[Federal Register Volume 61, Number 234 (Wednesday, December 4, 1996)]
[Notices]
[Pages 64381-64400]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-21204]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from November 8, 1996, through November 21, 1996.
The last biweekly notice was published on November 19, 1996.
NOTICE OF CONSIDERATION OF ISSUANCE OF AMENDMENTS TO FACILITY
OPERATING LICENSES, PROPOSED NO SIGNIFICANT HAZARDS CONSIDERATION
DETERMINATION, AND OPPORTUNITY FOR A HEARING
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules
Review and Directives Branch, Division of Freedom of Information and
Publications Services, Office of Administration, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, and should cite the
publication date and page number of this Federal Register notice.
Written comments may also be delivered to Room 6D22, Two White Flint
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15
p.m. Federal workdays. Copies of written comments received may be
examined at the NRC Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC. The filing of requests for a hearing and
petitions for leave to intervene is discussed below.
By January 3, 1997, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible
[[Page 64382]]
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 20555-0001, Attention: Docketing and
Services Branch, or may be delivered to the Commission's Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC,
by the above date. Where petitions are filed during the last 10 days of
the notice period, it is requested that the petitioner promptly so
inform the Commission by a toll-free telephone call to Western Union at
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union
operator should be given Datagram Identification Number N1023 and the
following message addressed to (Project Director): petitioner's name
and telephone number, date petition was mailed, plant name, and
publication date and page number of this Federal Register notice. A
copy of the petition should also be sent to the Office of the General
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001,
and to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: October 31, 1996
Description of amendment request: The proposed change would revise
the maximum allowable water temperature as measured at the respective
intake structures from 95 deg.F to 94 deg.F and will increase the
minimum main reservoir level from 205.7 feet mean sea level to 215 feet
mean sea level in Technical Specification (TS) 3/4.7.5, Ultimate Heat
Sink.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Since the proposed change does not affect the operation of any
accident initiating systems, the probability of occurrence of an
accident previously evaluated will not increase. Also, none of the
proposed changes will cause plant systems to operate outside their
design limits or create the likelihood of a radioactive release.
Therefore, there would be no increase in the consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No new component or system level interactions will be created by
the proposed change in ultimate heat sink level and temperature, and
no design limits will be exceeded. This change to [Technical]
Specification 3/4.7.5 is more conservative than the current
Specification limits and will serve only to restrict operation to a
higher reservoir level and lower temperature than was previously
allowed. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
The proposed amendment will establish a more conservative
minimum main reservoir level such that safety-related heat
exchangers served by Emergency Service Water will continue to remove
their design-basis accident heat loads. Establishing a higher
minimum reservoir level, combined with a more conservative reservoir
temperature assumption, will involve an increase in the margin of
safety. Also, the proposed change in maximum reservoir temperature
from 95 deg.F to 94 deg.F will not result in any reduction in the
margin of safety. A maximum pre-accident initial water temperature
of 94 deg.F is necessary to yield a post-accident (30-day)
calculated maximum inlet temperature less than or equal to the
design basis temperature of 95 deg.F. 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.
[[Page 64383]]
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
Attorney for licensee: William D. Johnson, Vice President and
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551,
Raleigh, North Carolina 27602
NRC Project Director: Mark Reinhart, Acting
Duke Power Company, Docket Nos. 50-413 and 50-414, Catawba Nuclear
Station, Units 1 and 2, York County, South Carolina
Date of amendment request: November 4, 1996
Description of amendment request: The proposed amendments would
eliminate from the Technical Specifications, Section 4.7.13.1, the
``during shutdown'' restriction pertaining to the 18-month Standby
Shutdown System (SSS) diesel generator inspection. Unlike Catawba
Nuclear Station, many nuclear plants do not have an SSS facility and
associated diesel generator. The requirements in the Technical
Specifications for the SSS diesel generator (shared between both units)
were patterned after similar requirements for the emergency diesel
generators. The current wording requires that both units be shut down
to perform the subject 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, which is presented below:
... The standard for determining that a Technical Specification
amendment request involves no significant hazards considerations
requires that operation of the facility in accordance with the
requested amendment will not:
1) Involve a significant increase in the probability or
consequences of an accident previously evaluated; or
2) Create the possibility of a new or different kind of accident
from any accident previously evaluated; or
3) Involve a significant reduction in the margin of safety.
Criterion 1
The proposed amendment seeks to change the surveillance
requirements to allow the SSS DG [diesel generator] periodic
inspection with one or both units on line. The surveillance can be
safely completed as proposed without affecting unit operation. The
equipment would not be removed from service for a time that would
exceed the current Limiting Condition For Operation or the
appropriate action statement would be entered. The probability or
consequences of any accident previously evaluated will not be
significantly increased because the removal of the SSS DG from
service can be safely performed while one or both units are
operating.
Criterion 2
The proposed amendment change does not change any actual
surveillance requirements. The change simply allows the 18 month SSS
DG inspection to be performed at different unit conditions. The
performance of the surveillance with the units operating do not
require any new component configurations that would reduce the
ability of any equipment to mitigate an accident. The station is not
degraded beyond that which has been previously evaluated. Therefore
the proposed change does not create the possibility of a new or
different kind of accident.
Criterion 3
The allowed outage time for the SSS DG, as specified by the
Limiting Condition For Operation, defines the required margin of
safety for equipment operability. Removing the SSS DG from service
for periodic inspection and returning it to service within the
allowed outage time 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
proposed amendments involve 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
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: October 30, 1996
Description of amendment request: The proposed changes would (1)
completely rewrite Technical Specification (TS) 4.4.2 to incorporate a
prestressed concrete containment surveillance program that is
consistent with Regulatory Guide 1.35, (2) modify TS 3.6.7 by
establishing new Limiting Conditions for Operation and required actions
related to the structural integrity of the reactor buildings, (3)
incorporate an editorial change to TS 6.6.3 to reference the relocated
tendon surveillance reporting requirements, and (4) modify TS 3.6.7
Bases to describe the Reactor Building post-tensioning 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) Will the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed amendment to Oconee Technical Specifications
involves the implementation of an enhanced surveillance program for
the reactor building prestressed concrete containment and the
assurance of appropriate station response to abnormal degradation of
the containment structure. The proposed change will move Oconee into
a surveillance program which is consistent with accepted industry
practice and a published NRC regulatory position. The adoption of
Regulatory Guide 1.35 as a basis for the periodic inspection of the
reactor building prestressed concrete containment and clearly
defined station response to any indication of structural
deterioration will assure acquisition of sufficient data to
demonstrate that structural integrity is maintained and, if
necessary, appropriate compensatory action(s) are taken. By assuring
that any adverse trends in the behavior of the prestressed concrete
containment are identified and acted upon in a timely manner, this
change does not increase the probability or consequences of an
accident previously evaluated.
2) Will the change create the possibility of a new or different
kind of accident from any previously evaluated?
No. The proposed amendment to Oconee Technical Specifications
involves the implementation of an enhanced surveillance program for
the reactor building prestressed concrete containment and the
assurance of appropriate station response to abnormal degradation of
the containment structure. By adopting Regulatory Guide 1.35 as a
basis for the surveillance inspection program for the reactor
building prestressed concrete containment and clearly defining
required station response to any indication of structural
deterioration, sufficient data will be obtained to demonstrate that
structural integrity is being maintained and that any adverse
behavioral trends are identified and acted upon in a timely manner.
Therefore, the proposed amendment does not create the possibility of
any type of accident: new, different or previously evaluated.
3) Will the change involve a significant reduction in a margin
of safety?
No. Margin of safety is associated with confidence in the
ability of the fission product barriers (i.e., fuel and fuel
cladding, Reactor Coolant System pressure boundary, and containment
structure) to limit the level of radiation dose to the public. The
proposed Technical Specifications amendment will move Oconee into a
surveillance program which is consistent with accepted industry
practice and a published regulatory position. By ensuring more
timely identification of, and response to, any adverse trend in the
behavior of the reactor building prestressed concrete containment,
continued maintenance of the structural integrity is enhanced.
Therefore, the ability of the containment structure to perform the
intended function of protecting the public
[[Page 64384]]
from radiation dose is further assured, and no reduction in any
existing margin of safety will occur.
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: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412,
Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: September 9, 1996
Description of amendment request: The proposed amendment would
modify the design features section (Section 5.0) of the Technical
Specifications (TSs) to make the design features section consistent
with the four criteria specified in the Commission's Policy Statement
on TSs (58 FR 39132) and with the guidance provided in the NRC's
Standard Technical Specifications, Westinghouse Plants (NUREG-1431,
Revision 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. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change reduces the content of the technical
specification (TS) design feature section consistent with the
Improved Standard Technical Specifications (ISTS) of NUREG-1431. The
information that has been removed is also contained in the UFSAR
[Updated Final Safety Analysis Report] or offsite dose calculation
manual (ODCM); therefore, duplication of the information is
eliminated to improve the use of the TS. Because the information
removed from the TS is maintained in the UFSAR or ODCM where changes
are controlled in accordance with regulatory requirements, there is
no reduction in commitment and adequate control is provided.
Elimination of information from the design feature section of the TS
which duplicates information in the UFSAR enhances the usability of
the TS without reducing commitments. These changes clarify and
improve the understanding and readability of the TS. Since the
requirements remain the same, these changes only affect the method
of presentation and would not affect possible initiating events for
accidents previously evaluated or any system functional requirement.
Therefore, the proposed changes would 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 relocation of existing requirements, the elimination of
requirements which duplicate existing information, and making
administrative improvements are all changes that are administrative
in nature. The proposed changes will not affect any plant system or
structure, not [nor] will they affect any system functional or
operability requirements. Consequently, no new failure modes are
introduced as a result of the proposed changes. The proposed changes
are consistent with the ISTS, for the most part, as plant-specific
information is included in this section. Therefore, the proposed
change will not create the possibility of a new or different type of
accident from any accident previously evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed changes are administrative in nature in that no
change to the design features of the facility are being made. The
design features section is being reformatted to be consistent, for
the most part, with the ISTS. The proposed changes do not affect the
UFSAR design bases, accident assumptions, or technical specification
bases. In addition, the proposed changes do not affect release
limits, monitoring equipment or practices. Therefore, the proposed
change will not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Aliquippa, PA 15001
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 24, 1996
Description of amendment request: The proposed amendment would
revise the technical specifications to remove accelerated testing
requirements for the standby diesel generators. The changes implement
the provisions of Generic Letter (GL) 94-01, ``Removal of Accelerated
Testing and Special Reporting Requirements For Diesel Generators'',
dated May 31, 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:
1. This request does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
This change will provide flexibility to structure the standby
diesel generator maintenance program based on the risk significance
of the structures, systems, and components that are within the scope
of the Maintenance Rule. The removal of the diesel generator
accelerated testing is acceptable as the maintenance rule applies
site and system specific performance criteria to monitor diesel
generator performance. This criteria includes a running availability
and reliability goal as well as specific goals to monitor
maintenance preventable functional failures. The performance
criteria for the diesel generator reliability and unavailability
established by the maintenance rule and the causal determinations
and corrective actions required for maintenance preventable
functional failures are considered to be an acceptable method for
monitoring diesel generator performance.
