[Federal Register Volume 59, Number 244 (Wednesday, December 21, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-31196]
[[Page Unknown]]
[Federal Register: December 21, 1994]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating LicensesInvolving
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 28, 1994, through December 9, 1994.
The last biweekly notice was published on December 7, 1994.
Notice Of Consideration Of Issuance Of Amendments To Facility
Operating Licenses, Proposed No Significant Hazards Consideration
Determination, And Opportunity For A Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Rules Review and
Directives Branch, Division of Freedom of Information and Publications
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555, and should cite the publication date and page
number of this Federal Register notice. Written comments may also be
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the NRC Public
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC.
The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By January 13, 1995, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC and at the local public
document room for the particular facility involved. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or an Atomic Safety and Licensing Board, designated by
the Commission or by the Chairman of the Atomic Safety and Licensing
Board Panel, will rule on the request and/or petition; and the
Secretary or the designated Atomic Safety and Licensing Board will
issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) the nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555, and to the
attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document room for
the particular facility involved.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit
Nos. 1, 2, and 3, Maricopa County, Arizona
Date of amendment requests: November 2, 1994
Description of amendment requests: The proposed amendment would
delete the Condenser Vacuum Exhaust release point reference on Figure
5.1-3 and combine it with the Plant Vent Exhaust release point on the
revised Figure 5.1-3. In addition to the figure change, Bases Section
3/4.3.3.6 is amended to note the deletion of radiation monitor RU-142
and the relocation of RU-144 and RU-146 from Table 3.3-13 (previously
deleted) to the Offsite Dose Calculation Manual (ODCM).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis about the issue of no significant hazards
consideration, which is presented below:
Standard 1--Does the proposed change involve a significant
increase in the probability or consequences of an accident
previously evaluated?
Combining the condenser vacuum and the plant vent exhausts has
no affect
[sic] on the operation of the radiation monitoring system or its
intended functions. Routing of the condenser vacuum exhaust to the
plant vent exhaust is in the same area as the old system and does
not affect accident initiation or consequences. The change has no
affect
[sic] on the operation of the plant. The radiation monitors
affected by this change do not provide engineered safety features or
protection system actuation signals. Therefore, the change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Standard 2 -- Does the proposed change create the possibility of
a new or different kind of accident from any accident previously
evaluated?
The overall system is designed to assist the operators in
evaluating and controlling the radiological consequences of normal
plant operations, anticipated operational occurrences, and
postulated accidents. The change does not affect the way the system
is operated. Therefore, the change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Standard 3 -- Does the proposed change involve a significant
reduction in a margin of safety?
Combining of the condenser vacuum and plant vent exhaust into a
single release path does not involve a significant reduction in a
margin of safety. The change involves the removal of one high range
monitor in the condenser vent, however, its function is provided by
the high range monitor in the plant vent. The ranges of the monitors
are the same. The existing plant effluent radiation monitors will
serve to monitor both the plant and condenser air removal system
effluent. The normal range monitors have the ability to adequately
detect radiation over five decades and these monitors will stay in
place and they have the ability to perform the anticipated radiation
release detection. Therefore, the proposed change does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Basis for proposed no significant hazards consideration
determination: I11Attorney for licensees: Nancy C. Loftin, Esq.,
Corporate Secretary and Counsel, Arizona Public Service Company, P.O.
Box 53999, Mail Station 9068, Phoenix, Arizona 85072-3999
NRC Project Director: Theodore R. Quay
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick
County,North Carolina
Date of amendments request: November 16, 1994 Description of
amendments request: The proposed Technical Specification (TS) change
would (1) revise TS 4.6.1.2 by removing the schedular requirements for
Type A overall integrated leakage rate tests to be performed at 40 plus
or minus 10 month intervals and replacing the acceptance criteria for
these Type A integrated leakage rate tests with a reference to the
containment integrated leakage testing requirements of Appendix J to 10
CFR Part 50, (2) delete TS 4.6.1.2.a through TS 4.6.1.2.c because they
are no longer needed, (3) revise TS 4.6.1.2.h to remove the prohibition
against applying TS 4.0.2 to the 40 plus or minus 10 month integrated
leakage rate test frequency, (4) delete Unit 1 one-time footnote *
located on TS page 3/4 6-3A and on Table 4.6.1.2-1 listed on TS page 3/
4 6-3B since the exception provision has expired, (5) delete Unit 1
one-time footnote ** located on TS page 3/4 6-3A since the exception
has expired, (6) delete Unit 2 footnote * located on TS page 3/4 6-3
because the exception constitutes an approved exemption.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
. The proposed amendments do not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The proposed amendments remove the detailed technical and
schedular information pertaining to primary containment integrated
leakage rate testing from the Technical Specifications and
references the corresponding requirements that are located in the
Appendix J to 10 CFR Part 50. As such, the proposed amendments are
an administrative change since the actual requirements for the
performance of primary containment integrated leakage rate testing
are not being changed. No safety-related equipment, safety function,
or plant operations will be altered as a result of the proposed
amendments. The change does not affect the design, materials, or
construction standards of the primary containment nor the test
methods, test acceptance criteria, or testing frequencies applicable
to primary containment integrated leakage rate testing. Based on the
above, the proposed license amendments do not create a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, no safety-related equipment, safety
function, or plant operations will be altered as a result of the
proposed change. The proposed amendments do not change the primary
containment design or the test methods, test acceptance criteria, or
testing frequencies for primary containment integrated leakage rate
testing. As such, the proposed license amendments cannot create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendments do not involve a significant
reduction in the margin of safety. The proposed amendments do not
involve any changes to the test methods, acceptance criteria, or
testing frequency for primary containment integrated leakage rate
testing. Thus, the proposed amendments will not affect the ability
of the primary containment to perform its intended safety function
and no margins of safety, as defined by the plant's accident
analyses, are impacted. Primary containment integrated leakage rate
testing will continue to be performed in accordance with the
regulatory requirements of Appendix J to 10 CFR Part 50. Based on
the above reasoning, the proposed license amendments do not involve
a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
NRC Project Director: William H. Bateman
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: September 28, 1994
Description of amendment request: Amendments will update the ``Loss
of Power'' functional unit of the Engineered Safety Features Actuation
System (ESFAS) Instrumentation tables within the Technical
Specifications for McGuire Nuclear Station.
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:
As required by 10 CFR 50.91, this analysis is provided
concerning whether the requested amendments involve significant
hazards considerations, as defined by 10 CFR 50.92. Standards for
determination that an amendment request involves no significant
hazards considerations are if operation of the facility in
accordance with the requested 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 requested amendments update the existing one-level
undervoltage protection to be exclusively for loss of voltage, and
add a second level of undervoltage protection to be exclusively for
degraded voltage.
In 48 FR 14870, the Commission has set forth examples of
amendments that are considered not likely to involve significant
hazards considerations. Example vi describes a change which either
may result in some increase to the probability or consequences of a
previously-analyzed accident or may reduce in some way a safety
margin, but where the results of the change are clearly within all
acceptable criteria with respect to the system or component
specified in the Standard Review Plan. The requested amendments are
similar to example vi in that they result in some increase to the
probability of a previously-analyzed accident, the Loss of Offsite
Power accident, but where the changes are clearly based on the
recommendations of Branch Technical Position PSB-1.
Criterion 1
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The requested amendments will involve some increase in
the probability of an accident previously evaluated. Automatic
separation from offsite power (a LOOP accident) will be more
probable because the voltage setpoints for the new relaying will be
higher than the settings for the existing relaying. The closer relay
settings are to 100% bus voltage, the more frequently actual bus
voltage can be expected to occur at or below the setpoint. The
occurrence of a LOOP presents a challenge to safety systems. More
probable (e.g., more frequent) LOOPs increase the frequency of
safety system challenges, which increases the probability of
malfunction of equipment important to safety. However, offsetting
this probability increase is a probability decrease due to the
protection of safety equipment from degraded voltage conditions,
given by the added protective relaying. The EPC system is required
to provide power for equipment used for accident mitigation and safe
shutdown. The ability of the EPC system to perform its required
safety functions will not be degraded by the implementation of this
TS change. No common failure modes are created between redundant EPC
system power trains. Therefore, the consequences of an accident or
malfunction of equipment important to safety evaluated in the SAR
are not increased.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. No new failure modes are created by the implementation of
this TS change. No accidents previously considered incredible are
made credible. The added protective relaying is expected to be as
reliable as the existing relaying. The added equipment is QA
Condition 1, and qualifications of equipment enclosures have been
maintained. Thus, the possibility of an accident or malfunction of
equipment of a different type than evaluated in the SAR will not be
created.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. The setpoints for the existing Loss
of Power protective relays are lowered by this TS change. The new
setpoints have been evaluated and will not prevent the protective
relaying from performing its required safety function. The fission
product barriers (RCS pressure boundary, containment, fuel pellets,
and cladding) are not degraded. No assumptions made in any accident
analysis are affected by the implementation of this TS change,
except as previously discussed for probability of a Loss of Offsite
Power. Therefore, the margin of safety as defined in the basis for
any Technical Specification is not decreased.
Based on the preceding analyses, Duke Power concludes that the
requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242
NRC Project Director: Herbert N. Berkow
Duquesne Light Company, et al., Docket Nos. 50-334 and 50-
412,Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport,
Pennsylvania
Date of amendment request: August 31, 1994
Description of amendment request: The proposed amendment would
revise Technical Specifications (TSs), Section 6, Administrative
Controls, and includes line-item improvements suggested by Generic
Letter 93-07. The proposed changes include the following:
1. Elimination of the references to specific frequencies for
each of the Technical Specification required audits.
2. Elimination of the references to reviews and audits of the
Emergency Plan and Security Plan.
3. Separation of the Inservice Inspection (ISI) and Inservice
Testing (IST) Programs surveillance requirements and removal of the
requirement that relief requests be granted before they are
implemented for both IST and ISI.
4. Editorial changes which were necessitated by a
reorganization.
5. Elimination of the reference to Appendix A of 10 CFR Part 55.
6. Elimination of the requirement to perform an independent fire
protection and loss prevention program inspection annually.
7. Inclusion of the Offsite Dose Calculation Manual and Process
Control Program and associated implementing procedures into the list
of required audits.
8. Updates of the Beaver Valley Power Station (BVPS) Unit 2
License Conditions to reflect completion of activities.
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 likelihood that an accident will occur is neither increased
or decreased by this proposed Technical Specification change which
only affects review and audit frequencies, removes redundancies in
the audit program, corrects editorial information, and updates the
Unit 2 license conditions. This Technical Specification change will
not impact the function or method of operation of plant equipment.
Thus, there is not a significant increase in the probability of a
previously analyzed accident due to this change. No systems,
equipment, or components are affected by the proposed change. Thus,
the consequences of a malfunction of equipment important to safety
previously evaluated in the Updated Final Safety Analysis Report are
not increased by this change.
