92 to show no significant hazards exist. The
proposed change will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Testing of model boiler specimens for free span tubing (no TSP
[tube support plate] restraint) at room temperature conditions show
burst pressures in excess of 5,000 psig for indications of ODSCC
[outside diameter stress corrosion cracking] with voltage
measurements as high as 19 volts. Burst testing performed on five
intersections pulled from the Kewaunee SGs [steam generators] with
up to a 2 volt indication showed measured tube burst in the range of
9,537 to 9,756 psig. Burst testing performed on pulled tubes from
other plants with up to 7.5 volt indications show burst pressures in
excess of 6,300 psi at room temperatures. Correcting for the effects
of temperature on material properties and the minimum strength
levels, tube burst capability significantly exceeds the safety
factor requirements of RG [Regulatory Guide] 1.121.
Tube burst criteria are inherently satisfied during normal
operating conditions due to the presence of the TSP. Test data
indicates that tube burst cannot occur within the TSP, even for
tubes with through wall EDM [electro-discharge machining] notches
0.75 inch long, when the notch is adjacent to the TSP. Since tube
burst is precluded during normal operating conditions, the criterion
that must be satisfied to demonstrate adequate tube integrity is a
safety margin of 1.43 times MSLB [main steam line break] pressure
differential. The BOC [beginning of cycle] structural limit for 7/8
inch diameter tubing is 8.82 volts. Applying an allowance of 20.5%
for NDE [nondestructive examination] uncertainty and 50% for crack
growth rate over an operating cycle results in a voltage repair
limit of 5.4 volts. The proposed repair limit of 2 volts is very
conservative when compared to the 5.4 volts taking into account the
low average growth rates experienced at Kewaunee and the high tube
burst pressures.
Relative to the expected leakage during accident condition
loadings, a plant specific calculation was performed to determine
the maximum primary-to-secondary leakage during a postulated MSLB
event. The evaluation considered both pre-accident and accident
initiated iodine spikes. The results of the evaluation show that the
accident spike yielded the limiting leak rate. This case was based
on a 30 rem thyroid dose at the site boundary and initial primary
and secondary coolant activity levels of 1.0 uCi/gm and 0.1 uCi/gm
dose equivalent iodine-131, respectively. A leak rate of 34.0 gpm
was determined to be the upper limit for allowable primary to
secondary leakage in the SG in the faulted loop. The SG in the
intact loop was assumed to leak at a rate of 0.1 gpm (150 gpd).
Application of the voltage-based repair limit will be
supplemented with a projected EOC [end of cycle] MSLB leakage
calculation and conditional burst probability assessment. The
methodology for performing these calculations will be in accordance
with the GL [generic letter]. Should the projected MSLB leakage be
exceeded indications will be repaired or removed from service until
the projected leakage is less than or equal to 34.0 gpm.
Application of the voltage-based repair limit will not adversely
affect SG tube integrity. Therefore, the proposed amendment will not
increase the probability or consequences of an accident previously
evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
Implementation of the proposed voltage-based repair limit will
not reduce the overall safety or functional requirements of the SG
tube bundles. The tube burst criteria will be satisfied during
normal operating conditions by the presence of the TSPs. The RG
1.121 criteria that must be satisfied during accident loading
conditions is 1.43 times MSLB differential pressure. Conservatively,
the existing data base of burst testing shows that the tube burst
margins can be satisfied with bobbin coil signal amplitudes of about
8.82 volts or less regardless of the depth of tube wall penetration.
The proposed repair criteria will be supplemented with a reduced
operating leakage requirement of 150 gpd through either SG to
preclude the potential for excessive leakage during operating
conditions. The 150 gpd restriction will provide for timely leakage
detection and plant shutdown in the event of the occurrence of an
unexpected single crack resulting in leakage that is associated with
the longest permissible crack length. The operating leakage limit is
based on leak-before break considerations, critical crack length and
predicted leakage.
The SG tube integrity will continue to be maintained through
inservice inspections and primary-to-secondary leakage monitoring.
Therefore, the proposed change will not create the possibility of a
new or different kind or accident.
3. Involve a significant reduction in the margin of safety.
Application of the voltage-based repair criteria has been
demonstrated to maintain tube integrity commensurate with the RG
1.121 criteria. RG 1.121 describes a method acceptable to the staff
for meeting GDCs [general design criteria] 2, 14, 15, 31 and 32.