The proposed change has no effect on the probability of the
initiation of an accident, because the emergency diesel generators
do not serve as the initiator of any event. Additionally, as diesel
generator performance will continue to be assured by the maintenance
rule, the proposed changes do not affect the ability to mitigate the
consequences of an accident previously evaluated. The changes do not
impact the diesel's design sources, operating characteristics,
system functions, or system interrelationships. The failure
mechanisms for the accidents previously analyzed are not affected
and no additional failure modes are created that could cause an
accident that has been previously evaluated. Since the diesel
generator performance and reliability will continue to be assured by
the maintenance rule, the proposed changes cannot involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. This request does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
This proposed change does not involve a change to the plant
design or operation. As a result, the proposed changes does not
affect any of the parameters or conditions that could contribute to
the initiation of any accidents. The proposed changes only affect
the methods used to monitor and assure diesel generator performance.
The performance criteria for both the diesel generator reliability
and unavailability established by the maintenance rule, and the
[[Page 64385]]
casual determinations and corrective actions required for
maintenance preventable functional failures, is considered by GL 94-
01 to be an acceptable method for monitoring diesel generator
performance.
No [system, structure, or component] SSC, method of operating,
or system interface is altered by this change. The changes do not
impact the diesel's design sources, operating characteristics,
system functions, or system interrelationships. The failure
mechanisms for the accidents are not affected, and no additional
failure modes are created. Because the diesel generator performance
and reliability will continue to be assured by the maintenance rule,
the proposed changes cannot create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The request does not involve a significant reduction in a
margin to safety.
The proposed changes only affect the methods used to monitor and
assure diesel generator performance and reliability. The performance
criteria for both the diesel generator reliability and
unavailability established by the maintenance rule, and the casual
determinations and corrective actions required for maintenance
preventable functional failures, is considered by GL 94-01 to be an
acceptable method for monitoring diesel generator performance. No
margin to safety as defined in the basis for any technical
specification is impacted by these changes. This change does not
impact any uncertainty in the design, construction, or operation of
any SSC. Diesel generator response to accident initiators is
unchanged. No SSC, method of operating, or system interface is
altered by this change. The changes do not impact the diesel's
design sources, operating characteristics, system functions, or
system interrelationships. Because the diesel generator performance
and reliability will continue to be assured by the maintenance rule,
the proposed changes cannot involve a significant reduction in the
margin to 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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: November 6, 1996
Description of amendment request: The proposed amendment would
revise the River Bend Station (RBS) Fire Hazards Analysis Report and
Safety Analysis Report to allow a deviation from 10 CFR Part 50,
Appendix R, Section III.G.2.c with respect to the requirement for an
area wide automatic fire suppression system in Fire Area C-16. The
deviation would allow a 1-hour barrier to separate redundant trains of
post fire safe shutdown equipment within the fire area and partial
sprays on the protected 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 request does not involve an increase in the probability
or consequences of an accident previously evaluated.
The event of concern is a fire in Fire Area C-16. The low fire
loading and minimal concentration of exposed combustible material in
Fire Area C-16 would limit fire spread. However, for this scenario,
all unprotected equipment in Fire Area C-16 will be assumed lost.
Fire Area C-16 contains cables for both Division I and Division II
components required for post fire safe shutdown. The loss of both
divisions of cables could preclude the ability of the plant to
achieve post fire safe shutdown. Protection of the required Division
II cables in a 1-hour fire barrier in conjunction with a partial
area, automatic suppression system installed above and below the
protected trays will ensure that post fire safe shutdown can be
achieved.
In summary, the probability of a fire occurring in Fire Area C-
16 is not affected. However, if a fire were to occur in Fire Area C-
16 which caused the loss of Division I powered components, Division
II powered components, by virtue of the 1-hour fire barrier and
partial area, automatic suppression system, would remain available.
The low fire loading and minimal concentration of exposed
combustible material in Fire Area C-16 would limit fire spread. The
proposed fire protection scheme provides a level of protection
commensurate with the original design. Therefore, this request does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
Fire Area C-16 will be protected by a partial area, automatic
suppression system installed above and below the protected cable
trays. Fire suppression systems are generally used to limit fire
spread, once the heat of the fire opens thermally sensitive
sprinklers. The low fire loading and minimal concentration of
exposed combustible material in Fire Area C-16 would aid in limiting
fire spread, and would also limit the severity of any plausible
fire. The previous analysis assumed all Division I components and
cables in the area would be lost, and that the installed fire
barrier would adequately protect the Division II cables routed
through C-16. The required Division II cables will be enclosed in a
1-hour fire barrier with a partial area, automatic suppression
system. These features provide a level of protection commensurate
with that of the previous design. In addition, the total combustible
loading in the area results in a maximum theoretical worst case fire
duration of less than 1-hour.
In summary, if a fire were to occur in Fire Area C-16 which
caused the loss of Division I powered components, post fire safe
shutdown could still be achieved using Division II. Therefore, this
request does not create the possibility of occurrence of a new or
different kind of accident from any accident previously evaluated.
3. The request does not involve a significant reduction in a
margin of safety.
In this case, the margin of safety is implicit rather than being
explicitly expressed as a numerical value. An implicit margin of
safety involves conditions for NRC acceptance. Since the RBS
Technical Specification Bases do not specifically address a margin
of safety for fire protection, the SAR [Safety Analysis Report], the
NRC's Safety Evaluation Report (SER), and appropriate other
licensing basis documents were reviewed to determine if the proposed
change would result in a reduction in a margin of safety. As stated,
in part, in Attachment 4 to NPF-47 [Facility Operating License; NPF-
47]:
EOI [Entergy Operations, Inc.] shall implement and maintain in
effect all provisions of the approved fire protection program as
described in the Final Safety Analysis Report for the facility
through Amendment 22 and as approved in the SER dated May 1984 and
Supplement 3 dated August 1985 subject to provisions 2 and 3 ....
As discussed in the Reason for Request, SSER [Supplemental
Safety Evaluation Report] 3 dated August 1985 states, in part:
On the basis of its evaluation the staff finds that the
applicant's fire protection program with approved deviations is in
conformance with the guidelines of BTP CMEB [branch technical
position, Chemical Materials and Engineering Branch] 9.5-1,
[S]sections III.G, III.J, and III.O of Appendix R to 10 CFR [Part]
50, and GDC [General Design Criteria] 3, and is, therefore,
acceptable.
Thus, the margin of safety in this case can be defined as
conformance with the specified fire protection guidelines.
10 CFR [Part] 50, Appendix R, Section III.G.2, requires, in
part, that redundant trains of post fire safe shutdown equipment
located in the same fire area be separated by a 1-hour fire barrier
and, in addition, that fire detection and an automatic fire
suppression system be installed in the are under consideration.
Since Fire Area C-16 will have a partial area, automatic suppression
system, this fire area would deviate from the requirements of 10 CFR
[Part] 50, Appendix R, Section III.G.2.c. However, as discussed
previously, the installed partial area, automatic suppression
system, the low fire loading and minimal amount of exposed
combustibles compensate for the lack of a total, area wide,
automatic fire suppression
[[Page 64386]]
system. There is no adverse impact on the ability to achieve and
maintain post fire safe shutdown. Therefore, this request 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: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant Units 1 and 2, St. Lucie County, Florida
Dates of amendment requests: October 28, 1996 (Two letters)
Description of amendment request: The licensee proposed to change
the St. Lucie Units 1 and 2 Technical Specifications (TS) to implement
10 CFR 50, Appendix J, Option B, for containment leakage testing by
referring to Regulatory Guide 1.163, ``Performance-Based Containment
Leakage-Test Program.'' Changes include relocating the details for
containment testing to the ``containment leakage rate testing program''
and adding the requirements of the containment leakage rate testing
program to TS 6.8.4, which describes facility programs. Changes are
also proposed to remove Tables 3.6-1, ``Containment Leakage Paths,''
and 3.6-2, ``Containment Isolation Valves'' from TS and relocate the
information to plant procedures.
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
due to the following reasons:
a)These proposed changes are all consistent with NRC
requirements and guidance for implementation of 10 CFR 50, Appendix
J, Option B, except for the use of Bechtel Topical Report BN-TOP-1
for type A testing. BN-TOP-1 has been previously approved for use in
accordance with 10 CFR 50 appendix J.
b) Based on industry and NRC evaluations performed in support of
developing Option B, these changes potentially result in a minor
increase in the consequences of an accident previously evaluated due
to the increased testing intervals. However, the proposed changes do
not result in an increase in the core damage frequency since the
containment system is used for mitigation purposes only.
c) These changes are expected to result in increased attention
to components with poor leakage test history as part of the
performance-based nature of Option B, such that the marginally
increased consequences from the expanded testing intervals may be
further reduced or negated.
Therefore, these changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.(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 modified specifications can not create the
possibility of a new or different kind of accident from any
previously evaluated since the proposed amendments will not change
the physical plant or the modes of plant operation defined in the
facility operating license. No new failure mode is introduced due to
the implementation of a performance-based program for containment
leakage rate testing, 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, structures, or
components.(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, structures, and components are basically unchanged by the
proposed amendments. The increase in intervals between leak-test
surveillances will not significantly reduce the margin of safety as
shown by findings in NUREG 1493, ``Performance-Based Containment
Leak-Test Program'', which was based on implementation of the
performance-based testing of Option B.