The proposed change affects audit frequencies, types of audits
listed in the technical specifications, references for some
technical specification sections, the time frame for Inservice
Testing (IST) and Inservice Inspection (ISI) relief request
submittals, and editorial changes necessitated by an internal
reorganization. As such, the proposed change has no impact on
accident initiators or plant equipment, and therefore, does not
affect the probabilities or consequences of an accident.
Therefore, the proposed change will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed technical specification revisions do not involve
changes to the physical plant or operations. Since program audits,
organizational titles, and technical specification references do not
contribute to accident initiation, a change related to the areas
listed in the description section [***] cannot produce a new
accident scenario or produce a new type of equipment malfunction.
Therefore, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the change involve a significant reduction in a margin
of safety?
The proposed change concerns the conduct of audits, technical
specification references, ISI and IST relief request submittals,
completed License conditions, and organizational title changes and
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are not affected by this
proposed change.
Therefore, use of the proposed Technical Specification would not
involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment involves no significant hazards consideration.
Local Public Document Room location: B. F. Jones Memorial Library,
663 Franklin Avenue, Alquippa, Pennsylvania 15001.
Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg,
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW.,
Washington, DC 20037.
NRC Project Director: Walter R. Butler
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 11, 1994, as supplemented
December 2, 1994.
Description of amendment request: The requested change would remove
cycle-specific variables from the Waterford 3 Technical Specifications
(TSs) and control them under a new document called the Core Operating
Limits Report (COLR). All cycle-specific limits that are to be included
in the COLR must be calculated using NRC approved methodologies. The
proposed change is consistent with the TS line-item improvement
guidelines provided by the NRC in Generic Letter (GL) 88-16, ``Removal
of Cycle-Specific Parameter Limits From Technical Specifications,''
dated October 3, 1988.
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. Removing cycle-specific variables from the TS and placing
them into a COLR, is consistent with the NRC guidance provided in GL
88-16. These changes are administrative in nature and have no impact
on plant operation or accident analyses. The TS will continue to
require operation within the core operational limits for each cycle
reload calculated by the approved reload methodologies. If these
limits are violated, Technical Specifications will continue to
ensure that the appropriate actions are taken.
The cycle-specific evaluation demonstrates that changes in the
fuel cycle design and the corresponding COLR do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Listing the NRC approved methodologies in the COLR as opposed to
the TS Administrative Controls section is purely an administrative
change in contrast to NUREG 1432. The proposed change requires the
use of NRC approved methodologies. Listing the approved
methodologies in the TS provides the potential for an increased
licensee and NRC administrative burden without a commensurate
increase in safety or control.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed changes, to relocate the cycle-specific
variables from TS to the COLR, are administrative in nature. No
change in the design, configuration, or method of operation of the
plant is made by this amendment. The cycle-specific variables will
continue to be calculated using NRC approved methods. TS will
continue to require operation within the required core operating
limits and appropriate actions will be taken if the limits are
exceeded.
Listing the NRC approved methodologies in the COLR as opposed to
the TS Administrative Controls section is purely an administrative
change in contrast to NUREG 1432. The proposed change requires the
use of NRC approved methodologies. Listing the approved
methodologies in the TS provides the potential for an increased
licensee and NRC administrative burden without a commensurate
increase in safety or control.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The margin of safety presently provided is not affected by
removing cycle-specific core operating limits from TS. The core
limits contained in the COLR are obtained through analyses using NRC
approved methodologies. The TS still: (1) require that the core be
operated within these limits and (2) specify appropriate actions to
be taken if the limits are violated. The cycle-
specific COLR limits for future reload will also be developed
based on NRC-approved methodologies. In addition, each reload will
involve a 10CFR 50.59 safety review to assure that operation of the
unit within the cycle-specific limits will not involve a reduction
in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
Safety and Significant Hazard Determination
Based on the above safety analysis, it is concluded that: (1)
the proposed change does not constitute a significant hazards
consideration as defined by 10CFR50.92; and (2) there is a
reasonable assurance that the health and safety of the public will
not be endangered by the proposed change; and (3) this action will
not result in a condition which significantly alters the impact of
the station on the environment as described in the NRC Final
Environmental Statement.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L
Street N.W., Washington, D.C. 20005-3502
NRC Project Director: William D. Beckner
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2,
Appling County, Georgia
Date of amendment request: October 13, 1994
Description of amendment request: The proposed amendments would
revise the Hatch Technical Specifications (TS) as follows:
1. Lower the anticipated transient without scram-recirculation pump
trip (ATWS-RPT) setpoint by approximately 2 feet 2 inches to minimize
the potential for recirculation pump trips following reactor scrams.
2. Allow restarting the recirculation pump following an RPT when
the temperature differential between the coolant at the reactor bottom
head and the reactor steam dome cannot be obtained, provided certain
conditions are met.
The licensee believes the above changes will aid in preventing
thermal stratification and unnecessary thermal cycles resulting from
the rapid cooldown of the bottom head region and the reduction in
reactor pressure to atmospheric conditions.
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:
Proposed Change 1
Proposed Change 1 does not involve a significant hazards
consideration, because it does not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Lowering the ATWS-RPT trip will not increase the probability of
occurrence of any design basis accident or transient, since this
change does not physically affect any component of the reactor
coolant pressure boundary (RCPB). Therefore, the probability of a
LOCA event is not increased. Lowering the ATWS-RPT water level
setpoint does not increase the probability of an ATWS event, since
no component of the CRD system or the reactor protection system is
being physically altered by this change. Also, the operation of
these two systems is not affected.
Reducing the setpoint may require installation of new slave trip
units; however, this addition does not increase the probability of
occurrence of accidents or transients. The new trip units will be
functionally identical to other slave trip units already in use at
Plant Hatch and are within the design capabilities of ATTS. In
conclusion, no safety-related plant system or component is being
affected in a manner that would render it more susceptible to
failure.
Lowering the setpoint does not result in an increase of the
consequences of a previously evaluated accident. GE reviewed the
proposed reduction and determined the results of the ATWS event with
the lowered setpoint remain acceptable. An approved analytical
method (REDY) was used to evaluate a bounding ATWS event -- LOFW
[loss of feedwater]. The results indicate that reactor power with
the new ATWS-RPT setpoint remains stable, with no unacceptable power
spikes. Hot and cold reactor shutdowns can still be ultimately
attained.
The consequences of non-ATWS events are not increased. For LOCA
events, reducing the recirculation pump low water level trip
setpoint allows the recirculation pumps to run longer. The forced
circulation provided by the recirculation pumps keeps the fuel
cooler for a longer period of time. The ECCS-LOCA analysis assumes
the pump trip and coastdown early in the event. Therefore, lowering
the ATWS-RPT makes the ECCS-LOCA analysis more conservative and, as
a result, it does not need to change.
Based on the above discussion, Proposed Change 1 does not
constitute an increase in the probability or consequences of a
previously analyzed accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
Lowering the ATWS-RPT trip will not alter the design or
operation of any safety-related system. The change may require
adding new slave trip units to ATTS; however, the new trip units
will be functionally identical to the equipment already in use at
Plant Hatch. Furthermore, the addition of this slave trip unit is
within the design capabilities of ATTS.
Since no new operation modes, accident scenarios, or failure
modes are introduced, Proposed Change 1 does not create the
possibility of a new type of accident.
3. Involve a significant reduction in the margin of safety.
As stated previously, reducing the ATWS-RPT low water level
setpoint will not cause unacceptable results for ATWS events.
Specifically, the LOFW event is bounding for all the ATWS events. An
evaluation using approved analytical methods indicates that reducing
the ATWS-RPT setpoint will not result in power instabilities or
unacceptable power spikes, or prevent the mitigation of an ATWS
event. (Reference Enclosure 1 [of the licensee's submittal],
Proposed Change 1).
The ATWS-RPT aids in maintaining the level above the top of the
active fuel. The reduction of core flow reduces the neutron flux and
thermal power and, therefore, the rate of coolant boil-off. However,
the setpoint reduction does not significantly reduce the margin of
safety since a substantial margin remains to the top of the active
fuel.
For non-ATWS events, delaying the RPT will provide a slight
improvement in the current ECCS-LOCA analysis, thereby improving the
margin of safety.
The margin of safety for transients is not reduced because plant
transient (MCPR) analyses do not take credit for the ATWS-RPT trip.
Proposed Change 2
Proposed Change 2 does not involve a significant hazards
consideration, because it does not:
1. Involve a significant increase in the probability of
occurrence or the consequences of a previously analyzed accident.
Allowing a recirculation pump restart within 30 minutes of a
trip, when the temperature differential is unknown, will not
increase the probability of occurrence of a previously analyzed
accident because this change does not physically alter the RCPB.
Additionally, the proposed change does not alter the design or
function of any safety-related systems.
Furthermore, no recirculation system equipment is being changed
as a result of this amendment. The start circuitry and trip
circuitry remain[s] unaffected. Operation of the recirculation
system with the reactor at power is also unaffected. As a result,
the probability of the chapter 14 and 15 events dealing with the
recirculation system are not increased; i.e., trip of one or both
recirculation pumps, recirculation pump seizure, recirculation flow
controller failure, etc.
The purpose of the 145 deg.F temperature differential
requirement is to avoid thermal shock caused by hot water on the
cold CRD stub tubes during recirculation pump restart. If the
temperature differential is unable to be determined, restart within
30 minutes of the trip will not increase the probability or severity
of thermal fatigue on the stub tubes. As discussed in Enclosure 1,
Basis for Proposed Change 2, stratification will not develop within
a 30-minute period following pump trip, thus, the temperature
differential will not exceed 145 deg.F. Additional caveats are
provided to insure the required temperature differential is met.
These involve certain conditions of ECCS injection, feedwater
temperature, and drive flow.
General Electric verified that this provision for recirculation
pump restart will not affect any plant safety analysis, including
radiological analysis. Therefore, the consequences of previously
analyzed events are not increased.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed recirculation pump restart provisions do not
introduce any new plant operating modes, accident scenarios, or
equipment failure modes. All other requirements for recirculation
pump restart; e.g., those addressing equipment protection and power
oscillations, will continue to apply.
3. Involve a significant reduction in the margin of safety.
The 145 deg.F differential temperature requirement is in place
to avoid thermal fatigue on the CRD stub tubes and the in-core
housing welds. Allowing the early restart with the listed caveats,
when temperature indication is not available, is acceptable because
the conditions for re-start insure that a stratified condition has
not yet developed. Thus, the cooler vessel structures at the vessel
bottom will not experience a severe thermal shock resulting from
exposure to hot water following the pump restart.
This change will actually aid in preventing the development of a
stratified condition, since the recirculation pumps will be
restarted before a stratified condition can develop, thereby helping
to maintain RCPB integrity. In the past, it has often been necessary
to depressurize the RPV [reactor pressure vessel] to atmospheric
pressure before the required temperature differential was met.