This is accomplished by determining the limiting degradation of SG
tubing as established by inservice inspection, beyond which tubes
should be removed from service. Upon implementation of the repair
criteria, even under the worst case conditions, the occurrence of
ODSCC at the TSPs is not expected to lead to a SG tube rupture event
during normal or faulted conditions. The most limiting event would
be a potential increase in leakage during a MSLB event. Excessive
leakage during a MSLB is precluded by verifying that the expected
EOC crack distribution of ODSCC indications at TSP locations would
result in an acceptably low primary-to-secondary leakage. Therefore,
the radiological consequences from tubes remaining in service is a
small fraction of the 10 CFR 100 limits.
The combined effects of a LOCA [loss-of-coolant accident] plus
SSE [safe shutdown earthquake] on the SGs were assessed as required
by GDC 2. This issue was addressed for the Kewaunee SGs through the
application of leak-before-break (LBB) principles to the primary
loop piping. Based on the results of this analysis, it is concluded
that the LBB is applicable to the Kewaunee primary loops and, thus,
the probability of breaks in the primary loop piping is sufficiently
low that they need not be considered in the structural design basis
of the plant. Excluding breaks in the primary loops, the LOCA loads
from the large branch lines were also assessed and found to be of
insufficient magnitude to result in SG tube collapse. Based on these
analysis results, no tubes are expected to collapse or deform to the
degree that the secondary-to-primary in-leakage would be increased
over currently expected levels. On this basis no tubes need to be
excluded from the voltage-based repair criteria for reasons of
deformation resulting from combined LOCA or SSE loadings.
Addressing the RG 1.83 considerations, implementation of the
voltage-based repair criteria will include a 100% bobbin coil probe
inspection of all tube-to-TSP intersections with known ODSCC down to
the lowest cold leg TSP identified. This will be supplemented by a
reduced operating leakage limit, enhanced eddy current data analysis
guidelines, MRPC [motorized rotating pancake coil] inspection
requirements and a projected EOC voltage distribution. It is
concluded that the proposed change will 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of Wisconsin,
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner,
P. O. Box 1497, Madison, Wisconsin 53701-1497
NRC Project Director: Gail H. Marcus
[[Page 16001]]
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: January 16, 1996
Brief description of amendments: The amendments revise the
Technical Specifications to reflect approval of the use of 10 CFR Part
50, Appendix J, Option B, for the Calvert Cliffs Nuclear Power Plant,
Unit Nos. 1 and 2, containment leakage rate test program for Type A
tests only.
Date of issuance: March 13, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 212 and 189
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 14, 1996 (61
FR 5810) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated March 13, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2 Calvert
County, Maryland
Date of application for amendments: November 30, 1995, as
supplemented by letter dated March 15, 1996.
Brief description of amendments: The amendments allow the
installation of tube sleeves as an alternative to plugging for repair
of steam generator (SG) tubes using repair techniques developed by
Westinghouse Electric Corporation. The November 30, 1995, letter also
requested approval of repair techniques developed by ABB Combustion
Engineering, Inc., for repairing SG tubes. The NRC staff is still
reviewing that portion of the request and will notice the results of
its review at a future date.
Date of issuance: March 22, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 213 and 190
Facility Operating License No. DPR-53 and DPR-69: Amendment revised
the Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
176) The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated March 22, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson
Steam Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of application for amendment: November 22, 1995
Brief description of amendment: The proposed change will delete the
qualifying statement, ''... provided the remaining systems are in
continuous operation,'' from TS Section 3.3.4.2.
Date of issuance: March 15, 1996
Effective date: March 15, 1996
Amendment No. 168
Facility Operating License No. DPR-23. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 6, l995 (60 FR
62487) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 15, 1996No significant
hazards consideration comments received: No
Local Public Document Room location: Hartsville Memorial Library,
147 West College Avenue, Hartsville, South Carolina 29550
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: July 18, 1994, as supplemented
by letters dated October 9, 1995, February 13 and March 8, 1996
Brief description of amendments: The amendments revise the current
combined Technical Specifications (TS) for Units 1 and 2 by separating
them into individual volumes for Unit 1 and Unit 2. In addition to the
changes required by the TS split, some administrative and editorial
changes were made, such as the correction of typographical errors and
the deletion of unnecessary blank pages.
Date of issuance: March 21, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: Unit 1 - 166 - Unit 2 - 148
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47166) The October 9, 1995, February 13 and March 8, 1996, letters
provided additional information that did not change the scope of the
July 18, 1994, application and the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
March 21, 1996 and Environmental Assessment dated February 7, 1996. No
significant hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
[[Page 16002]]
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of application for amendments: December 15, 1995, as
supplemented March 5, 1996
Brief description of amendments: These amendments (1) revise
Technical Specifications (TSs) 3/4.6.1.1, 3/4.6.1.2, 3/4.6.1.3, 3/
4.6.1.6, and associated Bases, (2) delete TS 6.9.2.g, and (3) add a new
TS 6.17. These changes make the TSs consistent with Option B of
Appendix J of 10 CFR Part 50 and the implementing guidance of
Regulatory Guide 1.163, ``Performance-Based Containment Leak Test
Program,'' dated September 1995. Option B of Appendix J permits
implementation of a performance-based leak rate test schedule in lieu
of the prescriptive requirements contained in Option A of Appendix J.