Therefore these changes do not involve a significant reduction
in the margin of safety.The NRC staff has reviewed the licensee's
analysis and, based on thisreview, 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: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light,
11770 US Highway 1, North Palm Beach, Fl 33408
NRC Project Director: Frederick J. Hebdon
Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389,St. Lucie Plant Units 1 and 2, St. Lucie County, Florida
Date of amendment request: October 30, 1996
Description of amendment request: The proposed amendments will
revise Technical Specification (TS) 3/4.9.10, ``Refueling Operations,
Water Level-Reactor Vessel.'' The Limiting Condition for Operation
(LCO) specified for the minimum allowed refueling water level is not
altered, but the Applicability, Action, and Surveillance Requirements
are changed to remove inconsistencies with the definition of Core
Alterations, and to achieve consistency with the generic Standard
Technical Specifications for Combustion Engineering Plants (NUREG-
1432). An editorial change is proposed for TS 3/4.9.9, ``Refueling
Operations, Containment Isolation System,'' and, for St. Lucie Unit 1,
the LCO is modified to conform with other related refueling
specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below.
(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.
Certain evolutions performed with the UGS [upper guide
structure] in place are not Core Alterations, and the revised LCO 3/
4.9.10 will allow these activities to be performed at water levels
other than prescribed by the existing LCO. Since these activities
are performed with the UGS in place, the probability that a fuel
handling accident would occur is not impacted by the proposed
changes. The minimum water level required for Core Alterations and
movement of irradiated fuel in containment is not altered by the
proposed changes, nor are any assumptions or conditions changed that
were used as inputs to the evaluation of fuel handling accident
consequences. The changes to Specification 3/4.9.9 are
administrative in nature and resolve an inconsistency between the
operability requirements for the containment isolation system and
the containment penetrations that the system would isolate at PSL1
[Plant St. Lucie 1]. Therefore, operation of either facility in
accordance with its proposed amendment would not involve a
significant 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
[[Page 64387]]
kind of accident from any accident previously evaluated.
The proposed changes are administrative in nature, in that the
changes do not involve the addition or modification of equipment nor
do they alter the design of plant systems. New failure modes are not
introduced, and the physical plant or the modes of plant operation
defined in the Facility License are not altered. Therefore,
operation of either facility in accordance with its proposed
amendment would 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.
The safety margin associated with a fuel handling accident is
determined, in part, by the minimum refueling water level allowed
for conducting Core Alterations and movement of irradiated fuel in
containment. The minimum water level required by LCO 3/4.9.10, or
other factors considered as inputs to the safety analysis, is not
changed by the proposed amendments. The revised applicability
requirements for LCO 3/4.9.9 at PSL1 will allow the containment
isolation system to be inoperable only during those Mode 6
conditions where Core Alterations or irradiated fuel movements
within containment are not in progress, or each required containment
penetration is otherwise closed. Therefore, operation of either
facility in accordance with its proposed amendment would not involve
a significant 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 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendmentrequest involves no significant hazards consideration.
Local Public Document Room location: Indian River Junior College
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light,
11770 US Highway 1, North Palm Beach, Fl 33408
NRC Project Director: Frederick J. Hebdon
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: October 23, 1996, as supplemented by
letter dated November 6, 1996.
Description of amendment request: The proposed amendment would
revise Technical Specification 3.4.6.1, regarding reactor coolant
system leakage detection instrumentation, to adopt the requirements
found in NUREG-1431, ``Standard Technical Specifications Westinghouse
Plants,'' for the reactor coolant system leakage detection
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 involved a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change reduces the number of containment
atmospheric radioactivity channels which must be OPERABLE when
operating in MODES 1, 2, 3, and 4 from two to one. This change does
not significantly increase the probability or consequences of a
previously evaluated accident since the plant will continue to have
diverse and independent means of detecting significant changes in
the amount of leakage from the RCS [reactor coolant system]; the
normal sump level and flow monitoring system, at least one of the
two containment atmospheric radiation monitors, and the periodic
precision RCS water inventory balance required by Technical
Specification surveillance requirement 4.4.6.2.1.c. In addition, STP
[South Texas Project] design includes advanced trending displays
which can assist in detecting leakage based on changes in the volume
control tank or pressurizer level. Other instruments, which are not
listed in the Technical Specification related to leakage, but which
can provide indication of leakage, are the containment pressure,
temperature and humidity indicators. Good operating practice and
commercial risk associated with long term inoperability of both
monitors assures that an inoperable containment atmospheric
radiation monitor will be promptly returned to service.
The proposed change also revises the limitation on continued
operation with both containment atmospheric radiation monitors
inoperable from 72 hours to 30 days. This change is based on the
continued availability of diverse and redundant instrumentation
discussed above to detect and indicate RCS leakage.
The Actions required as a result of this change include analysis
of a containment atmospheric grab sample or performance of a
precision RCS water inventory balance in accordance with
surveillance requirement 4.4.6.2.1.c. The containment normal sump
level flow monitoring system will also promptly identify changes in
RCS leakage. Other installed instrumentation, such as containment
pressure, temperature, and humidity, will provide indications of
significant increases in leakage. Slower increases will be detected
by the daily inventory balance or the daily grab samples analysis,
and the three day inventory balance.
Inoperability of the on-line automatic containment normal sump
level and flow monitoring system can be compensated for by the
performance of a daily manual calculation, a precision RCS inventory
balance as described in surveillance requirement 4.4.6.2.1.c, or the
other available indications of increases in leakage such as the
containment atmospheric radiation monitoring instruments and
installed containment temperature, pressure and humidity
instrumentation. The STP control room design also incorporates
features which allow rapid detection of unexpected changes in the
volume control tank and pressurizer level through available
instrument trend displays. The combination of the compensatory
measures, diverse and separate channels, and non-TS [non-technical
specification] required instrumentation provides a sufficient level
of detection to assure prompt identification and quantification of
leakage with an inoperable containment normal sump level and flow
monitoring system. The allowable outage time of 30 days provides
assurance the normal containment sump level and flow monitoring
system will be returned to service in a reasonable amount of time.
Based on the continued availability of adequate and redundant
instrumentation to detect changes in RCS leakage rate, this change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed change does not require the installation of any new
or different kind of equipment. Nor does the change involve any
significant new or different MODE of operation of the plant. The
proposed change reduces the number of required containment
atmospheric radiation monitors, and provides a 30 day allowed outage
time for either the containment atmosphere radioactivity monitor or
the containment normal sump level and flow monitoring system.
Therefore, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
In addition, as described above, the proposed change does not
significantly reduce a margin of safety. Small changes in RCS leak
rates are typically detected over a relatively long period of time.
Diverse instrumentation continues to be available to plant operators
which will assist in early detection of any change. The STP design
provides additional non-Technical Specification human factors which
assist in assuring any changes in leakage will be quickly detected.
The proposed change extends the amount of time that the
containment atmospheric radiation monitors may be inoperable. The
extension is based on the continued availability of equipment which
provides a level of detection capability adequate to detect
increases in RCS leakage and which continues to be diverse and
independent. This protection is afforded by the continued
OPERABILITY of the containment normal sump level and flow monitoring
system, the daily performance of a precision RCS
[[Page 64388]]
inventory balance as described by surveillance requirement
4.4.6.2.1.c or the daily analysis of containment atmospheric grab
samples, and other instrumentation such as pressure, temperature and
humidity indicators.
The combination of the compensatory measures, diverse and
separate channels, and non-TS required instrumentation provides a
sufficient level of detection to assure prompt identification and
quantification of leakage with an inoperable containment normal sump
level and flow monitoring system. Additionally, the compensatory
measure of performing either a daily manual calculation or precision
RCS inventory balance, provides assurance that the level of safety
is maintained when the containment normal sump level and flow
monitoring system is inoperable. The allowable outage time of 30
days provides assurance the normal containment sump level and flow
monitoring system will be returned to service in a reasonable amount
of time.
Based on the compensatory actions and available installed
equipment, the proposed changes do not represent a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis &
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
NRC Project Director: William D. Beckner
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: August 15, 1996
Description of amendment requests: The proposed amendments would
revise the Containment Cooling Systems Limiting Conditions for
Operation Technical Specifications to bring them into conformance with
recently completed system analyses by no longer permitting both
containment spray pumps to be inoperable at the same time.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Operation of the Prairie Island plant in accordance with the
proposed changes does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
None of the proposed changes involve a physical modification to the
plant.
These changes will require operability of at least one
containment spray pump at all times and reduces the spray additive
tank allowable outage time from 72 hours to 24 hours. Both of these
changes are more conservative and safer than currently required in
the Prairie Island Technical Specifications. These proposed changes
do allow one containment fan cooler train out of service for 7 days
instead of 72 hours as allowed by current Technical Specifications.
Recent plant analyses confirm that one containment fan cooler train
with one containment spray train is sufficient to meet the system
design bases. Since the probability of an accident occurring is low
while one containment fan cooler train is out of service, the
probability and consequences of an accident are not significantly
increased.
In total these changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated
because the proposed changes, in themselves, do not introduce a new
mode of plant operation, surveillance requirement or involve a
physical modification to the plant.
The proposed changes do require more restrictive, safer
containment spray train operability. The proposed changes also allow
one containment fan cooler train to be out of service for 7 days
instead of 72 hours as allowed by the current Technical
Specifications. However, this change does not create the possibility
of a new kind of accident.
The proposed changes do no alter the design, function, or
operation of any plant components and therefore, no new accident
scenarios are created.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be created
by these amendments.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety. This License Amendment Request
require[s] one containment spray train to be operable at all times
which is more restrictive than current Technical Specifications and
thus the margin of safety is not reduced.
This License Amendment Request will also allow one containment
fan cooler train to be out of service for 7 days instead of 72 hours
as allowed by the current Technical Specifications. Since the
remaining containment cooling components can mitigate an accident
and the probability of a design basis accident are low during this
time, this change does not significantly reduce the plant margin of
safety.
Therefore, a significant reduction in the margin of safety would
not be involved with these amendments.
Based on the evaluation described above, and pursuant to 10 CFR
Part 50, Section 50.91, Northern States Power Company has determined
that operation [of] the Prairie Island Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by Nuclear
Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.