Proposed Change 2 should reduce the number of times depressurization
is required, thus avoiding unnecessary thermal cycles on the RPV.
Therefore, the margin of safety regarding the protection of RPV
components from severe thermal stresses, and the integrity of the
RCPB has not been reduced, and may actually increase.
The margin of safety in existing plant analyses is not reduced,
because none of the analyses are adversely affected as a result of
allowing the pump restart within 30 minutes of the RPT, as indicated
in GE's review of plant transient and accident analyses.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Appling County Public Library,
301 City Hall Drive, Baxley, Georgia 31513
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
NRC Project Director: Herbert N. Berkow
Georgia Power Company, Oglethorpe Power Corporation, Municipal
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant (VEGP), Units 1
and 2, Burke County, Georgia
Date of amendment request: October 3, 1994
Description of amendment request: This amendment would replace the
reactor coolant system heatup and cooldown limitations for VEGP Units 1
and 2, contained in Technical Specification figures 3.4-2a through 3.4-
3b, and the maximum allowable nominal power-operated relief valve
(PORV) setpoint for the cold overpressure protection system. These
changes are the results of new analyses that account for the
nonconservatisms identified in NRC Information Notice 93-58, the
results of reactor pressure vessel surveillance capsule examinations,
and recently issued ASME Code Case N-514.
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:
Conformance of the proposed amendment with the standards for a
determination of no significant hazards, as defined in the three
factor test of 10 CFR 50.92, is shown in that the proposed
amendment:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The revised heatup and cooldown limits and PORV setpoints ensure
that the Appendix G pressure/temperature limits are not exceeded and
therefore, help ensure that RCS integrity is maintained. The changes
do not result in a condition where the design, material, and
construction standards of the RCS are altered. In addition, the
safety function of the COMS (cold over-pressure mitigation system),
which is related to accident mitigation, has not been degraded.
Therefore, the probability of an accident is not increased by the
PORV setpoint change.
The changes do not adversely affect the integrity of the RCS
such that its function in the control of radiological consequences
is affected. In addition, the changes do not affect any fission
barrier. The changes do not degrade or prevent the response of the
COMS or other safety-related system to accident scenarios, as
described in FSAR chapter 15. In addition, the changes do not alter
any assumption previously made in the radiological consequence
evaluations nor affect the mitigation of the radiological
consequences of an accident described in the FSAR. Therefore, the
consequences of an accident previously evaluated in the FSAR will
not be increased.
Thus, operation of VEGP Units 1 and 2 in accordance with the
proposed license amendment, does not involve a significant increase
in the probability or consequences of any accident previously
evaluated.
2. Does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The changes do not cause the initiation of any accident nor
create any new credible limiting single failure for safety-related
systems and components. The changes do not result in any event
previously deemed incredible being made credible. As such, it does
not create the possibility of an accident different than any
evaluated in the FSAR.
The changes do not have any effect on the ability of the safety-
related systems to perform their intended safety functions. The
changes do not create failure modes that could adversely impact
safety-related equipment. Therefore, it will not create the
possibility of a malfunction of equipment important to safety
different than previously evaluated in the FSAR. Thus, the proposed
license amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does not involve a significant reduction in a margin of
safety.
The evaluation has shown that the PORV setpoints ensure that the
Appendix G pressure/temperature limits are not exceeded. The
analysis to support the proposed PORV setpoint change demonstrates
that the appropriate criteria, including that of ASME Code Case N-
514, are met for the postulated RCS pressures and temperatures. An
adequate margin of safety against vessel failure is assured, in
part, by the safety factors identified in Appendix G to Section III
of the ASME Boiler and Pressure Vessel Code, and [in] the basis for
ASME Code Case N-514 as well as [in the] added margin to prevent
lifting of the PORVs. The heatup and cooldown limits are designed to
prevent nonductile failure of the reactor vessel and take into
account the results of surveillance capsule Y on the reactor vessel
materials for VEGP Unit 1. The actuation of the safety-related
components and responses of the safety-related systems will remain
as modeled in the safety analyses. The changes will have no adverse
[effect] on the availability, operability, or performance of the
COMS. Therefore, the changes will not reduce the margin of safety,
as described in the bases to any Technical Specification.
Thus, [this] proposed license amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Burke County Public Library,
412 Fourth Street, Waynesboro, Georgia 30830.
Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308
NRC Project Director: Herbert N. Berkow
Gulf States Utilities Company, 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: January 14, 1994, as supplemented by
letter dated November 10, 1994.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) by removing component lists
from the TSs in accordance with NRC Generic Letter (GL) 91-08 and by
removing the schedule for withdrawal of reactor vessel material
specimen capsules from the TSs in accordance with GL 91-01. This
proposed amendment was originally noticed in the Federal Register on
May 23, 1994, (59 FR 26675). The licensee's letter dated November 10,
1994, provides clarification of the wording in the proposed TSs and
does not change the proposed determination that the amendment request
involves no significant hazards consideration. However, the notice is
being repeated here.
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:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change will not result in any hardware or operating
changes. The proposed change is based upon Generic Letters 91-01 and
91-08 and merely removes component lists, removes details relating
to the component lists, provides clarifying information supporting
the removal of the component listings, or removes details (which are
considered administrative) that are no longer applicable to the
Technical Specifications. The components listed in the affected
Technical Specifications are assumed in the mitigation of accident
and transient events. The removal of tabular component listings from
the Technical Specifications does not impact affected component
OPERABILITY requirements. Technical Specifications will continue to
require the components to be OPERABLE. Action statements and
surveillance requirements for the components will also remain in the
Technical Specifications. The tabular component lists are relocated
to the Technical Requirements Manual which will be in accordance
with the change control provisions specified in the Administrative
Controls Section of the Technical Specifications (Specification
6.5.2). Therefore, this change is administrative in nature and does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes to parameters governing normal plant
operation. The proposed change will not impose any different
requirements and adequate control of information will be maintained.
No new failure modes are introduced. Therefore, this proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumption. The proposed
changes do not alter the scope of equipment currently required to be
OPERABLE or subject to surveillance testing, nor do the proposed
changes affect any instrument setpoints or equipment safety
functions. Therefore 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: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005
NRC Project Director: William D. Beckner
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of amendment request: November 10, 1994
Description of amendment request: The proposed amendment revises
the Duane Arnold Energy Center (DAEC) Technical Specification (TS)
Section 3.2.A to refer to the Offsite Dose Assessment Manual (ODAM) for
the setpoint of the Offgas Stack Radiation Monitor and makes the
``Applicable Operating Mode'' and the ``Action'' statements for these
instruments consistent with the required function. The Action statement
for the other instruments which initiate Secondary Containment
isolation is also revised to be consistent with the current practice
and with the function of those instruments. The Basis is also revised
to add further description of the function and requirements.
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 provided below:
1) The proposed amendment will not involve a significant
increase in the probability or consequences of an accident
previously evaluated because the instruments will still be required
to be operable to initiate an isolation at a setpoint which will
assure that the offsite dose limits are preserved, as designed, or
else administrative controls will be established for the venting of
primary containment. Through either means, offsite releases will be
maintained within the limits established in the ODAM. The change to
the applicable operating mode simply will require that the
instruments be operable when they are assumed to be operable in
previously analyzed accidents. The change to the required action
when the TS requirement cannot be met will assure that the flow path
from containment is isolated or that positive control is established
so that any offsite radioactive gaseous release is within the limits
analyzed in the ODAM.
2) The proposed amendment will not create the possibility of a
new or different kind of accident from any previously evaluated
because the affected instruments are inputs to the secondary
containment isolation and the revised specification will assure that
they are operable or adequately compensated when they are assumed to
perform their function. The instruments initiate a secondary
containment isolation in the event that high radiation levels are
detected in the monitored effluent.
3) The proposed amendment will not involve a significant
reduction in a margin of safety because the revised applicability
statement will assure that the instruments are operable when they
are required to perform their function. The proposed compensatory
action allows administrative control of the isolation valves when
the instruments are inoperable and it is necessary to continue
venting. This allowance recognizes that venting is a controlled
evolution and that operator action would be adequate to prevent
excessive releases in the event of high radioactivity in the offgas
piping. The revision to the setpoint will not affect system
operation, but will continue to assure that the gaseous effluents
released are within the limits specified in the ODAM.
In summary, the proposed changes do not change the probability
or consequences of an accident previously evaluated, do not create
the possibility for a new or different kind of accident and do not
involve a 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 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: Cedar Rapids Public Library,
500 First Street, S.E., Cedar Rapids, Iowa 52401.
Attorney for licensee: Jack Newman, Kathleen H. Shea, Newman,
Bouknight & Edgar, PC, 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Leif J. Norrholm
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: October 7, 1994
Description of amendment requests: The proposed amendments would
remove the requirements for the Nuclear Safety and Design Review
Committee (NSDRC) to audit, and for the Plant Nuclear Safety Review
Committee (PNSRC) to review, the Emergency and Security plans and
implementing procedures. The composition of the PNSRC and the NSDRC
would also be revised to reflect organizational changes. Changes would
be made to the delegation of responsibility by the Site Vice President/
Plant Manager, and title corrections would be made on all pages
affected by the above changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:We [the licensee] have
evaluated the proposed T/S changes and have determined that the changes
should involve no significant hazards consideration. The proposed
amendment involves changes to the administrative controls section of
the T/Ss only. Because all changes reflect organizational/title changes
only or guidance from GL 93-07, they do not:
1) involve a significant increase in the probability or
consequence of an accident previously evaluated;
2) create the possibility of a new or different kind of accident
from any accident previously evaluated; or
3) involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske
MemorialLibrary, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: November 16, 1994
Description of amendment requests: The proposed amendments would
allow core offload 100 hours after core subcriticality instead of the
168 hours currently required. Also included in this submittal are minor
typographical corrections to Figure 5.6-1, ``Normal Storage Pattern
(Mixed Three Zone), and Figure 5.6-2, ``Interim Storage Pattern
(Checkerboard).
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:
We [the licensee] have evaluated the proposed T/S, editorial and
clarification changes and have determined that they do not represent
a significant hazards consideration based on the criteria
established in 10 CFR 50.92(c). Operation of Cook Nuclear Plant in
accordance with the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Although one of the proposed changes results in initiation of
core offload earlier after subcriticality than is currently allowed,
it does not increase the probability or consequences of an accident
previously evaluated. The bulk pool water temperatures, fuel rod
clad temperatures, and pool wall concrete temperatures will be
within acceptable limits as shown in Attachment 2 [of the November
16, 1994, submittal]. In addition, the subject change will not
result in an uncontrolled release of radiation to the environment
and will not initiate an accident. The remaining changes are
editorial in nature and have no [e]ffect on probability or
consequences of a postulated accident.
(2) Create the possibility of a new or different kind of
accident from an accident previously evaluated.