These amendments remove from the TSs the prescriptive requirements of
Option A concerning test frequencies and test methodology. These
amendments also include minor administrative and editorial changes to
add consistency between the Bases and the TSs and provide additional
clarification.
Date of issuance: March 19, 1996
Effective date: Both units, as of the date of issuance, to be
implemented within 60 days.
Amendment Nos.: 197 and 80
Facility Operating License Nos. DPR-66 and NPF-73: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
179) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 19, 1996.No significant hazards
consideration comments received: No.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
Florida Power and Light Company, et al., Docket No. 50-389, St.
Lucie Plant, Unit No. 2, St. Lucie County, Florida
Date of application for amendment: August 16, 1995
Brief description of amendment: This amendment modifies Technical
Specification 3.6.6.1, Shield Building Ventilation System (SBVS), to
more effectively address the design functions performed by the SBVS for
both the Shield Building and the Fuel Handling Building.
Date of issuance: March 20, 1996
Effective date: March 20, 1996
Amendment No.: 81
Facility Operating License No. NPF-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 27, 1995 (60
FR 49937) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 20, 1996. No significant
hazards consideration comments received: No.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mill
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: August 10, 1995, as supplemented
on December 21, 1995, and February 22, 1996
Brief description of amendment: The amendment deletes a Technical
Specification (TS) reference to the reactor trip input to the reactor
building isolation system, changes the surveillance frequency for the
sodium hydroxide storage tank and station battery, and removes an
inappropriate reference in the TS bases section to testing that is not
required by the TSs themselves.
Date of issuance: March 21, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 200
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 27, 1995 (60
FR 58401). The December 21, 1995 and February 22, 1996 letters did not
change the staff's determination hazards consideration exist. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated March 21, 1996No significant hazards
consideration comments received: No
Local Public Document Room location: Law/Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: December 19, 1995 and
supplemented February 16, 1996 (AEP:NRC:1215B&D)
Brief description of amendments: The amendments modify the
technical specifications to replace the existing scheduling
requirements for overall integrated and local containment leakage rate
testing with a requirement to perform the testing in accordance with 10
CFR Part 50, Appendix J, Option B. Option B allows test scheduling to
be adjusted based on past performance.
Date of issuance: March 19, 1996
Effective date: March 19, 1996, with full implementation within 45
days
Amendment Nos.: Unit 1 - 209, Unit 2 -193
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 22, 1996 (61 FR
1632) The February 16, 1996 supplement made only a minor change to the
proposed technical specifications that provided consistency between the
wording for Units 1 and 2. The change did not affect the staff's
proposed finding that the amendments involve no significant hazards
consideration.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 19, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085.
The Cleveland Electric Illuminating Company, Centerior Service
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania
Power Company, Toledo Edison Company, Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: February 17, 1996
Brief description of amendment: The amendment allows one main steam
line's leakage rate to be as high as 35 standard cubic feet per hour
(scfh) as long as the total leakage through all four main steam lines
does not exceed 100 scfh until the end of Operating Cycle 6.
Date of issuance: March 18, 1996
Effective date: March 18, 1996
Amendment No.: 83
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications. The Commission's related evaluation of the
amendment and final no significant hazards consideration determination
is contained in a Safety Evaluation dated March 18, 1996. Public
comments requested as to proposed no significant hazards consideration:
Yes (61 FR 7823 dated February 29, 1996). That notice provided an
opportunity to submit comments on the Commission's
[[Page 16003]]
proposed no significant hazards consideration determination. No
comments have been received. The notice also provided for an
opportunity to request a hearing by April 1, 1996, but indicated that
if the Commission makes a final no significant hazards consideration
determination any such hearing would take place after issuance of the
amendment.
Local Public Document Room location: Perry Public Library, 3753
Main Street, Perry, Ohio 44081
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: December 19, 1994 (TXX-94274), as
supplemented by letter dated January 25, 1996 (TXX-96026)
Brief description of amendments: These changes allowed testing of
Reactor Protection System and Engineered Safety Features Actuation
System instrument channels with the channel under test in bypass in
order to reduce the vulnerability to spurious trips during surveillance
testing.