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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Northern States Power Company, Docket Nos. 50-282 and 50-306,
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue
County, Minnesota
Date of amendment requests: September 24, 1996, as supplemented
October 17, 1996.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TS) for the Prairie Island Nuclear
Generating Plant to allow use of an alternate steam generator tube
repair criteria (elevated F-star or EF*) in the tubesheet region when
used with the repair method of additional roll expansion. The
amendments incorporate revised acceptance criteria for tubes with
degradation in the tubesheet region and enable the licensee to avoid
unnecessary plugging and sleeving of steam generator tubes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The supporting technical and safety evaluations of the subject
criterion
[[Page 64389]]
demonstrate that the presence of the tubesheet will enhance the tube
integrity in the region of the hardroll by precluding tube
deformation beyond its initial expanded outside diameter. The
resistance to both tube rupture and tube collapse is strengthened by
the presence of the tubesheet in that region. The results of
hardrolling of the tube into the tubesheet is an interference fit
between the tube and the tubesheet. Tube rupture cannot occur
because the contact between the tube and tubesheet does not permit
sufficient movement of tube material. The radial preload developed
by the rolling process will secure a postulated separated tube end
within the tubesheet during all plant conditions. In a similar
manner, the tubesheet does not permit sufficient movement of tube
material to permit buckling collapse of the tube during postulated
LOCA [loss-of-coolant accident] loadings.
The EF* length of roll expansion is sufficient to preclude tube
pullout from tube degradation located below the EF* distance,
regardless of the extent of the tube degradation. The existing
Technical Specification leakage rate requirements and accident
analysis assumptions remain unchanged in the unlikely event that
significant leakage from this region does occur. As noted above,
tube rupture and pullout is not expected for tubes using the EF*
criterion. Any leakage out of the tube from within the tubesheet at
any elevation in the tubesheet is fully bounded by the existing
steam generator tube rupture analysis included in the Prairie Island
Plant USAR [updated safety analysis report]. For plants with partial
depth roll expansion like Prairie Island, a postulated tube
separation within the tube near the top of the roll expansion (with
subsequent limited tube axial displacement) would not be expected to
result in coolant release rates equal to those assumed in the USAR
for a steam generator tube rupture event due to the limited gap
between the tube and tubesheet. The proposed plugging criterion does
not adversely impact any other previously evaluated design basis
accident.
Leakage testing of roll expanded tubes indicates that for roll
lengths approximately equal to the EF* distance, any postulated
faulted condition primary to secondary leakage from EF* tubes would
be insignificant.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
Implementation of the proposed EF* criterion does not introduce
any significant changes to the plant design basis. Use of the
criterion does not provide a mechanism to initiate an accident
outside of the region of the expanded portion of the tube. Any
hypothetical accident as a result of any tube degradation in the
expanded portion of the tube would be bounded by the existing tube
rupture accident analysis. Tube bundle structural integrity will be
maintained. Tube bundle leaktightness will be maintained such that
any postulated accident leakage from EF* tubes will be negligible
with regard to offsite doses.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The use of the EF* criterion has been demonstrated to maintain
the integrity of the tube bundle commensurate with the requirements
of Reg Guide 1.121 [Bases for Plugging Degraded PWR Steam
Generator Tubes] (intended for indications in the free
span of tubes) and the primary to secondary pressure boundary under
normal and postulated accident conditions. Acceptable tube
degradation for the EF* criterion is any degradation indication in
the tubesheet region, more than the EF* distance below the bottom of
the transition between the roll expansion and the unexpanded tube.
The safety factors used in the verification of the strength of the
degraded tube are consistent with the safety factors in the ASME
[American Society of Mechanical Engineers] Boiler and Pressure
Vessel Code used in steam generator design. The EF* distance has
been verified by testing to be greater than the length of roll
expansion required to preclude both tube pullout and significant
leakage during normal and postulated accident conditions. Resistance
to tube pullout is based upon the primary to secondary pressure
differential as it acts on the surface area of the tube, which
includes the tube wall cross-section, in addition to the inner
diameter based area of the tube. The leak testing acceptance
criteria are based on the primary to secondary leakage limit in the
Technical Specifications and the leakage assumptions used in the
USAR accident analyses.
Implementation of the tubesheet plugging criterion will decrease
the number of tubes which must be taken out of service with tube
plugs or repaired with sleeves. Both plugs and sleeves reduce the
RCS (reactor coolant system) flow margin; thus, implementation of
the EF* criterion will maintain the margin of flow that would
otherwise be reduced in the event of increased plugging or sleeving.
Based on the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the USAR or the Technical Specification
Bases.
Based on the evaluation described above, and pursuant to 10 CFR
Part 50, Section 50.91, Northern States Power Company has determined
that operation of the Prairie Island Nuclear Generating Plant in
accordance with the proposed license amendment request does not
involve any significant hazards considerations as defined by NRC
regulations in 10 CFR Part 50, Section 50.92.
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: Minneapolis Public Library,
Technology and Science Department, 300 Nicollet Mall, Minneapolis,
Minnesota 55401
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and
Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
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: June 10, 1996, as supplemented July 25,
1996
Description of amendment request: The proposed amendment would
change the differential temperature Technical Specification Allowable
Values and Trip Setpoints for the Reactor Water Cleanup penetration
room steam leak detection function.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability [of occurrence] [sic] or consequences of an
accident evaluated.
FSAR section 5.2.5.1.3 addresses the ambient and differential
room ventilation temperature leakage detection. This section states:
``...switch setpoints are based on the temperature rise
resulting from a leak at system conditions corresponding to full
reactor power.''
NRC Safety Evaluation on the RWCU system steam leak detection
system (related to Amendment Number 123 to License NPF-14 and
Amendment Number 90 to License NPF-22) reviewed and found acceptable
the PP&L criteria for calculating the leak detection setpoints for
the RWCU system, which include:
1. Setpoints are selected to detect and isolate a leak that is
normally less than 25 gpm and below the flow rate corresponding for
the critical crack size for the system piping.
2. Setpoints are set high enough to avoid inadvertent isolation
caused by normal temperature transients or abnormal transients
caused by non-leak conditions (such as loss of ventilation).
This NRC SER also stated that a leak rate of 25 gpm is less than
those leak rates associated with the onset of unstable pipe
ruptures. This fact is also shown in FSAR figure 5.2-10. This value
of 25 gpm constitutes the design basis for the steam leak detection
system.
The mixing and liquid energy addition assumption changes in the
analysis do not affect this design basis. The analysis calculates
the resulting room temperature increase from a 25 gpm leak. In fact,
the new assumptions provide a more accurate yet conservative
prediction of room temperature increases. Therefore, operation of
the system is improved.
[[Page 64390]]
The proposed change leads to higher calculated room temperatures
to be used in the differential temperature setpoint calculations.
The engineering study was reviewed to determine if the higher
calculated temperatures would have a negative impact on the High
Energy Line Break and Leak Analysis environmental study which
provides the basis for equipment qualification.
In determining the room temperatures, the engineering study
considers ambient temperature setpoints at which the leaks will be
isolated. The proposed action will not change the ambient
temperature setpoints, and actuation of these instruments will
ensure that the results of the engineering study will not be
adversely affected. Therefore, no impact on equipment qualification
is being introduced by this change.
FSAR chapter 15 was reviewed for potential impacts on the
accident analyses. The 25 gpm leak outside containment is not
specifically analyzed in FSAR chapter 15. However, other conditions
which result in coolant leakage outside containment are analyzed in
section 15.6.2 (Instrument Line Break) and 15.6.4 (Steam System
Piping Break Outside Containment). As stated in the NRC SER, the 25
gpm RWCU leak rate is bounded by the analysis in FSAR section
15.6.4. FSAR section 15.6.2 also states that leak detection
actuations will initiate operator actions, a fact that is not
affected by the proposed change. Therefore, based on a review of
FSAR chapter 15 it was concluded that no impact on the analyzed
accident scenarios is created by the proposed change.
Based on the above discussions, it is demonstrated that the
proposed change will not adversely impact system function or
equipment. System performance will actually be improved since the
new setpoints eliminate spurious isolations resulting from a less
accurate model. The setpoint change has no impact on any equipment
important to safety or any accidents previously analyzed in the
FSAR. Therefore, the proposed change does not involve a significant
increase in the probability of occurrence 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 action does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
Neither the system design basis nor the system function will be
adversely affected. System performance will be enhanced since
spurious differential temperature actuations will be reduced as a
result of using the more accurate, yet conservative, COTTAP model.
In addition to this, redundant temperature isolation function will
continue to be provided by the existing high ambient temperature
detectors.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed action does not involve a significant reduction in
a margin of safety. The Technical Specification basis for the
setpoints is to detect a leak below the flow rate corresponding to
critical crack size for the system piping. As stated previously, the
25 gpm flow rate is an acceptable flow rate and is used to calculate
the new temperatures.
Although the newly calculated RWCU penetration room temperatures
will be higher (due to the improved model), the isolation actuation
will be initiated by the high ambient temperature detectors before
the penetration room temperatures reach the newly calculated values,
as would happen under the old model. Therefore, system response is
not adversely affected.
The current temperature values lead to differential temperature
setpoints which are too low, causing spurious isolations. The use of
the new temperature values will reduce the number of spurious
isolations, reducing unnecessary challenges to safety systems during
normal plant operations.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: September 18, 1995
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Table 4.3.1.1-1, ``Reactor
Protection System Instrumentation Surveillance Requirements'' to
reflect the change in the calibration frequency for the Local Power
Range Monitor (LPRM) signal from every 1000 Effective Full Power Hours
(EFPH) to every 2000 Megawatt Days per Standard Ton (MWD/ST).
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 Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The change in the calibration frequency of the Local Power Range
Monitor (LPRM) signal does not make any physical change to the fuel
or the manner in which the fuel responds to a transient or accident.
The proposed TS change does not affect the fundamental method by
which the LPRMs are calibrated. Also, the LPRM calibration frequency
is not considered an initiator of any events analyzed in the SAR.
Therefore, calibrating the LPRMs on a different frequency will not
increase the probability of occurrence of an accident previously
evaluated in the SAR.