As previously stated, the earlier fuel movement change will not
result in bulk pool water, fuel rod clad, or concrete temperatures
which would initiate bulk pool boiling, challenge fuel rod integrity
or jeopardize the structural integrity of the pool. This change will
also have no impact on the criticality, structural, seismic, or
dropped assembly accident analysis previously performed and accepted
by the NRC. Consequently, the proposed T/S change does not create
the possibility of a new or different kind of accident from any
previously analyzed. The remaining changes have no [e]ffect on [the]
nature or probability of a postulated accident.
(3) Involve a significant reduction in a margin of safety.
The proposed change for earlier fuel movement will not result in
bulk pool water temperatures, fuel rod clad temperatures or concrete
temperatures which would initiate bulk pool boiling, challenge fuel
rod integrity or jeopardize the structural integrity of the pool.
This proposed change will not affect the results of any other
analysis associated with the spent fuel pool. It is, therefore,
concluded that this change poses no significant reduction in a
margin of safety. The remaining changes have no [e]ffect on the
nature or probability of a postulated accident.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of amendment requests: November 18, 1994
Description of amendment requests: The license amendment requests
propose a change to Technical Specification (T/S) 4.0.5 for both units
to delete the wording ``except where specific written relief has been
granted by the Commission pursuant to 10 CFR 50, Section
50.55a(g)(6)(i).'' This change, which is consistent with guidance in
the November 1993 draft NUREG-1482, ``Guidelines for Inservice Testing
at Nuclear Power Plants,'' would allow the licensee to implement
certain 10 CFR 50.55a relief requests while the relief requests were
being reviewed by the NRC.
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:
We [the licensee] have evaluated the proposed T/S change and
have determined that the change involves no significant hazards
consideration. Operation of Cook Nuclear Plant in accordance with
the proposed amendment will not:
(1) Involve a significant increase in the probability or
consequences of an accident.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The inspections required under Section XI are intended to
show the operational readiness of the applicable components, and
exceptions to the Code are allowed. When taking relief from Code
requirements, alternate requirements are developed which provide a
high level of confidence that components will perform their intended
function.
The proposed change does not alter the Code requirements or
lessen our obligations under existing regulations. Its only effect
is to allow implementation of Code relief prior to obtaining NRC
written approval. The proposed T/S change is consistent with NUREG-
1431, and, as such, has been found to be acceptable by the NRC.
Therefore, we believe that implementation of this change will not
involve a significant increase in the probability or consequences of
a previously analyzed incident.
(2) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed amendment does not create the possibility of a new
or different kind of accident from any previously evaluated. Typical
relief requests involve using alternative testing methods or
increasing the time interval between tests. Each proposed
alternative must assure that the component will perform its intended
function. The proposed change involves no physical changes to the
plant; therefore, we believe that implementation of this change will
not introduce a new of different kind of accident than previously
analyzed.
(3) Involve a significant reduction in a margin of safety.
The proposed amendment does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The inspections required under Section XI are intended to
show the operational readiness of the applicable components, and
exceptions to the Code are allowed. When taking relief from Code
requirements, alternate requirements are developed which provide a
high level of confidence that components will perform their intended
function.
The proposed change does not altar the Code requirements or
lessen our obligations under existing regulations. Its only effect
is to allow implementation of Code relief prior to obtaining NRC
written approval. The proposed T/S change is consistent with NUREG-
1431, and, as such, has been found to be acceptable by the NRC.
Therefore, we believe that implementation of this change will not
result in a significant reduction of the margin of safety.
NRC staff has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
NRC Project Director: John N. Hannon
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of amendment request: November 14, 1994
Description of amendment request: The proposed license amendment
would revise Technical Specification 4.5.1.e.2.e) to reduce the leak
rate test pressure for the Automatic Depressurization System (ADS)
nitrogen receiving tanks from 385 psig to 365 psig. This pressure
reduction would be made to reduce potential degradation of the rupture
disk installed on each ADS nitrogen receiving tank during periodic leak
testing of the receiving tanks. Plant operating experience has shown
that leak rate testing at 385 psig occasionally results in inadvertent
failure of the rupture disks. Testing at the reduced pressure would be
consistent with the manufacturer's recommendations.
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 operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The ADS is required to effect or support the safe shutdown of
the reactor. This is accomplished by the blowdown of steam from a
group of seven designated ADS SRVs [safety/relief valves] to the
suppression pool. The proposed change to the test pressure does not
affect any accident precursors. Therefore, the proposed change
cannot increase the probability of an accident previously evaluated.
In the event the nitrogen gas supply from the nitrogen gas
storage tanks is lost, a minimum nitrogen pressure of 334 psig in
the ADS nitrogen receiver tanks assures a five-day supply of
nitrogen to the ADS accumulators. The proposed change to
Surveillance Requirement 4.5.1.e.2.e) would decrease the leak rate
test pressure of the ADS nitrogen receiver tanks from 385 psig to
365 psig. Since the proposed test pressure remains well above the
design minimum pressure of 334 psig, the surveillance test continues
to ensure that the actual leakage of the safety related ADS
accumulator pneumatic supply system is bounded by the leakage
assumptions contained in the system design. In addition, the
surveillance test continues to ensure that the ADS nitrogen receiver
tanks are capable of providing a 5-day supply of nitrogen to the ADS
accumulators. Therefore, the proposed change does not significantly
increase the consequences of a previously evaluated accident.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed leak rate test pressure of 365 psig for the ADS
nitrogen receiver tanks is above the minimum design pressure to
assure a 5-day supply of nitrogen is available to the ADS
accumulators if makeup from the high pressure nitrogen gas storage
tanks is lost. With the proposed change, the ADS will continue to
perform its safety function of effecting and supporting the safe
shutdown of the reactor. The nitrogen receiving tank test pressure
is not a precursor for any new or different accident and the change
does not affect the operation of the system in any way.
Accordingly, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
The operation of Nine Mile Point Unit 2, in accordance with the
proposed amendment, will not involve a significant reduction in a
margin of safety.
The operation of the ADS SRVs, in conjunction with the LPCI [low
pressure coolant injection] mode of RHR [residual heat removal
system] and/or the LPCS [low pressure core spray] system, functions
as an alternative to the HPCS [high pressure core spray system] for
protection against fuel cladding damage upon a small break loss-of-
coolant accident. The blowdown of steam by these SRVs depressurizes
the reactor, allowing injection by the low-pressure coolant
injection sources. With the proposed change, the ADS will continue
to perform its intended safety function of effecting and supporting
the safe shutdown of the reactor as an alternate to the HPCS. The
proposed test pressure of 365 psig remains well above the minimum
acceptable pressure for the nitrogen receiver tanks of 334 psig.
Therefore, the change will not involve a significant reduction in
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Project Director: Michael J. Case
Northeast Nuclear Energy Company, et al., Docket No. 50-336,
Millstone Nuclear Power Station, Unit No. 2, New London County,
Connecticut
Date of amendment request: November 30, 1994
Description of amendment request: The proposed amendment would
modify the Technical Specifications by adding a footnote to Limiting
Conditions for Operation (LCOs) 3.8.1.1.b and 3.8.1.2.b which will
denote that 24,000 gallons of fuel oil is capable of supporting the
operation of one emergency diesel generator (EDG) for at least 4 days
and the other EDG for 1 hour with the EDGs loaded to the continuous
rated load of 2750 kW.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
...The proposed changes do not involve a significant hazards
consideration because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases
Section 3/4.8 will revise the Millstone Unit No. 2 design
requirements regarding the volume of EDG fuel oil which is required
to be stored onsite. The new rationale indicates that 24,000 gallons
of safety-related fuel oil would support the operation of one EDG
for at least four days with the other EDG running for at least one
hour. These run-times assume the EDGs are loaded to the continuous
rated loading of 2750 kW.
The proposed changes have no effect on EDG operation and
reliability. They provide additional operational flexibility,
because the EDG loading can be varied without the EDG minimum run-
time being altered. Also, an EDG run-time of at least four days
provides significant time to replenish fuel oil from onsite and
offsite sources even in the event of a hurricane or seismic event.
Based on the above, there is no effect on the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The EDGs are required to operate in response to a loss of
offsite power. The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b
and Bases Section 3/4.8 do not change the manner in which the EDGs
respond to a design basis accident. Also, the proposed changes do
not introduce any new failure mechanisms. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases
Section 3/4.8 have no effect on EDG operation and reliability. They
provide additional operational flexibility, because the EDG loading
can be varied without the EDG minimum run-time being altered.
An EDG run-time of at least four days provides significant time
to replenish EDG fuel oil via onsite or offsite sources even in the
event of a hurricane of seismic event. EPIP 4400 requires that the
need to order EDG fuel oil be evaluated within four hours of a loss
of offsite power event. Also, the high reliability of the electrical
grid and the high probability that offsite power would be restored
within 24 hours reduces the need to rely on extended EDG operation.
Millstone Unit No. 2 has more margin than is indicated by the
new design requirements. The EDG run-time will be significantly
greater than four days, because the electrical loading on the EDGs
will be less than the continuous rated loading, and electrical loads
will be shed through normal recovery actions following a design
basis accident.
Based on the above, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel,
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT
06141-0270.
NRC Project Director: Phillip F. McKee
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: July 11, 1994
Description of amendment requests: The proposed amendments would
change license condition 2.C.4 of each license to conform to the
standard fire protection license condition as stated in Generic Letter
(GL) 86-10, ``Implementation of Fire Protection Requirements.'' In
addition, the amendments would delete the fire protection program
elements from the Technical Specifications and incorporate, by
reference, the NRC-approved Fire Protection Program and major
commitments, including the fire hazards analysis, into the Updated
Safety Analysis Report. Guidance for these proposed changes is also
provided in GL 88-12, ``Removal of Fire Protection Requirements from
Technical Specifications.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) The proposed amendment[s] will not involve a significant
increase in the probability or consequences of [an] accident
previously evaluated.
The requested changes are administrative in nature in that they
move fire protection requirements from the Technical Specifications
to the Fire Protection Program and associated implementing
procedures following the guidance provided in GL 86-10 and GL 88-12.
The requested changes will not revise the requirements for fire
protection equipment operability, testing or inspections. The
amendment would give added responsibility to the Operations
Committee for review of the Fire Protection Program in accordance
with the guidance given in GL 86-10 and 88-12 including special
reporting requirements associated with limiting conditions for
operation for fire protection systems.
The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident, nor do they affect
any assumptions or conditions in any of the accident analyses. Since
the accident analyses remain bounding, their radiological
consequences are not adversely affected.
Therefore, the probability or consequences of an accident
previously evaluated are not affected.
(2) The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The requested changes are administrative in nature in that they
move fire protection requirements from the Technical Specifications
to the Fire Protection Program and associated implementing
procedures following the guidance provided in GL 86-10 and GL 88-12.