Date of issuance: March 14, 1996
Effective date: March 14, 1996
Amendment Nos.: Unit 1 - 47; Unit 2 - 33
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: February 1, 1995 (60 FR
6312) The additional information contained in the supplemental letter
dated January 25, 1996, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated March 14, 1996. No significant hazards consideration
comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, TX 76019.
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 requests: November 21, 1995 (TXX-95289), as
supplemented by letters dated February 22 (TXX-96061 and TXX-96062) and
28, (TXX-96068), and March 13, 1996 (TXX-96090).
Brief description of amendments: The amendments allowed both doors
of the containment personnel airlock to be open during fuel movement
and core alterations, providing one airlock door is capable of being
closed and the water level in the refueling pool is maintained.
Date of issuance: March 18, 1996
Effective date: March 18, 1996
Amendment Nos.: Unit 1 - 48; Unit 2 - 34
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 3, 1996 (61 FR
185) The additional information contained in the supplemental letters
dated February 22 (2 letters) and 28, and March 13, 1996, were
clarifying in nature and thus, within the scope of the initial notice
and did not affect the staff's proposed no significant hazards
consideration determination.The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated March 18, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government 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: May 26, 1994, as supplemented
January 5, April 25 and October 12, 1995, and February 2 and March 1,
1996.
Brief description of amendments: These amendments revise the
Technical Specifications by extending the operation of both units with
the current heatup and cooldown limit curves to 23.6 effective full
power years. The basis for TS Section 15.3.1.B, ``Pressure/Temperature
Limits,'' is also revised to reflect the methodology for the curve
compilation.
Date of issuance: March 20, 1996
Effective date: March 20, 1995
Amendment Nos.: 168 and 172
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
37093). The supplemental submittals provided additional information
that did not change the initial proposed no significant hazards
consideration determination.The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated March 20, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been
[[Page 16004]]
issued without opportunity for comment. If there has been some time for
public comment but less than 30 days, the Commission may provide an
opportunity for public comment. If comments have been requested, it is
so stated. In either event, the State has been consulted by telephone
whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By May 10, 1996, the licensee
may file a request for a hearing with respect to issuance of the
amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.714 which is available at the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC and at the local public document room for the
particular facility involved. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, 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
[[Page 16005]]
granted based upon a balancing of the factors specified in 10 CFR
2.714(a)(1)(i)-(v) and 2.714(d).
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of application for amendment: March 23, 1996
Brief description of amendment: The amendment modifies Technical
Specification (TS) 4.8.2.1.c, ``DC Sources - Operating,'' to specify
that the provisions of TS 4.0.1 and 4.0.4 are not applicable. This
provision expires upon entry into Mode 4 coming out of the sixth
refueling outage or upon any deep discharge cycle of the battery.
Date of issuance: March 23, 1996
Effective date: March 23, 1996
Amendment No.: Unit 2 - 94
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 23, 1996.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
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
NRC Project Director: William H. Bateman
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of application for amendment: March 26, 1996
Brief description of amendment: The amendment revises Technical
Specification (TS) 3/4.9.6 to allow the refueling machine overload
cutoff limit to be increased to as much as 2000 pounds, from the
current 1600 pound limit, in an effort to free the stuck fuel assembly
from core location A-06. The additional 400 pound increase will be
applied in 50 pound increments. This change will expire when the fuel
assembly located at core location A-06 is successfully withdrawn.
Date of issuance: March 26, 1996
Effective date: March 26, 1996, to be implemented prior to entry
into Mode 4 from the current refueling outage.
Amendment No.: Unit 2 - 95
Facility Operating License No. NPF-51: The amendment revised the
Technical Specifications.Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 26, 1996.
Local Public Document Room location: Phoenix Public Library, 1221
N. Central Avenue, Phoenix, Arizona 85004
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
NRC Project Director: William H. Bateman
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: March 29, 1996
Brief description of amendment: The amendment clarifies the testing
requirements and updates the regulatory and industry guidance
references for charcoal adsorber units addressed by TS 4.6.4.4,
Hydrogen Purge System; TS 4.6.5.1, Emergency Ventilation System; and TS
4.7.6.1, Control Room Emergency Ventilation System.
Date of issuance: March 29, 1996
Effective date: March 29, 1996
Amendment No.: 209
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No.The Commission's related
evaluation of the amendment, finding of emergency circumstances, and
final determination of no significant hazards consideration are
contained in a Safety Evaluation dated March 29, 1996.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Gail H. Marcus
Dated at Rockville, Maryland, this 3rd day of April 1996.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation.
[FR Doc. 96-8786 Filed 4-9-96; 8:45 am]
BILLING CODE 7590-01-F