The resulting nodal power uncertainty does not exceed the nodal
power uncertainty accounted for in the existing Minimum Critical
Power Ratio (MCPR) Safety Limit; thus, the MCPR Safety Limit is not
affected by this TS Change, and, therefore, the initial conditions
of any accident are unchanged. Since the calibration frequency
change will not affect the course of any evaluated accident, the
consequences of an accident previously evaluated in the SAR will not
be increased.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The change in the calibration frequency of the Local Power Range
Monitor (LPRM) signal does not make any physical change to the plant
or the manner in which the equipment responds to a transient or
accident. The proposed TS change does not introduce a new mode of
plant operation and does not involve the installation of any new
equipment or instrumentation. The fuel will continue to be operated
to the same safety limits since the Minimum Critical Power Ratio
(MCPR) Safety Limit remains unchanged due to this TS change.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident, from any
accident previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
The following TS Bases were reviewed for potential reduction in
the margin of safety:
2.0 Safety Limits and Limiting Safety System Settings;
3/4.1 Reactivity Control Systems;
3/4.2.1 Average Planar Linear Heat Generation Rate;
3/4.2.3 Minimum Critical Power Ratio:
3/4.2.4 Linear Heat Generation Rate;
3/4.3.1 Reactor Protection System Instrumentation;
3/4.3.6 Control Rod Block Instrumentation;
3/4.3.7.7 Traversing In-Core Probe System;
The GE Thermal Analysis Basis (GETAB) determination of the
Minimum Critical Power Ratio (MCPR) Safety Limit allows a maximum
total nodal uncertainty of the Traversing In-Core Probe (TIP)
readings of which the Local Power Range Monitor (LPRM).
Update uncertainty is a part. The change in LPRM calibration
frequency results in an LPRM Update uncertainty which, when combined
with the other uncertainties which comprise the total TIP readings
uncertainty, yields a total TIP readings nodal power uncertainty of
less than the allowed GETAB uncertainty. Thus the change in LPRM
[[Page 64391]]
calibration frequency will not affect the MCPR Safety Limit.
The LPRMs are utilized as input to the Average Power Range
Monitor (APRM) and Rod Block Monitor (RBM) systems. The primary
safety function of the APRM system is to initiate a scram during
core-wide neutron flux transients before the actual core-wide
neutron flux level exceeds the safety analysis design basis. This
prevents fuel damage from single operator errors or equipment
malfunctions. The APRMs are calibrated at least once per week to the
plant heat balance, utilize a radially and axially diverse group of
LPRMs as input and are utilized to detect changes in average, not
local, power changes. Therefore, the effects of changing the LPRM
calibration frequency on the APRM system responses will be minimal
due to any individual LPRM drift being practically canceled out (due
to diversity of input) and/or due to the frequent recalibration of
the APRMs to an independent power calculation (the heat balance).
Thus, changing the LPRM calibration frequency will not impact the
capability of the APRM system to perform the scram function, and
there is no impact on transient delta-CPRs.
The RBM system is utilized in the mitigation of a Rod Withdrawal
Error (RWE) event. The RBM system is designed to prevent the
operator from increasing the local power significantly when
withdrawing a control rod. Under Average Power Range Monitor - Rod
Block Monitor Technical Specifications/Maximum Extended Load Line
Limit Analysis (ARTS/MELLLA) on each selection of a control rod, the
average of the assigned, unbypassed LPRMs is adjusted to equal a
100% reference signal for each of the two RBM channels. Each RBM
channel automatically limits the local thermal margin changes by
limiting the allowable change in local average neutron flux to the
RBM setpoint. If the local average neutron flux change is greater
than that allowed by the RBM setpoint, within either RBM channel,
the rod withdrawal permissive is removed preventing further rod
movement. Since the change in local neutron flux is calculated from
the change in the average of the LPRM readings, and calibrated on
every rod selection to the reference signal, offsets in individual
LPRM readings due to calibration differences are effectively
eliminated for a given RBM setpoint. Therefore, the constraints on
the withdrawal of any given rod are unchanged, and there will not be
any increase in RWE delta-CPR.
Since the MCPR Safety Limit is unaffected and the delta-CPR
values are unchanged, the cycle CPR Operating limits are unchanged
due to this TS change. Therefore, the proposed change in the
frequency of LPRM signal calibration does not result in 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 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: May 3, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would revise TS Surveillance Requirements
4.6.5.3.a and 4.6.5.4.a to modify specific requirements to perform
surveillance flow testing of the Standby Gas Treatment and Reactor
Enclosure Recirculation Systems from monthly to quarterly.
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 Technical Specifications (TS) changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
The proposed TS changes do not involve any physical changes to
plant systems or equipment. The proposed TS changes only change the
Surveillance Requirements (SRs) surveillance test frequency
pertaining to flow testing of the SGTS and RERS from monthly to
quarterly. The periodic surveillance test frequencies provide
adequate assurance that the equipment tested will remain in an
operable condition. The test frequency interval for the flow testing
of the SGTS and RERS was determined from the regulatory position in
USNRC Regulatory Guide 1.52, ``Design, Testing, and Maintenance
Criteria for Post Accident Engineered-Safety-Feature Atmosphere
Clean-up System Air Filtration and Adsorption Units of Light-Water-
Cooled Nuclear Power Plants''. As stated in Regulatory Position
C.4.d, ''... each Engineered Safety Feature (ESF) atmosphere cleanup
train should be operated at least 10 hours per month, with the
heaters on (if so equipped), in order to reduce the buildup of
moisture on the absorbers and HEPA filters.''
System operation on a monthly basis for the purpose of
preventing moisture buildup on the absorbers as described in R.G.
1.52 is not required at Limerick due to the continuous dry
instrument air purge described previously in the Safety Assessment
section of this submittal. Therefore a change in the interval
between tests from monthly to quarterly will not result in moisture
accumulation which would reduce the capability of the absorber to
remove the iodine species from the exhaust air flow stream.
The SGTS components are common to both units and must be run
with the associated RERS for the surveillance test for each unit.
The currently specified test frequency results in the SGTS being run
at least twice per month or as many as eight (8) times per quarter
for this surveillance, in addition to other required system
surveillance tests which require the use of the components in this
system. A change in surveillance test frequency from monthly to
quarterly would reduce the wear on system components and thereby
reduce the associated system downtime for maintenance and repairs.
The consequent increased availability provides greater assurance
that the system will be able to perform its mitigation function
following any postulated accident.
Surveillance test frequency on a quarterly interval is
considered adequate to verify operability, as demonstrated by the
required quarterly test interval for other equipment important to
safety which have a similar function, such as the requirement for
quarterly verification of the isolation time of the secondary
containment and refueling area isolation valves, as required by LGS
TS Sections 4.6.5.2.1 and 4.6.5.2.2.
Therefore, the proposed TS changes do not involve an increase in
the probability or consequences of an accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed TS changes only involve changes to the frequency in
which the specified surveillances tests are performed. The proposed
TS changes do not physically change the design or intended function
of the systems, structures, or components associated with the SGTS
or RERS. There will be no change to the existing redundancy of
systems and components. The proposed change in surveillance test
frequency will not introduce the possibility of any failure
mechanisms of a different type than those already evaluated in the
SAR. The existing components will not be used in any different
manner and no new components will be added. Therefore with no
physical changes and no new or different manner of system operation,
no new failure mechanisms or equipment failure modes are created.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
The margin of safety as defined in the LGS TS Bases has not been
reduced. The specific basis for the 31 day surveillance interval is
not given in the LGS TS Bases section nor in the LGS UFSAR Sections
6.5.1 or 9.4.2 which discuss the subject systems. However,
Regulatory Position C.4.d of Regulatory Guide 1.52, Revision 2,
relating to maintenance requirements, recommends:
Each ESF atmosphere cleanup train should be operated
at least 10 hours per month, with the heaters on (if so equipped),
in order to reduce the buildup of moisture on the absorbers and HEPA
filters.''
[[Page 64392]]
The Bases for Surveillance Requirements (SR) 3.6.4.3.1 in the
Standard Technical Specifications for General Electric Plants, BWR/
4, which corresponds to the subject LGS TS test, also notes the need
for ten (10) hours of operation per month for elimination of
moisture in the filters.
The basis for the requirement for a monthly test with the
heaters energized is clearly related to the desired elimination of
moisture in the filters and absorbers. However, LGS UFSAR Table 6.5-
2 states that LGS does not conform to R.G. 1.52, Position C.4.d
because the SGTS and RERS trains are ``continuously purged with dry
instrumentation air to prevent build-up of moisture.'' UFSAR
Sections 6.5.1.1.2 and 6.5.1.3.2 provide additional discussion of
this method of moisture control.
Therefore, the proposed TS 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of amendment request: September 27, 1996
Description of amendment request: The proposed Technical
Specifications (TS) changes would increase the Reactor Enclosure
Secondary Containment maximum inleakage rate. This change will also
impact secondary containment drawdown time and system flow rate
assumptions, thereby, affecting charcoal filter bed efficiency and post
accident dose analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed TS changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Changing the Reactor Enclosure post drawdown inleakage rate from
1250 cfm to 2500 cfm does not involve any changes to the function or
operation of any plant component or safety related system. The
Reactor Enclosure Recirculation System (RERS) and the Standby Gas
Treatment System (SGTS) will maintain their design function by
mitigating the radiological consequences of the analyzed accident
and mitigating the post LOCA temperatures within the Reactor
Enclosures. No analyzed accident initiating events are impacted, no
new accident initiators are created, and no new failure modes are
created. There are no changes to the redundancy, separation, quality
assurance or fire protection requirements for the associated
components and systems.
The proposed changes to the LGS adsorber bed residence time will
no longer fully meet the literal design guidance provided in
Regulatory Guide (RG) 1.52, ``Design, Testing, and Maintenance
Criteria for Post Accident Engineered-Safety-Feature Atmosphere
Cleanup System Air Filter and Adsorption Units of Light-Water-Cooled
Nuclear Power Plants,'' Revision 2, March 1978. This is because
LGS's unique, yet more conservative, adsorber bed design is not
addressed by the RG residence time design guidance. However, the LGS
SGTS charcoal adsorbers still conform to the design function
described in RG 1.52, based on the following: The LGS design with
increased inleakage will continue to conform to the three conditions
specified by RG 1.52, Position C.6.a, in order to maintain an
assigned decontamination efficiency of 99%; there is a conservative
amount of charcoal adsorber material provided by the LGS design,
based on calculations performed in accordance with RG 1.3
``Assumptions Used For Evaluating The Potential Radiological
Consequences of a Loss of Coolant Accident For Boiling Water
Reactors; and the LGS charcoal bed design is more conservative than
the RG 1.52 design guidance, based on data (i.e., Iodine Penetration
vs. Air Velocity) published by the charcoal manufacturer.
Therefore, the probability of occurrence and the consequences of
a malfunction of equipment important to safety is not increased.
Also, the probability of occurrence of an accident previously
evaluated is not increased. However, the proposed changes do affect
the leak tightness of the Unit 1 and Unit 2 Reactor Enclosure, which
increases the consequences of a postulated accident previously
evaluated.