The requested changes will not revise the requirements for fire
protection equipment operability, testing or inspections. The
amendment would give added responsibility to the Operations
Committee for review of the Fire Protection Program in accordance
with the guidance given in GL 86-10 and 88-12 including special
reporting requirements associated with limiting conditions for
operation for fire protection systems.
The proposed changes do not involve any change to the
configuration or method of operation of any plant equipment that is
used to mitigate the consequences of an accident.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be
created.
(3) The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
The requested changes are administrative in nature in that they
move fire protection requirements from the Technical Specifications
to the Fire Protection Program and associated implementing
procedures following the guidance provided in GL 86-10 and GL 88-12.
The requested changes will not revise the requirements for fire
protection equipment operability, testing or inspections. The
amendment would give added responsibility to the Operations
Committee for review of the Fire Protection Program in accordance
with the guidance given in GL 86-10 and 88-12 including special
reporting requirements associated with limiting conditions for
operation for fire protection systems.
Therefore, a significant reduction in the margin of safety would
not be involved.
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
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station, Unit No. 1, Washington County, Nebraska
Date of amendment request: November 11, 1994
Description of amendment request: The proposed amendment to the
Technical Specifications (TSs) would make administrative changes to TS
5.2 and 5.5. These changes reflect organizational changes in OPPD
senior management, delete specific titles of personnel on the Safety
Audit and Review Committee (SARC) and Plant Review Committee (PRC), and
make changes to SARC reviews and audits.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes are administrative changes to reflect
organizational changes in Omaha Public Power District (OPPD) Senior
Management, remove specific titles from the membership of the Plant
Review Committee (PRC) and the Safety Audit and Review Committee
(SARC), add minor clarifications to SARC reviews and audits and
delete statements concerning the frequency of SARC audits from the
Technical Specifications (TS).
The proposed change to revise the overall corporate
responsibility for plant nuclear safety from the Senior Vice
President to Vice President is administrative in nature as it only
reflects an organizational change. Section 12 of the Updated Safety
Analysis Report describes the management structure and reporting
responsibilities of OPPD. Section 12 provides an organizational
chart to differentiate the Vice President in charge of nuclear
activities from other Vice Presidents within OPPD. Therefore,
changing the corporate reporting responsibility does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes to the membership of the PRC and SARC are
administrative in nature since only the specific titles of the
members are being removed from the TS. The management level and
expertise of personnel who are PRC or SARC members is not being
changed. The review of plant operations is still required to be in
compliance with ANSI N18.7-1976 and Regulatory Guide 1.33, Revision
2, as committed to in the Fort Calhoun Station Quality Assurance
(QA) Program. Any changes in the QA Program which reduce the
effectiveness of the program must be approved by the NRC in
accordance with 10 CFR 50.54(a)(3). Therefore, the proposed changes
to the membership of the PRC and SARC do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Clarifications of SARC reviews and audits and the deletion of
SARC audit frequencies from the TS are administrative changes. The
audit frequencies are required by the NRC approved QA Program and
any changes that could reduce the effectiveness of the QA Program
must be approved by the NRC in accordance with 10 CFR 50.54(a)(3).
Therefore, the clarifications and deletion of the specific audit
frequencies do not involve a significant increase in the probability
or consequences of an accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed changes are administrative in nature to reflect
organizational changes in OPPD Senior Management, remove specific
titles from the membership of the PRC and SARC, provide minor
clarifications of SARC reviews and audits and delete statements
concerning the frequency of SARC audits from the TS. The proposed
changes do not revise any equipment setpoints, change the manner in
which any plant equipment is operated, or propose any new operating
modes. Therefore, the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(3) Involve a significant reduction in a margin of safety.
The proposed changes revise organizational and administrative
requirements contained within the Administrative Controls section of
the TS. The proposed changes do not revise any equipment setpoints,
change the manner in which any plant equipment is operated, or
propose any new operating modes. Therefore, the proposed changes do
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: W. Dale Clark Library, 215
South 15th Street, Omaha, Nebraska 68102
Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875
Connecticut Avenue, NW., Washington, DC 20009-5728
NRC Project Director: Theodore R. Quay
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: October 28, 1994
Description of amendment request: The proposed changes to the
Technical Specifications (TS) for the two units would add reference
20 (Unit 1) and reference 18 (Unit 2) to Section
6.9.3.2 as ``PL-NF-90-001, Supplement 1, 'Application of Reactor
Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating
Changes and Use of RETRAN MOD 5.1', September 1994''. These changes
would add changes to the methodology that the licensee is using to
perform its nuclear fuel reload analysis for the two units.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Incorporation of these proposed minor changes into PP&L's NRC
approved methodology for performing reload licensing analysis is
considered to be an enhancement to the currently approved
methodology. Upgrading of the RETRAN code allows for taking
advantage of state-of-the-art technology, while utilization of the
generic correlation for the LOFWH event supports consistency in
licensing analysis performance. Results of incorporating these
changes will not significantly increase the probability or the
consequences of an accident previously evaluated.
II. Create the possibility of a new or different kind of
accident from any accident previously evaluated.
As stated above, the incorporation of these minor changes are
considered enhancements, allowing PP&L to more efficiently and cost
effectively continue to perform future reload licensing analysis.
Therefore, the incorporation of these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
III. Involve a significant reduction in a margin of safety.
In addition to the extensive testing perform by EPRI, PP&L has
performed its own comparison tests utilizing RETRAN MOD005.1 in
place of MOD004 for four licensing transients that use the RETRAN
code. Results of this comparison were essentially the same for both
codes and support this proposed change. Also, the Loss of Feedwater
Heating event is not a limiting event for establishing MCPR
Operating Limits for Susquehanna. Therefore, the incorporation of
these changes will have no impact on current safety margins, nor
will they 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: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: October 28, 1994
Description of amendment request: The proposed changes to the
Technical Specifications (TS) for the two units would make a number of
administrative changes. These would include changing the title of the
positions of Superintendent of Plant to the Vice President-Nuclear
Operations, and changing the title of Vice-President-Nuclear Operations
to Senior Vice President-Nuclear for the listing of the assignment of
certain duties in various sub-sections of Section 6.0 of the TS. Other
proposed changes would be the deletion of a number of footnotes
indicating times, dates, and events that are no longer applicable, the
addition of a footnote to Section 6.5.1.2 indicating that the Station
Duty Manager shall act as a PORC [Plant Operations Review Committee]
chairman in the absence of the Vice President-Nuclear Operations, and
the change of the Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report in Table 4.11.2.1.2-1
footnote g.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposal to change the title of Superintendent of Plant to
Vice President-Nuclear Operations and Vice President-Nuclear
Operations to Senior Vice President-Nuclear (for certain duties) is
administrative in nature and does not compromise the minimum
qualifications or training required for these positions. Therefore,
the proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change for the removal of footnotes that reference
periods of time, dates, and events that have since past is justified
based on the fact that they are no longer applicable. Because
operators must perform unnecessary applicability reviews on these no
longer applicable footnotes, removing the footnotes decreases the
potential for confusion and incorrect actions. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed change to add a footnote indicating the Station
Duty Manager shall act as PORC chairman in the absence of the Vice
President-Nuclear Operations will ensure continuous leadership of
PORC and will enhance performance by ensuring a responsible
individual is available during all shifts. Therefore, the proposed
change does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
Changing the Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report was previously approved
in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating
this change into footnote g of Table 4.11.2.1.2-1 will maintain
accuracy and consistency. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposal to change the title of Superintendent of Plant to
Vice President-Nuclear Operations and Vice President-Nuclear
Operations to Senior Vice President-Nuclear (for certain duties) is
administrative in nature and does not compromise the minimum
qualifications or training required for these positions. Also, the
change does not diminish the responsibilities or functions of these
positions. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed change for the removal of footnotes that reference
periods of time, dates, and events that have since past is justified
based on the fact that they are no longer applicable. Because
operators must perform unnecessary applicability reviews on them,
removing the no longer applicable footnotes decreases the potential
for confusion and incorrect actions, thereby enhancing the safe
operation of Susquehanna SES. Therefore, the proposed change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change to add a footnote indicating the Station
Duty Manager shall act as PORC chairman in the absence of the Vice
President-Nuclear Operations will ensure continuous leadership of
PORC and will enhance performance by ensuring a responsible
individual is available during all shifts. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Changing Semiannual Radioactive Effluent Release Report to
Annual Radioactive Effluent Release Report was previously approved
in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating
this change into footnote g of Table 4.11.2.1.2-1 will maintain
accuracy and consistency. Therefore, the proposed change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
For the reasons discussed in items I and II above, as well as
the enclosed Safety Assessment, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: October 28, 1994
Description of amendment request: The amendment would delete the
requirements for chlorine detection and the associated Bases from the
Technical Specifications for each unit as a result of the removal of
bulk quantities of gaseous chlorine from the Susquehanna Steam Electric
Station. Specifically, Sections 3.3.7.8 and the associated Surveillance
Requirements in Section 4.3.7.8 would be deleted. In addition, Bases 3/
4.3.7.8 would also be deleted.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Review of the various design basis accidents identified in
Chapter 15 of the Susquehanna SES Final Safety Analysis Report
(FSAR) concluded that none of these accidents are affected by
deletion of the chlorine detection requirements from Technical
Specifications. With the elimination of bulk quantities of gaseous
chlorine from use at Susquehanna SES the probability of control room
inhabitability due to a gaseous chlorine release has actually
decreased. Therefore, this proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed change involves only the deletion of the chlorine
detection system Technical Specifications based upon a plant
modification to remove gaseous chlorine as a biocide from
Susquehanna SES and replace it with a nonoxidizing biocide. The
release of chlorine from an off-site source is bounded by Reg. Guide
1.95 in that manual isolation capability for the control room
ventilation system is acceptable. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed change would not alter the margins of safety
provided in the existing FSAR analysis (Sections 2.2.3.1.3 and 6.4)
for chlorine release events since the basis for the existing margin
of safety, which are the Reg. Guide 1.95 requirements, are not
altered by the change. As stated above, since gaseous chlorine is no
longer used for open cooling water treatment at Susquehanna SES and
since the nonoxidizing biocide is relatively nontoxic to humans,
safety margin has actually increased. Therefore, the proposed change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: November 11, 1994
Description of amendment request: The amendment would extend the
Main Turbine Valve surveillance test interval from a weekly basis to no
greater than 92 days for all Main Turbine Stop, Control, and Combined
Intermediate 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:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed Technical Specification change to a quarterly
turbine inlet valve surveillance test interval is based on
maintaining the turbine missile generation probability within the
NRC criteria as stated in Table 3.1 of NUREG-1048. This, combined
with the NRC acceptable strike-and-damage probability as specified
in NUREG-1048, will keep the probability of unacceptable damage to
safety-related structures, systems, and components from turbine
missiles acceptably low (i.e., <>-7) . Thus, the NRC
acceptable risk rate of <>-7/yr. is not changed and there is
no increase in the probability of an accident previously evaluated.
The proposed Technical Specification change to the turbine inlet
valve surveillance interval does not effect the sequence of events
or the consequences of an accident previously evaluated. The
surveillance interval does not affect the strike and damage scenario
of an accident previously evaluated. Thus, the radiological
consequences of an accident previously evaluated will not be
increased.
II. This proposal does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed Technical Specification change to the turbine inlet
valve surveillance interval does not affect the surveillance test
characteristics. There are no new surveillance testing requirements.
Surveillance testing of these valves does not create the possibility
for a new or different kind of accident from any accident previously
evaluated.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed Technical Specification change to the turbine inlet
valve surveillance interval is based on maintaining the same margin
of safety as previously determined by the NRC and does not reduce
the margin of safety. In fact, the reduction in the testing rate
will reduce the potential for testing related transients, which have
been credited with causing 18 reactor scrams in the period 1985
through 1992.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037
NRC Project Director: John F. Stolz
Philadelphia Electric Company, Public Service Electric and Gas
Company, Delmarva Power and Light Company, and Atlantic City
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: November 14, 1994
Description of amendment request: The proposed changes relocate
audit topics and frequencies, Nuclear Review Board review requirements
and requirements associated with the independent Safety Engineering
Group function from the Peach Bottom Atomic Power Station, Units 2 and
3 Technical Specifications to licensee controlled documents.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the changes relocate requirements from the TS to licensee
controlled documents consistent with the NRC Final Policy Statement
on TS Improvements. Any changes to the licensee controlled documents
will be evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59
as appropriate. Therefore, these changes will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2) The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because the changes will not alter the plant or the manner in which
the plant is operated. The changes will not involve a design change
or introduce any new failure modes. The changes will not alter
assumptions made in the safety analysis and licensing basis.
Adequate control of information will be maintained. Therefore, these
changes will not create the possibility of a new or different kind
of accident from any accident previously evaluated.
2) The proposed changes do not involve a significant reduction
in a margin of safety because they have no impact on any safety
analysis assumptions. The requirements to be transposed from the TS
to licensee controlled documents are the same as the existing TS.
Any future changes to licensee controlled documents will be
evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59 as
appropriate. Because the proposed changes are consistent with NUREG-
1433, as modified by approved generic change BWOG-09, and the change
controls for proposed relocated details and requirements provide an
equivalent level of regulatory authority, revising the TS to reflect
the approved level of detail and requirements ensures no 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 Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and
General Counsel, Philadelphia Electric Company, 2301 Market Street,
Philadelphia, Pennsylvania 19101
NRC Project Director: John F. Stolz
Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296,
BrownsFerry Nuclear Plant, Units 1, 2 and 3, Limestone County,
Alabama
Date of amendment request: November 15, 1994 (TS 350)
Description of amendment request: The proposed change would remove
the frequency for each of the audits specified in the administrative
controls section of the technical specifications (TS). The requirements
to perform the audits would be retained, but the frequency for their
performance would be controlled by a requirement to be added to the
Nuclear Quality Assurance Plan. This would require that the audits
listed in the TS be performed on a biennial frequency. In addition, the
proposed change would remove the requirement to perform site
Radiological Emergency Plan and Physical Security/Safeguard Contingency
Plan reviews and audits from the TS, since these requirements presently
exist in the respective Plans.
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:
TVA has concluded that operation of BFN units 1, 2, and 3 in
accordance with the proposed change to the technical specifications
does not involve a significant hazards consideration. TVA's
conclusion is based on its evaluation in accordance with 10 CFR
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).
TVA's conclusion is based on the following:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by this Technical Specification change which only
affects review and audit frequencies. This Technical Specification
change will not impact the function or method of operation of plant
equipment. Thus, there is not a significant increase in the
probability of a previously analyzed accident due to this change. No
systems, equipment, or components are affected by the proposed
change. Thus, the consequences of a malfunction of equipment
important to safety previously evaluated in the UFSAR are not
increased by this change.
The proposed change only affects review and audit frequencies.
As such, the proposed change has no impact on accident initiators or
plant equipment, and thus, does not affect the probabilities or
consequences of an accident.
Therefore, we conclude that this change does not significantly
increase the probabilities or consequences of an accident.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since program audits do not contribute to
accident initiation, a change related to audit functions cannot
produce a new accident scenario or produce a new type of equipment
malfunction. Also, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation, and, thus, does not create the possibility of a new or
different kind of accident. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change concerning conduct of reviews and audits
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are no affected by this proposed
change.
Therefore, use of the proposed Technical Specification would not
involve any reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: November 15, 1994 (TS 94-12)
Description of amendment request: The proposed change would remove
the frequency for each of the audits specified in the administrative
controls section of the technical specifications (TS). The requirements
to perform the audits would be retained, but the frequency for their
performance would be controlled by a requirement to be added to the
Nuclear Quality Assurance Plan. This would require that the audits
listed in the TS be performed on a biennial frequency. In addition, the
proposed change would remove the requirement to perform site
Radiological Emergency Plan and Physical Security/Safeguard Contingency
Plan reviews and audits from the TS, since these requirements presently
exist in the respective Plans.
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 standards used to arrive at a determination that a Technical
Specification change request involves no significant hazards
consideration are included in the Commission's regulations, 10 CFR
50.92, which states that no significant hazards considerations are
involved if the 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. Each standard is addressed as
follows:
1. Operation of the facility in accordance with the proposed
technical specifications would not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The likelihood that an accident will occur is neither increased
or decreased by the Technical Specification change which only
affects review and audit frequencies. This Technical Specification
change will not impact the function or method of operation of plant
equipment. Thus, there is not a significant increase in the
probability of a previously analyzed accident due to this change. No
systems, equipment, or components are affected by the proposed
changes. Thus, the consequences of a malfunction of equipment
important to safety previously evaluated in the FSAR are not
increased by this change.
The proposed change only affects review and audit frequencies.
As such, the proposed change has no impact on accident initiators or
plant equipment, and thus, does not affect the probabilities or
consequences of an accident.
Therefore, we conclude that this change does not significantly
increase the probabilities or consequences of an accident.
2. Operation of the facility in accordance with the proposed
technical specifications would not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not involve changes to the physical
plant or operations. Since program audits do not contribute to
accident initiation, a change related to audit functions cannot
produce a new accident scenario or produce a new type of equipment
malfunction. Also, this change does not alter any existing accident
scenarios. The proposed change does not affect equipment or its
operation, and, thus, does not create the possibility of a new or
different kind of accident. Therefore, the proposed change does not
create the possibility of a new or different kind of accident.
3. Operation of the facility in accordance with the proposed
technical specifications would not involve a significant reduction
in a margin of safety.
The proposed change concerning conduct of reviews and audits
does not directly affect plant equipment or operation. Safety limits
and limiting safety system settings are no affected by this proposed
change.
Therefore, use of the proposed Technical Specification would not
involve any reduction in the margin of safety.
The NRC has reviewed the licensee's analysis and, based on
thisreview, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Chattanooga-Hamilton County
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
NRC Project Director: Frederick J. Hebdon
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: September 8, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 4.2.2.2, 4.2.2.4, and 6.9.19 to
incorporate a penalty in the Core Operating Limit Report (COLR) to
account for FQ increases greater than 2 percent between
measurements.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes to the Technical Specifications do not
involve a significant hazards consideration because operation of
Callaway Plant in accordance with these changes would not:
1) Involve a significant increase in the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report. There is no increase in the probability of
occurrence or the consequences of an accident. The removal of
FQ(Z) penalty values from the Callaway Plant Technical
Specifications and the creation of cycle-specific FQ(Z) values
in the COLR has no influence or impact on the probability or
consequences of any accident previously evaluated. The cycle-
specific FQ(Z) values, although not in Technical
Specifications, will be followed in the operation of the Callaway
Plant. The proposed amendment still requires exactly the same
actions to be taken when or if FQ(Z) limits are exceeded as is
required by current Technical Specifications.
2) Create a possibility of a new or different kind of accident
from any previously evaluated in the safety analysis report. There
is no new type of accident or malfunction created and the method and
manner of plant operation will not change. As stated earlier, the
removal of the cycle-specific FQ(Z) value has no influence or
impact, nor does it contribute in any way to the probability or
consequences of an accident. No safety-related equipment, safety
function, or plant operation will be altered as a result of this
proposed change. The cycle-specific FQ(Z) values are calculated
using NRC approved methods. The Technical Specifications will
continue to require operation within the required FQ(Z) limits
and appropriate actions will be taken when or if limits are
exceeded.
3) Involve a significant reduction in a margin of safety. This
is based on the fact that no plant design changes are involved and
the method and manner of plant operation remains the same. The
margin of safety is not affected by change and removal of FQ(Z)
penalty values from the Technical Specifications. The margin of
safety presently provided by current Technical Specifications
remains unchanged. The current FQ(Z) limits remain unchanged
and the current safety analysis limits remain valid and unaffected
by this change. The proposed amendment continues to require
operation within the core limits as obtained from the NRC approved
design methodology and appropriate actions to be taken when or if
FQ(Z) limits are violated remain unchanged.
Given the above discussions as well as those presented in the
Safety Evaluation, the proposed change does not adversely affect or
endanger the health or safety of the general public or involve a
significant safety hazard.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: Leif J. Norrholm
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: September 12, 1994
Description of amendment request: The proposed amendment would
revise Technical Specification 3.7.1.1 (Tables 3.7-1 and 3.7-2). Tables
2.2-1 and 3.3-2, and Bases 3/4.7. Tables 3.7-1, 3.7-2, 3.3-2, and 2.2-1
would be revised to provide appropriate margin to relax main steam line
safety valve setpoint tolerance. Bases 3/4.7 would be revised to
incorporate the methodology used to determine the maximum allowable
power level associated with inoperable main steam line safety 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:
The proposed changes to the Technical Specifications do not
involve a significant hazards consideration because operation of
Callaway Plant in accordance with these changes would not:
1. Involve a significant increase in the probability of
occurrence or the consequences of an accident or malfunction of
equipment important to safety previously evaluated in the safety
analysis report. The main steam line safety valves are designed to
mitigate transients by preventing overpressurization of the main
steam system. The proposed change does not alter this design basis.
The revised analysis shows that the probability or
consequences of all previously analyzed accidents are not
changed by increasing the setpoint tolerance of the safety valves.
Therefore, there is no increase in the probability of occurrence or
the consequences of any accident.
2. Create the possibility of a new or different kind of accident
from any previously evaluated in the safety analysis report. There
is no new type of accident or malfunction created, the method and
manner of plant operation will not change nor is there a change in
the method in which any safety related system performs its function.
Any main steam safety valve lifting at the extremes of the proposed
tolerance will not result in low lift setpoint that is less than the
normal no load system pressure or a high lift setpoint that allows
main steam system overpressurization.