Changing the Reactor Enclosure post drawdown inleakage rate from
1250 cfm to 2500 cfm will result in an increase in the calculated
LOCA/LOOP Design Basis Accident (DBA) off-site and on-site doses. 10
CFR Part 100, and 10 CFR Part 50 Appendix A, General Design Criteria
(GDC) 19, establish reference dose values used to determine site
suitability and provide reasonable assurance that the facility can
be operated following the analyzed accident without undue risk to
the health and safety of the public. The proposed TS changes will
increase the SGTS drawdown time from 2 minutes and 20 seconds to 15
minutes and 30 seconds. The drawdown time increase will not prevent
the RERS/SGTS from performing all of their safety related functions.
However, because it is conservatively assumed that all radioactive
material released during the drawdown period is unfiltered, and
because the drawdown period has been extended whereby more
unfiltered radioactive material is assumed to be released following
the DBA, there is a corresponding increase in the calculated
Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and
Control Room doses. It is also assumed that the SGTS exhausts at the
maximum inleakage rate throughout the entire DBA evaluation period
(i.e., 30 days) where an increase in the maximum inleakage rate
would also contribute to higher postulated EAB, LPZ, and Control
Room doses. However, the proposed calculated doses do not exceed 10
CFR Part 100, or 10 CFR Part 50, Appendix A, DGC 19 reference doses.
Since the proposed doses resulting from the changes remain below
10 CFR Part 100, and 10 CFR Part 50, Appendix A, these proposed
changes will not significantly increase the consequences of an
accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Changing the Reactor Enclosure post drawdown inleakage rate from
1250 cfm to 2500 cfm is not an accident initiator nor does it result
in the occurrence of an accident. The changes do not affect the
function or operation of any plant component or safety related
system nor do they create any new failure modes.
In addition, the proposed changes do not involve any changes to
the function or operation of any plant system or component nor will
they adversely affect the Reactor Enclosure post LOCA environmental
conditions. Furthermore, these changes will not create any new or
different failure modes for the equipment important to safety within
the Reactor Enclosure Secondary Containment.
Therefore, the possibility of an accident of a different type or
a different type of malfunction of equipment important to safety
than previously evaluated is not created.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
Changing the Reactor Enclosure post drawdown inleakage rate from
1250 cfm to 2500 cfm will result in reducing the margin of safety as
defined in the LGS Updated Final Safety Analysis Report (UFSAR)
relative to the off-site and on-site doses following a LOCA/LOOP
DBA, and an increase of the UFSAR specified system drawdown time.
From a system perspective, increasing the SGTS drawdown time from 2
minutes and 20 seconds to 15 minutes and 30 seconds will not prevent
the RERS/SGTS from performing all of their safety related functions.
There will be a postulated increase in the corresponding EAB, LPZ,
and Control Room doses, since it is assumed that fuel damage occurs
coincident with the LGS DBA (i.e, at time = 0), all radioactive
material released during the drawdown time is unfiltered, and the
drawdown time is proposed to be extended whereby more unfiltered
radioactive material could be released. It is also assumed that the
SGTS exhausts at the maximum inleakage rate throughout the entire
DBA evaluation period (i.e., 30 days) where an increase in the
maximum inleakage
[[Page 64393]]
rate would also contribute to higher postulated EAB, LPZ, and
Control Room doses. However, these calculated doses will remain
below 10 CFR Part 100, and 10 CFR Part 50, Appendix A, GDC 19
reference doses.
Therefore, these 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: Pottstown Public Library, 500
High Street, Pottstown, PA 19464
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, PA 19101
NRC Project Director: John F. Stolz
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: October 1, 1996
Description of amendment request: The proposed amendment would
allow for a one-time extension of the surveillance intervals for the
containment isolation valve (CIV) seat leakage test, the isolation
valve seal water test, the boron injection tank leakage test, the
containment spray nozzle test, and the city water backup to the
auxiliary boiler feed pump test. These tests would be performed during
the refueling outage scheduled to begin in April 1997.
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:
Regarding the Containment Isolation Valve seat leakage and
Isolation Valve Seal Water tests:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The probability of a previously evaluated accident will
not increase because CIV leakage does not provide any role in
accident initiation. The CIVs provide containment isolation
following a design basis accident.
The consequences of an accident previously evaluated will not
significantly increase because the CIV leakage measurements contain
significant margin to a more restrictive criteria based on the
requested surveillance interval extension. As discussed in Section
II, ``Evaluation of Changes,'' [see application dated October 1,
1996] based on an evaluation of past CIV leak tests, the proposed
change will not result in an increase in containment leakage because
the measured leakage in previous CIV leak tests shows large margin
to a more restrictive criteria based on the requested surveillance
interval extension. Also, the latest test of IVSWS [isolation valve
seal water system] satisfied the established acceptance criteria.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any previously
evaluated. The proposed change only provides for a relatively short,
one-time extension of the current leak-test interval for certain
containment isolation valves. The proposed change does not involve
the addition of any new or different type of equipment, nor does it
involve operating equipment required for safe operation of the
facility in a manner different from that addressed in the Final
Safety Analysis Report. Therefore, the proposed change will not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed amendment does not involve a significant reduction
in a margin of safety. The proposed change, for a one-time extension
of the test interval, will not result in a significant reduction in
a margin of safety because the test interval is being extended by
only a short period and the measured leakage in previous CIV leak
tests shows large margin to a more restrictive criteria based on the
surveillance interval extension. In addition, the online leakage
monitoring capability of the WCCPPS [weld channel containment
penetration pressurization system] helps ensure that changes in CIV
leakage during the extension period will be detected. Therefore,
this change does not create a significant reduction in a margin of
safety.
Regarding the Boron Injection Tank (BIT) leakage test:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed change will not degrade the
integrity of the BIT piping outside containment because no time
dependent failure trends were observed in the review of past test
results. The probability of a previously evaluated accident will not
be increased because BIT leakage does not provide any role in
accident prevention. The BIT leakage test only verifies that the BIT
and associated piping meet specified leakage limits.
The consequences of an accident previously evaluated will not
significantly increase because the BIT leakage test results show
large margins to the allowable leakage limit.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any previously
evaluated. The proposed change does no[t] involve the addition of
any new or different type of equipment, nor does it involve
operating equipment required for safe operation of the facility in a
manner that's different from that addressed in the Final Safety
Analysis Report. Also, the increased surveillance interval (one-time
only) will not adversely affect the integrity of the BIT piping and
will not result in any new failure modes. Therefore, the proposed
change will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed license amendment does not involve a significant
reduction in a margin of safety. Because of the large margin between
the previous test and the allowable leak rate limit, the proposed
change, for a one-time extension of the test interval, for the BIT
leakage test does not adversely affect the performance of any safety
related system, component, and does not result in increased severity
of any of the accidents considered in the Final Safety Analysis
Report. Based on past test results, the one-time extension of the
leak test interval does not involve a significant reduction in a
margin of safety.
Regarding the Containment Spray Nozzle test:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. As discussed in Section II, ``Evaluation of
Changes,'' [see application dated October 1, 1996] based on an
evaluation of past test results the proposed change will not degrade
the reliability of the Containment Spray Nozzles because no time
dependent failure trends were observed in the data review. The
probability of a previously evaluated accident will not be increased
because the Containment Spray Nozzles do not provide any role in
accident prevention. The Containment Spray Nozzles provide a uniform
spray distribution for containment cooling following postulated
post-accident conditions.
The consequences of an accident previously evaluated will not
increase because the Containment Spray Nozzle reliability is not
degraded by this change.
[[Page 64394]]
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any previously
evaluated. The proposed change does not involve the addition of any
new or different type of equipment, nor does it involve operating
equipment required for safe operation of the facility in a manner
that is different from that addressed in the Final Safety Analysis
Report. Also, the increased surveillance interval (one-time only)
w[i]ll not adversely affect the functioning of the Containment Spray
Nozzles and will not result in any new failure modes. Therefore, the
proposed change will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed license amendment does not involve a significant
reduction in a margin of safety. The proposed change, for a one-time
extension of the test interval, for the Containment Spray Nozzles
does not adversely affect the performance of any safety related
system, component, or instrument, or safety system setpoints and
does not result in increased severity of any of the accidents
considered in the Final Safety Analysis Report. Based on past test
results, the one-time extension of the functional test interval will
not adversely affect the functioning of the Containment Spray
Nozzles. Therefore, this change does not create a significant
reduction in a margin of safety.
Regarding the City Water Backup test:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Response:
The proposed license amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated. The proposed change will not degrade the
reliability of the City Water Backup Supply Valves for the AFW
[auxiliary feedwater] System because no time dependent failure
trends were observed in the review of past test results. The
probability of a previously evaluated accident will not increase
because the City Water Backup Supply Valves for the AFW System do
not provide any role in accident prevention. The City Water Backup
Supply Valves for the AFW System only provides a diverse source of
water for the AFW system.
The consequences of an accident previously evaluated will not
significantly increase because the City Water Backup Supply Valves
for the AFW System are not assumed to function to mitigate any
analyzed accident.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any accident previously
evaluated?
Response:
The proposed license amendment does not create the possibility
of a new or different kind of accident from any previously
evaluated. The proposed change does not involve the addition of any
new or different type of equipment, nor does it involve operating
equipment required for safe operation of the facility in a manner
that is different from that addressed in the Final Safety Analysis
Report. Also, the increased surveillance interval (one-time only)
will not adversely affect the functioning of the City Water Backup
Supply Valves for the ABFP [auxiliary boiler feedpump] and will not
result in any new failure modes. Therefore, the proposed change will
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed amendment does not involve a significant reduction
in a margin of safety. The proposed change, for a one-time extension
of the test interval, for the City Water Backup Supply Valves for
the ABFP does not adversely affect the performance of any safety
related system, component, or instrument, or safety system setpoints
and does not result in increased severity of any of the accidents
considered in the Final Safety Analysis Report. Based on past test
results, the one-time extension of the functional test interval will
not adversely affect the functioning of the City Water Backup Supply
Valves for the AFW System. Therefore, this change does not create a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: S. Singh Bajwa, Acting
Public Service Electric & Gas Company, Docket No. 50-354, Hope
Creek Generating Station, Salem County, New Jersey
Date of amendment request: October 25, 1996
Description of amendment request: The proposed change to Hope Creek
Technical Specification (TS) 3/4.1.3.5, ``Control Rod Scram
Accumulator'', would: 1) permit a separate entry into a Technical
Specification action statement for each inoperable control rod; 2)
provide more specific applicability for required actions in operational
condition 1 or 2 with one inoperable control rod scram accumulator
(reactor pressure of greater than or equal to 900 psig would be
specified); 3) provide more specific actions (verify charging water
pressure) for two or more inoperable control rod scram accumulators and
reactor pressure is greater than or equal to 900 psig; 4) provide more
specific actions when reactor pressure is less than 900 psig and one or
more control rod scram accumulators are inoperable (verify insertion of
control rods associated with inoperable accumulators and verify that
charging water header pressure is greater than or equal to 940 psig);
and 5) provide specific actions in operational condition 5 with one or
more withdrawn control rods inoperable; and 6) eliminate the
requirements to perform a 18-month channel functional test of the leak
detectors and the 18-month channel calibration of the pressure
detectors.