3. Involve a significant reduction in a margin of safety. This
is based on the fact that no plant design changes are involved and
the method and manner of plant operation remains the same. With the
increased setpoint tolerance, the main steam safety valves will
still prevent pressure from exceeding 110 percent of design pressure
in accordance with the ASME code. All FSAR accident analysis
conclusions remain valid and unaffected by this change.
Given the above discussions as well as those presented in the
Safety Evaluation, the proposed change does not adversely affect or
endanger the health or safety of the general public or involve a
significant safety hazard.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
NRC Project Director: Leif J. Norrholm
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: July 9, 1993, with supplemental
information provided October 8, 1993, October 25, 1993, January 6,
1994, February 2, 1994, May 3, 1994, May 13, 1994, September 26, 1994,
and October 12, 1994.
Description of amendment request: The proposed amendment would
modify the operating license and several Technical Specifications (TS)
to allow an increase in licensed power level by 4.9%. This would raise
the licensed power level from the current 3323 MWt to 3486 MWt, but
will not affect the basic fuel design or fuel operating limits. The
uprate in power will be accomplished by expanding the existing power-
to-flow map to allow an increase in core flow along the flow control
lines, with an associated increase in core power. The increased flow
and power will also cause an increase in operating reactor vessel steam
dome pressure that will require an associated TS change.
Other TS changes proposed to reflect necessary modifications to
address the power uprate are (1) an increase in the average power range
monitor (APRM) and flow biased scram and rod block monitor (RBM)
setpoints, (2) an increase in reactor steam pressure limits, (3) an
increase in main steam line (MSL) isolation valve and tunnel high
differential temperature to reflect the increased operating pressure,
(4) revised temperature/pressure limit curves to reflect the higher
neutron flux over vessel life, (5) an increase in calculated peak
containment pressure, (6) an increase in the pressure at which reactor
core isolation cooling (RCIC) testing occurs, and (7) an increase in
the safety relief valve (SRV) setpoints.
The licensee, in the above reference letters, also proposed changes
to the TS that are not associated with the power uprate. These proposed
changes are (1) a change of the reactor protection system and End-of-
Cycle/Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a
variable setpoint based on power level, (2) an increase in the SRV
setpoint tolerance, and (3) an increase in the number of automatic
depressurization system (ADS) valves that are allowed out of service.
Although not part of the change to the license or associated TS,
the licensee is also revising the bases to reflect the TS changes and
power uprate.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92 (c). The NRC staff's review is
presented below:
1. Does the amendment involve a significant increase in the
probability or consequences of an accident previously evaluated?
* [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]Plant
operation at a higher power level is accomplished by increasing reactor
core flow along flow control lines to achieve the desired increase in
steam flow to the turbine/generator. The maximum allowable reactor
recirculation flow rate remains unchanged from the original plant
design analysis. The increased reactor core flow requires a
corresponding increase in feedwater flow that remains well within the
design of the feedwater system components. The increased power also
requires a small increase in reactor pressure. Safety relief valve and
power, pressure, and flow-related instrumentation trip setpoints are
increased slightly to accommodate the power uprate, to maintain
approximately the same level of trip avoidance and safety system
challenges as before the uprated power condition.
The plant is operated in the same manner at uprated power as it is
at the currently licensed power level, including the methods and
sequences of system and component operation. Since the level of trip
avoidance and safety system challenges remains approximately the same,
the frequency of operational responses to these events is not
increased. Reactor fuel operating limits, designed to protect the fuel
cladding, are maintained and thus provide the same level of protection
as before the uprated power condition. The original design and
regulatory criteria established for plant equipment, including ASME
code, IEEE standards, NEMA standards, and Regulatory Guide criteria,
are still imposed and met for operation at the uprated power level. In
addition, the reactor vessel and internals, reactor connecting piping,
balance of plant piping, primary containment, and related systems and
components still meet the pre-uprate design and licensing criteria. The
power uprate does not change the likelihood of failure of these systems
or components. Thus, the probability of an accident previously
evaluated is not significantly affected by the proposed power uprate.
The consequences of postulated, power-dependent accidents are
proportional to the power level assumed in the safety analysis. This is
because potential offsite doses increase proportionately to reactor
power since the radiological source term is directly proportional to
reactor power. The meteorological factors are unaffected by the
proposed power uprate. The current accident analyses are based on
104.4% of original rated power. The accident analyses for power uprate
are based on 1.02 x 104.9% (or 107%) of original rated power. Thus,
power uprate increases postulated consequences of an accident by less
than 3% over previous postulated consequences, while still remaining
well within the 10 CFR Part 100 limits. In addition, a spectrum of
hypothetical accidents and transients were investigated for power
uprate, and the bounding events met the same regulatory criteria to
which WNP-2 is currently licensed, including Maximum Average Planar
Linear Heat Generation Rate (MAPLHGR), Operating Limit Minimum Critical
Power Ratio (OLMCPR), 10 CFR 50.46 and 10 CFR Part 50-Appendix K, for
fuel cladding integrity, and containment criteria in 10 CFR Part 50-
Appendix A, Criterion 38 and Criterion 50.
The results of these analyses demonstrate that operation at the
proposed power uprate level does not significantly increase the
probability or consequences of any accident previously evaluated.
* [Increase in the maximum allowable reactor steam dome operating
pressure limit]
The operating pressure limit is increased by the same amount as the
nominal operating pressure increase for power uprate. This change to
the dome operating pressure limit is consistent with and meets the
current design criteria used for evaluation of steady state operating
conditions and for the most limiting event for vessel overpressure
protection. Operation at a higher pressure results in the plant
operating closer to the criteria used in the design analysis. However,
since operation of the plant within design limits is considered to
result in a very low probability of failure of systems or components,
this small increase in maximum operating pressure is considered to have
a negligible increase in failure probability. Thus the proposed change
does not significantly increase the likelihood of failure of existing
systems or components, and does not significantly affect the
probability of an accident previously evaluated.
The power uprate overpressure protection analysis results show the
peak reactor pressure vessel (RPV) pressure will remain below the ASME
code limit, keeping any postulated radiological consequences within the
bounds of existing analyses. Thus the consequences of an accident
previously evaluated are not significantly affected.
* [Increase in the average power range monitor (APRM) and flow
biased scram and rod block monitor (RBM) setpoints]
These scrams and rod blocks are designed to prevent fuel damage due
to power during postulated events or anticipated operational
occurrences. The setpoints for these scrams and rod blocks are
increased by the same amount of the proposed increase in licensed
power. The change in these setpoints does not affect the operation of
any system or component, nor does it affect the circuitry that provides
the protective function. Thus the increase in setpoints does not affect
the probability of any accident previously evaluated.
The increased setpoints maintain the same difference between
licensed power level and scram setpoints, and between the scram line
and rod block lines on the extended load line limit curve. The
increased setpoints result in higher postulated source terms for
accidents previously analyzed. The proposed change does not affect the
response or operation of mitigative equipment. The licensee's analyses
supporting power uprate demonstrate that the increased values do not
significantly affect the consequences of an accident. The proposed
change also maintains the current level of scram avoidance with
associated avoidance of unnecessary challenges to plant equipment,
while providing protection for the fuel, reactor systems, and
containment to meet current design requirements. Thus the proposed
change does not significantly increase the consequences of accident
previously analyzed.
* [Increase in main steam line (MSL) high flow differential
pressure setpoint to reflect the increased operating pressure]
The MSL high flow differential pressure setpoint is increased to
reflect the higher steam flows necessary to operate the plant at the
higher power. The proposed change does not affect the design,
construction, or operation of existing plant equipment, nor does it add
new equipment. The proposed change does not, therefore, affect the
probability of accidents previously analyzed. The increased setpoint
does not affect the maximum closure time for the main steam isolation
valves, thus the release of fission products during postulated accident
is not changed from current design analyses. The proposed change does
not, therefore, affect the consequences of accidents previously
analyzed.
* [Revised temperature/pressure limit curves to reflect the higher
neutron flux over vessel life]
The temperature/pressure limit curves are being modified to reflect
the increased exposure to neutron flux over the life of the vessel, to
retain the existing margin to brittle fracture over vessel life. The
plant will continue to be operated in conformance to the pressure/
temperature limits, retaining the existing probability of overcooling
events and associated postulated brittle fracture of the reactor
vessel. If a failure were to occur, the mode of failure is unaffected
by the proposed change, thus the consequences of a postulated failure
of the reactor vessel is unchanged by the proposed amendment.
* [Increase in the pressure at which reactor core isolation cooling
(RCIC) testing occurs]
The pressure at which the RCIC system is tested is being increased
to ensure adequate system operation with the increased reactor system
pressure required for power uprate. No credit for RCIC system operation
is taken in any accident analysis nor is RCIC included as an accident
initiator, thus this change does not affect the probability or
consequences of an accident previously evaluated.
* [Change of the reactor protection system and End-of-Cycle/
Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a
variable setpoint based on power level]
This setpoint is being changed from a specific value of main
turbine first stage pressure to a value to be calculated based on
variable plant conditions. The basis for the setpoint will remain
unchanged at a value equivalent to thermal power less than 30% of rated
thermal power. This change is proposed because turbine first stage
pressure can vary for a given thermal power level depending on the
amount of subcooling in the core, which itself is highly dependent on
feedwater temperatures. This proposed change does not affect the
reliability of operation of the trip circuitry, nor does it affect the
design or operation of plant equipment or safety systems. Thus the
change does not affect the probability of accidents previously
evaluated. In addition, the proposed change retains the same basis for
establishing the setpoint as currently stated in the TS, thus the
setpoint will be established at the level (30% rated thermal power)
assumed in the accident analysis. Thus the proposed change does not
affect the consequences of any accidents previously evaluated.
* [Increase in the safety relief valve (SRV) setpoints]
The licensee proposes to increase the setpoints of the two lowest-
set SRVs to accommodate the change in maximum operating pressure after
power uprate. This increase in SRV setpoints maintains approximately
the same difference between maximum operating RPV pressure and the
lowest SRV setpoint as currently exists. As discussed in the above
section regarding increased operating pressure, increasing the pressure
at which the lowest SRV actuates would allow operation at a slightly
higher pressure than currently allowed. This results in the plant
operating closer to the criteria used in the design analysis. However,
since operation of the plant within design limits is considered to
result in a very low probability of failure of systems or components,
this small increase in maximum operating pressure is considered to have
a negligible increase in failure probability. Thus the proposed change
does significantly increase the likelihood of failure of existing
systems or components, and does not affect the probability of an
accident previously evaluated.
With the increased SRV setpoint, the analysis results show that the
peak RPV pressure will remain below the ASME code limit, keeping any
postulated radiological consequences within the bounds of existing
analyses. Thus the consequences of an accident previously evaluated are
not affected.*
[Increase the value of Pa (the pressure at which primary
containment is tested). Add a new definition for Pa in the
``Definition'' section of the TS that gives a specific value of
Pa, and simplify the TS by deleting the specific value of Pa
and 1.10Pa from other locations in the TS]
The analysis to support the power uprate resulted in a higher
calculated peak containment pressure in response to postulated
accidents. To maintain the validity of the radiological analysis, the
containment leak rate testing must be based on a pressure greater than
or equal to the peak calculated containment pressure. The proposed
change does not affect the design, construction, or operation of
existing plant systems or components, and therefore does not affect the
probability of accidents previously evaluated.