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 change incorporates the appropriate content of the improved
BWR/4 Standard Technical Specifications, NUREG-1433, for Control Rod
Scram Accumulators.
The proposed Technical Specification and required Action
completion times are consistent with or more conservative than those
approved for use in the improved Technical Specifications for
inoperable control rod scram accumulators. In addition, the proposed
surveillance requirements for the control rod scram accumulators are
sufficient to adequately demonstrate operability as stated in the
Bases for the improved Technical Specifications. Further, the
proposed changes enhance the current Hope Creek Technical
Specifications by reflecting improved techniques collectively
learned by the industry. Therefore, the proposed changes do not
significantly increase the risk or consequences of any accidents
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Neither the mechanism for initiating or completing a scram is
modified by this proposed change. There are no physical changes to
plant equipment proposed in the application. The proposed change
does not create a means by which the scram function could be impeded
or prevented. The proposed change is functionally equivalent to the
current Technical Specifications, but provides additional
operational flexibility to diagnose and resolve equipment issues
that do not impact operability of the control rods before taking
proscriptive actions which
[[Page 64395]]
result in significant plant transients (i.e. full power scram).
3. The proposed change does not involve a significant reduction
in a margin of safety.
The operability of the accumulators and the scram function of
the control rod drive system protects the Safety Limit Minimum
Critical Power Ratio as well as the 1% cladding plastic strain fuel
design limit. The proposed change does not reduce a margin of safety
as defined in the Bases of the Technical Specification since the
proposed change does not affect the maximum allowable scram times
for control rods, nor does it change the maximum allowable number or
minimum separation of inoperable control rods. The proposed change
does not modify any instrument setpoints or functions. The proposed
change will either maintain the present margins of safety or
increase them, by reducing the need for unnecessary challenges to
the reactor protection system and resulting plant shutdowns, while
still maintaining the capability to complete a reactor scram.
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: Pennsville Public Library, 190
S. Broadway, Pennsville, New Jersey 08070
Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
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: October 29, 1996
Description of amendment request: The proposed amendment would
revise the mode of applicability for the motor-driven auxiliary
feedwater (AFW) pump actuation on opening of the main feedwater (MFW)
pump breakers to correct an error introduced during Amendment No. 61.
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 less restrictive changes discussed in Section C.1 [of the
licensee's application] do not involve a significant hazards
consideration as discussed 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 correct an error which was introduced in Amendment No.
61 to the Ginna Station technical specifications. The changes revert
the mode of applicability for the motor-driven AFW pump actuation on
the opening of the MFW pump breakers to what existed previously. The
change is essentially correction of a typographical error that was
caused through use of the electronic version of NUREG-1431 in
preparation of the Ginna Station ITS [Improved Technical
Specifications]. There have been no subsequent plant modifications
or changes to the accident analysis which would invalidate the
previous NRC acceptance of only requiring this Function above 5%
power. The accident analyses do not credit automatic initiation of
AFW on MFW pump trip in MODE 2. As such, these changes 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 which existed prior to
Amendment No. 61. 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.
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 there have been no subsequent plant modifications or
changes to the accident analysis which would invalidate the previous
NRC acceptance of only requiring this Function above 5% power. 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 Project Director: S. Singh Bajwa, Acting
Rochester Gas and Electric Corporation, Docket No. 50-244, R. E.
Ginna Nuclear Power Plant, Wayne County, New York
Date of amendment request: October 29, 1996
Description of amendment request: The proposed amendment would
revise the Required Actions for the auxiliary feedwater (AFW) pump
actuation on Steam Generator Level (SG) - Low Low logic to be
consistent with those specified in NUREG-1431.
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 less restrictive changes discussed in Section C.1 [of the
licensee's application] do not involve a significant hazards
consideration as discussed 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 with respect to the Required Actions for AFW actuation on SG
Level - Low Low logic provide consistency with NUREG-1431 by
requiring an inoperable channel to be placed in the tripped
condition within 6 hours. The affected logic then requires 1 of 2
channels in order to actuate such that there is no impact on any
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.
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 AFW actuation on SG Level - Low Low still remains
capable of performing its function with an inoperable channel placed
in the tripped configuration. These changes are also consistent with
those provided in NUREG-1431. 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
[[Page 64396]]
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 Project Director: S. Singh Bajwa, Acting
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: September 4, 1996
Description of amendment request: The proposed amendment to the
Technical Specifications would allow the use of four lead test
assemblies (advanced zirconium-based alloys) in the North Anna, Units 1
and 2, reactor cores.
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 the four FCF [Framatome Cogema Fuels] lead test
assemblies will not:
1.Involve a significant increase in the probability of
occurrence or the consequences of an accident previously evaluated.
The FCF lead test assemblies are very similar in design to the
Westinghouse fuel that comprises the remainder of the core. The
reload core design for North Anna cycles which incorporate the lead
test assemblies will meet all applicable design criteria. In
addition, the performance of the ECCS [emergency core cooling
system] at North Anna Units 1 and 2 will not be affected by the
insertion of the four lead test assemblies, so the criteria of 10
CFR 50.46 will be satisfied for use of these assemblies with fuel
rods, guide thimble tubes, and instrumentation tubes fabricated with
advanced zirconium-based alloys. The use of these fuel assemblies
will not result in a change to the North Anna Units 1 and 2 reload
design and safety analysis limits. The existing safety analyses
based on the resident Westinghouse fuel will remain applicable for
cycles which incorporate the lead test assemblies. Therefore,
neither the probability of occurrence nor the consequences of any
accident previously evaluated is significantly increased.
2. Create the possibility for a new or different type of
accident from any accident previously evaluated. The FCF lead test
assemblies are very similar in design (both mechanical and
composition of materials) to the resident Westinghouse fuel. North
Anna cores which incorporate the lead test assemblies will be
designed to meet all applicable design criteria and ensure that all
pertinent licensing basis criteria are met. Demonstrated adherence
to these standards and criteria precludes new challenges to
components and systems that could introduce a new type of accident.
North Anna safety analyses based on the resident Westinghouse fuel
will remain applicable for cores containing the lead test
assemblies. All design and performance criteria will continue to be
met and no single failure mechanisms have been created. In addition,
the use of these fuel assemblies does not involve any alteration to
plant equipment or procedures which would introduce any new or
unique operational modes or accident precursors. Therefore, the
possibility for a new or different kind of accident from any
accident previously evaluated is not created.
3. Involve a significant reduction in the margin of safety. The
use of the FCF lead test assemblies does not change the performance
requirements on any system or component such that any design
criteria will be exceeded, and will not cause the core to operate in
excess of pertinent design basis operating limits. North Anna reload
core designs for cycles which incorporate the lead test assemblies
will specifically evaluate any pertinent differences between the
lead test assemblies and the resident fuel, and will take into
consideration the normal core operating conditions allowed in the
Technical Specifications. Safety analyses based on the resident
Westinghouse fuel will remain applicable for cores incorporating the
FCF lead test assemblies. Analyses or evaluations will be performed
each cycle to confirm that the criteria in 10 CFR 50.46 will be met.
Therefore, the margin of safety as defined in the Bases to the North
Anna Units 1 and 2 Technical Specifications is not significantly
reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Acting Project Director: Mark Reinhart
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: November 6, 1996
Description of amendment request: The proposed changes will modify
the requirements for isolated loop startup to permit filling of a
drained isolated loop via backfill from the reactor coolant system
through partially open stop loop valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of the North Anna Power Station [in]
accordance with the proposed changes will not:
1. Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated. The
probability of occurrence of a positive reactivity addition accident
is not being increased by the proposed Technical Specification
change. The proposed restrictions on boron concentration and mixing,
reactor coolant system inventory and reactivity and count rate
monitoring provide a level of protection against reactivity addition
accidents which is equivalent to that currently in place.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated. The proposed change does not
introduce any new or unique failure modes or accident precursors.
Eliminating the operability requirements for the loop stop valve
interlocks does not create any new or different kind of accident
scenario. Loop startup accidents in the various modes of operation
have been analyzed. Operation of the loop stop valves will not
change. New requirements have been imposed for the case of
backfilling a drained loop from the reactor coolant system to ensure
that core cooling and reactivity control are preserved throughout
the backfill evolution.
3. Involve a significant reduction in any margin of safety. The
new Technical Specification loop isolation and startup requirements
for temperature, boron concentration, and shutdown margin fulfill
the function of the loop stop valve interlocks. Therefore, the
margin of safety as defined in any Technical Specification bases is
not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Mark Reinhart (Acting)
[[Page 64397]]
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of amendment request: October 31, 1996
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TS) by deleting the requirement for an annual submittal of a
description of changes made pursuant to 10 CFR 50.59. Consistent with
10 CFR 50.59(b)(2), a description of changes will subsequently be
included with the KNPP Updated Safety Analysis Report (USAR) update in
accordance with 10 CFR 50.71(e). Additionally, the proposed amendment
would correct minor administrative inconsistencies in the TS Table of
Contents and in a footnote reference.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c)The NRC staff's review is
presented below:
On August 31, 1992 (57 FR 39353), the NRC amended 10 CFR
50.59(b)(2) to reduce the regulatory burden on nuclear licensees. This
action revised the requirements for the annual submission of reports
for facility changes under 10 CFR 50.59. This action did not affect the
substance of the evaluation or the documentation required for 10 CFR
50.59 type changes. It only affected the interval for submission of the
information to the NRC. Instead of submitting the information annually,
the information can be submitted on a refueling cycle basis, provided
the interval between successive reports does not exceed 24 months.
In order to take advantage of this reduction in regulatory burden,
the licensee has proposed an amendment to remove the submittal of a
report of facility changes under 10 CFR 50.59 from the Technical
Specification list of annual reporting requirements. Additionally, the
licensee has proposed corrections to minor administrative
inconsistencies in the TS Table of Contents and in a footnote
reference. The proposed changes are administrative only and do 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.
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
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.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, Massachusetts
Date of application for amendment: May 1, 1996, as supplemented
August 12, 1996.
Brief description of amendment: The amendment approves relocation
of the administrative controls related to the quality assurance review
and audit requirements of Section 6, Technical Specifications 6.5.B.8,
``Nuclear Safety Review and Audit Committee-Audits,'' from the Pilgrim
Station Technical Specifications to the Boston Edison Quality Assurance
Manual (BEQAM). This change is in accordance with the guidance
contained in NRC Administrative Letter 95-06, ``Relocation of Technical
Specification Administrative Controls Related to Quality Assurance.''
In addition, the Safety Evaluation includes the NRC staff review and
approval of the BEQAM changes in support of this amendment.
Date of issuance: November 12, 1996
Effective date: November 12, 1996
Amendment No.: 168
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28605) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 12, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion
Nuclear Power Station, Units 1 and 2, Lake County, Illinois
Date of application for amendments: August 29, 1996, as
supplemented on September 20, 1996, and October 4, 1996.
Brief description of amendments: The amendments change the
Technical Specifications to implement 10 CFR Part 50, Appendix J,
Option B, by referring to Regulatory Guide 1.163, ``Performance-Based
Containment Leakage-Test Program,'' with an exception as detailed in
the licensee's application.
Date of issuance: November 12, 1996
Effective date: Immediately, to be implemented within 30 days.
Amendment Nos.: 175 and 162
[[Page 64398]]
Facility Operating License Nos. DPR-39 and DPR-48: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52964). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 12, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of application for amendment: August 14, 1996, as supplemented
October 18, 1996, and related application of January 18, 1996
Brief description of amendment: The amendment revises the technical
specifications (TS) to allow one-cycle deferral of the inspection of
reactor coolant pump (RCP) flywheels.
Date of issuance: November 7, 1996
Effective date: November 7, 1996
Amendment No.: 175
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 24, 1996 (61
FR 50054) The October 18, 1996, letter provided an updated TS page.
This change was within the scope of the original application and did
not change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendment is contained in a Safety Evaluation dated November 7, 1996.No
significant hazards consideration comments received: No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423
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: December 14, 1994, as
supplemented by letters dated May 16 and August 29, 1996
Brief description of amendments: The amendments will incorporate
guidance and recommendations for diesel generators contained in NUREG-
1366, ``Improvements to Technical Specifications Surveillance
Requirements,'' Generic Letter (GL) 93-05, ``Line-Item Technical
Specifications Improvements to Reduce Surveillance Requirements for
Testing During Power Operations,'' GL 94-01, ``Removal of Accelerated
Testing and Reporting Requirements for Emergency Diesel Generators,''
and NUREG-1431, ``Revised Standard Technical Specifications for
Westinghouse PWRs.''
Date of issuance: November 12, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment Nos.: 170 and 152
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 5, 1996 (61 FR
28612) The August 29, 1996, letter provided clarifying information that
did not change the scope of the December 14, 1996, application and the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 12, 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
Entergy Gulf States, Inc., Cajun Electric Power Cooperative, and
Entergy Operations, Inc., Docket No. 50-458, River Bend Station,
Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: August 1, 1996
Brief description of amendment: The amendment revises the technical
specifications to incorporate requirements for limiting the time that
the hydrogen mixing isolation valves on the drywell are open. The
amendment also changes the time from 7 days to 31 days to determine the
cumulative time the valves are open.
Date of issuance: November 12, 1996
Effective date: November 12, 1996
Amendment No.: 89
Facility Operating License No. NPF-47. The amendment revised the
Technical Specifications/operating license.
Date of initial notice in Federal Register: September 25, 1996 (61
FR 50343) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 12, 1996.No significant
hazards consideration comments received. No.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, LA 70803
Entergy Operations, Inc., System Energy Resources, Inc.,
SouthMississippi Electric Power Association, and Entergy
Mississippi, Inc., Docket No. 50-416, Grand Gulf Nuclear Station,
Unit 1, Claiborne County, Mississippi
Date of application for amendment: May 9, 1996, as supplemented by
letter dated August 27, 1996.
Brief description of amendment: The amendment changed Surveillance
Requirements (SRs) 3.4.4.3, Safety/Relief Valves, 3.5.1.7, Automatic
Depressurization System Valves, and 3.6.1.6.1, Low-Low Set Valves, of
the Technical Specifications and allows the licensee to perform the
surveillance of the relief mode of operation of the safety/relief
valves on the main steam lines without physically lifting the disk of a
valve off the seat at power. The changes stated that the required
operation of the valve to verify is that the relief-mode actuator
strokes when the valve is manually actuated and the frequency of the
surveillances are in accordance with the inservice testing program for
the valves.
Date of issuance: November 18, 1996
Effective date: November 18, 1996
Amendment No: 130
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47971) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 18, 1996. No
significant hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, MS 39120.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: June 3, 1996, as supplemented
October 23, 1996
Brief description of amendment: The amendment clarifies a
restriction on shutdown margin monitor operability while changing
operational modes, so that it only limits reactivity changes caused by
boron dilution and rod withdrawal. The amendment also corrects a
technical specification numerical reference so that the specification
number cited is in agreement with Amendment 99, dated December 29,
1994.
Date of issuance: November 14, 1996
[[Page 64399]]
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendment No.: 131
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 20, 1996 (61 FR
31559) The October 23, 1996, letter provided clarifying information
that did not change the scope of the June 3, 1996, application and the
initial proposed no significant hazards consideration determination.The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated November 14, 1996No significant hazards
consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, 574 New London Turnpike,
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49
Rope Ferry Road, Waterford, CT 06385.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: May 30, 1996
Brief description of amendment: The proposed change to the
anticipated transient without scram recirculation pump trip logic for
the James A. Fitzpatrick Nuclear Power Plant allows for a high pressure
trip setpoint which is dependent upon the number of safety/relief
valves which are out of service.
Date of issuance: November 7, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 237
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34896) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 7, 1996.No significant
hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Power Authority of the State of New York, Docket No. 50-333, James
A. FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of application for amendment: May 30, 1996, as supplemented
October 17, and November 8, 1996
Brief description of amendment: The proposed amendment changes the
FitzPatrick safety limit minimum critical power ratio from its current
value of 1.07 for two recirculation loop operation to 1.09 and from
1.08 to 1.10 for single recirculation loop operation for the Cycle 13
operation.
Date of issuance: November 14, 1996
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 238
Facility Operating License No. DPR-59: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 3, 1996 (61 FR
34896) The October 17 and November 8, 1996 letters provided
supplemental information that did not change the initial no significant
hazards consideration determination.The Commission's related evaluation
of the amendment is contained in a Safety Evaluation dated November 14,
1996.No significant hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama
Date of amendment request: August 23, 1996, as supplemented by
letters dated September 16, November 6, 11 and 14, 1996
Brief description of amendment: The amendment changes the Technical
Specifications (TS) to allow installation of laser welded elevated
tubesheet sleeves. Specifically, the amendment is for one cycle only
for Farley Unit 2. Permanent, generic TS changes for Westinghouse laser
welded sleeves for both units will be submitted prior to the next Unit
1 refueling outage currently scheduled for spring 1997.
Date of issuance: November 20, 1996
Effective date: As of the date of issuance to be implemented within
30 days
Amendment No.: 117
Facility Operating License No. NPF-8: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 11, 1996 (61
FR 47982) The September 16, November 6, 11 and 14, 1996, letters
provided clarifying information that did not change the scope of the
August 23, 1996, application and the initial proposed no significant
hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 20, 1996.No significant hazards
consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: July 17, 1995.
Brief description of amendments: These amendments revise the
frequency of surveillance requirements for certain plant protective
system instrumentation contained in Technical Specifications (TS)
3.3.1, ``Reactor Protective System (RPS) Instrumentation - Operating,''
TS 3.3.2, ``Reactor Protective System (RPS) Instrumentation -
Shutdown,'' TS 3.3.3, ``Control Element Assembly Calculators (CEACs),''
TS 3.3.4, ``Reactor Protective System (RPS) Logic and Trip
Initiation,'' TS 3.3.5, ``Engineered Safety Features Actuation System
(ESFAS) Instrumentation,'' and TS 3.3.6, ``Engineered Safety Features
Actuation System (ESFAS) Logic and Manual Trip.''
Date of issuance: November 18, 1996
Effective date: November 18, 1996, to be implemented within 30 days
of the date of issuance.
Amendment Nos.: Unit 2 - 133 ; Unit 3 - 122
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 30, 1995 (60 FR
45185) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 18, 1996.No significant
hazards consideration comments received: No.Temporary
Local Public Document Room location: Science Library, University
of California, P. O. Box 19557, Irvine, California 92713
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: September 4, 1996
Brief description of amendment: This amendment revises Technical
[[Page 64400]]
Specification (TS) 6.2.3, ``Facility Staff Overtime,'' by removing
specific overtime limits and working hours and by adding procedural
controls to perform a monthly review of overtime hours.
Date of issuance: November 8, 1996
Effective date: November 8, 1996, to be implemented not later than
90 days after issuance
Amendment No.: 212
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 9, 1996 (61 FR
52970) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 8, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Toledo, William
Carlson Library, Government Documents Collection, 2801 West Bancroft
Avenue, Toledo, Ohio 43606
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application for amendment: July 18, 1996
Brief description of amendment: The amendment adopts ASTM D-3803-
1989 as the laboratory testing standard for charcoal samples from the
charcoal adsorbers in the auxiliary/fuel building emergency exhaust
system.
Date of issuance: November 13, 1996
Effective date: November 13, 1996, to be implemented within 30 days
of the date of issuance.
Amendment No.: 118
Facility Operating License No. NPF-30: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 14, 1996 (61 FR
42285) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 13, 1996.No significant
hazards consideration comments received: No.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Dated at Rockville, Maryland, this 26th day of November 1996.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II Office of Nuclear Reactor
Regulation
[Doc. 96-30714 Filed 12-3-96; 8:45 am]
BILLING CODE 7590-01-F