The increase in test pressure does not affect the acceptance
criteria for the test, which is based on acceptable leakage. Offsite
dose projections are based in part on the leakage criteria, and since
the proposed change does not affect the leakage limits, the change does
not affect the consequences of accidents previously evaluated.
The addition of a definition of Pa, the delineation of the
specific value in the definition and associated deletion of the
specific value in individual TS, is a purely administrative change that
does not affect plant design, construction, or operation. This part of
the proposed change does not, therefore, affect the probability or
consequences of an accident previously evaluated.
* [Increase in the SRV setpoint tolerance]
The licensee is proposing to increase the setpoint tolerance for
the SRVs from +1/-3% to plus or minus 3% of the setpoint. This proposed
change does not affect how the SRVs operate in response to accidents or
abnormal operating occurrences, and thus does not affect the
probability of an accident previously evaluated.
The proposed increase in setpoint tolerance results in a
potentially higher pressure at which an SRV would lift. This increase
in pressure is approximately 23 psig, based on the proposed maximum
system pressure and proposed setpoint tolerance. This 2% increase in
maximum lift pressure would result in a proportional increase in
possible offsite dose due to a postulated event. The dose increase
would be something less than 2%, since the increased pressure would
result in a corresponding increase in differential pressure, with
attendant flow losses and subsequent filtration through accident
mitigation systems accounting for the lower dose. This small increase
in postulated offsite dose is not considered a significant increase in
the consequences of accidents previously evaluated.
* [Increase in the number of automatic depressurization system
(ADS) valves that are allowed out of service]
The proposed change would allow one ADS valve to be out of service
without time limit, and changing the minimum number of ADS valves
required to be in service from seven to six. The change would allow two
ADS valves (compared to the current one) to be out of service for up to
14 days, and would require plant shutdown within 12 hours for three or
more ADS valves (compared to the current two or more) out of service.
The proposed change would reduce the likelihood of an inadvertent
opening ADS/SRV as an initiating event if one SRV were out of service.
Having more than one ADS/SRV out of service would also reduce this
likelihood, although the reduction would have minimal effect since the
probability of multiple ADS/SRVs being out of service with a concurrent
lift and failure to close of an SRV is small. Thus the proposed change
would have little effect on the probability of accidents previously
evaluated.
The ADS/SRVs serve to limit overpressurization of the RPV and
connected primary piping. Having ADS valves out of service would
increase the consequences compared to the current TS which limit any
ADS valves out of service for more than 14 days. Reanalysis of the
effect of ADS valves on protection of the RPV demonstrated that five
ADS valves would prevent RPV overpressurization as the previous
analysis using six ADS valves. Thus the proposed change would not
affect the consequences of accidents previously evaluated.
2. Does the amendment create the possibility of a new or different
kind of accident from any accident previously evaluated?
* [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]
Equipment that could be impacted by power uprate has been evaluated
by the licensee. The proposed change has not introduced any new
operating modes, equipment lineups, accident scenarios, or equipment
failure modes. The full spectrum of accident considerations defined in
Regulatory Guide 1.70 has been reviewed, and no new or different kind
of accident has been identified. Power uprate uses existing technology
and applies it within the capabilities of existing plant equipment in
accordance with existing regulatory criteria including NRC-approved
codes, standards, and methods. General Electric has designed to higher
power levels than the uprated power of WNP-2, and no new power-
dependent accidents have been identified. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
* [Increase in the maximum allowable reactor steam dome operating
pressure limit]
The proposed change does not involve the addition of new equipment,
nor has it introduced any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Increase in the average power range monitor (APRM) and flow
biased scram and rod block monitor (RBM) setpoints]
The proposed setpoint changes do not involve the addition of new
equipment, nor do they introduce any new operating modes, equipment
lineups, accident scenarios, or equipment failure modes. The proposed
changes do not, therefore, create the possibility of a new or different
kind of accident from any accident previously evaluated.
* [Increase in main steam line (MSL) high flow differential
pressure setpoint to reflect the increased operating pressure]
The proposed setpoint change does not involve the addition of new
equipment, nor does it introduce any new operating modes, equipment
lineups, accident scenarios, or equipment failure modes. The proposed
change does not, therefore, create the possibility of a new or
different kind of accident from any accident previously evaluated.
* [Revised temperature/pressure limit curves to reflect the higher
neutron flux over vessel life]
The proposed change does not involve the addition of new equipment,
nor does it introduce any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Increase in the pressure at which reactor core isolation cooling
(RCIC) testing occurs]
The proposed change does not involve the addition of new equipment,
nor does it introduce any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Change of the reactor protection system and End-of-Cycle/
Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a
variable setpoint based on power level]
The proposed change does not involve the addition of new equipment,
nor does it introduce any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Increase in the safety relief valve (SRV) setpoints]
The proposed change does not involve the addition of new equipment,
nor does it introduce any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Increase the value of Pa (the pressure at which primary
containment is tested). Add a new definition for Pa in the
``Definition'' section of the TS that gives a specific value of
Pa, and simplify the TS by deleting the specific value of Pa
and 1.10 Pa from other locations in the TS]
The proposed change does not involve the addition of new equipment,
nor do it introduce any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Increase in the SRV setpoint tolerance]
The proposed change does not involve the addition of new equipment,
nor does it introduce any new operating modes, equipment lineups,
accident scenarios, or equipment failure modes. The proposed change
does not, therefore, create the possibility of a new or different kind
of accident from any accident previously evaluated.
* [Increase in the number of automatic depressurization system
(ADS) valves that are allowed out of service]
The proposed change does not involve the addition of new equipment,
nor does it introduce any new operating modes, accident sequences, or
equipment failure modes. The proposed change does allow a new equipment
lineup by allowing operation with one ADS valve out of service with no
limitation, which is not allowed by the current TS. In addition, two
ADS valves can be out of service for an extended period of time (14
days), with shutdown within 12 hours required only for three or more
valves inoperable compared to the current two valves. This allowed
equipment lineup does not alter the operation of the ADS valves or
their impact as potential event initiators as discussed in the FSAR.
The proposed change does not, therefore, create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the amendment involve a significant reduction in a margin
of safety?*
[Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]
The plant was originally designed for operation at 105% rated steam
flow. The proposed change therefore does not affect the fuel design or
safety limits. The Maximum Axial Power Linear Heat Generation Rate
(MAPLHGR) limits remain the same. The Operating Limit Minimum Critical
Power Ratio (OLMCPR) is expected to increase by approximately 2%, which
will ensure that the margin to the Safety Limit MCPR is not affected.
The entire plant design has been reviewed to ensure that plant
equipment will perform properly and will still meet original design and
licensing criteria. The safety margins prescribed by the Code of
Federal Regulations have been maintained by meeting the appropriate
regulatory criteria. The margins provided by the application of the
ASME design acceptance criteria have been maintained, as well as other
margin-assuring acceptance criteria.
The emergency core cooling system-loss of coolant accident analysis
was conservatively performed based on two ADS valves out of service and
power level corresponding to 110% of original rated steam flow (plus 2%
power uncertainty factor). The analyzed results remain well below the
safety margin established at the 2200 deg.F peak centerline temperature
regulatory limit.
The overpressurization and containment analyses were repeated based
on 110% of original rated steam flow, resulting in a slightly higher
peak reactor vessel pressure. The increased pressure remains below the
acceptance criteria for the design basis, which the licensee identified
as below the ASME code limit and below the applicable TS limit.
From the containment analysis, the peak containment pressure
increases from 34.7 psig to 35.1 psig. This remains below the
containment design pressure of 45 psig.
The postulated radiological doses of design basis events, including
the bounding analysis involving loss of coolant accident, were
calculated based on the uprate power level. The results remain within
the design basis established by 10 CFR Part 100.
Based on these considerations, the proposed increase in licensed
power level does not significantly reduce any margins of safety.
* [Increase in the maximum allowable reactor steam dome operating
pressure limit]
The maximum pressure is increased by the same amount as the nominal
operating pressure increase for power uprate. The increased pressure
remains below the acceptance criteria for the design basis, which the
licensee identified as below the ASME code limit and below the
applicable TS limit. Thus the proposed TS change does not reduce the
margin to safety.
* [Increase in the average power range monitor (APRM) flow biased
scram and rod block monitor (RBM) setpoints]
The APRM flow biased scram setpoints were increase by the same
percentage of power as the uprated power. This maintains the same
margin to the trip setpoint while maintaining the same scram avoidance
as originally designed. The current margin (9%) between the scram line
and rod block line is maintained for power uprate. Thus the proposed TS
changes do not reduce any margin to safety.
* [Increase in main steam line (MSL) high flow isolation
differential pressure setpoint to reflect the increased operating
pressure]
The revised safety analysis retains the current analytic basis of
140% of rated steam flow to ensure the same level of scram avoidance is
maintained. This results in approximately a 10% increase in the
differential pressure required to trip the plant. This change does not,
however, affect the assumed closure times for the main steam isolation
valves in design analyses. The closure times determine the analyzed
release of fission products during postulated accidents. Since the
closure time is unaffected by the proposed change, the change does not
affect the margin of safety.
* [Revised temperature/pressure limit curves to reflect the higher
neutron flux over vessel life]
The proposed increase in licensed power also increases the neutron
fluence on the reactor pressure vessel over the license period (40
years). The analysis for the pressure/temperature limit curves was
updated to incorporate the increased fluence, and the revised curves
maintain the same margin to postulated brittle fracture of the reactor
vessel as the current TS curves. The proposed change does not,
therefore, change a margin of safety.
* [Increase in the pressure at which reactor core isolation cooling
(RCIC) testing occurs]
The proposed change would increase the test pressure for RCIC to
conform to the increased system pressure resulting from the power
uprate. Increasing the test pressure periodically verifies that RCIC
will operate at the increased pressure resulting from power uprate. In
addition, RCIC is not credited in safety analyses for assuring that
margins of safety are maintained. The proposed change, therefore, does
not affect any margin of safety.
* [Change of the reactor protection system and End-of-Cycle/
Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a
variable setpoint based on power level]
The safety analyses assume an EOC/RPT at 30% power. This proposed
change will allow the setpoint to vary relative to turbine first stage
pressure (the current TS setpoint), but will assure that the setpoint
is set based on a fixed 30% power. This assures that the EOC/RPT
maintains the existing margin of safety.
* [Increase in the safety relief valve (SRV) setpoints]
The two low set SRV setpoints are being increased by the same
amount of operating system pressure increase resulting from the power
uprate. The SRVs at the new setpoints will relieve pressure to ensure
the reactor coolant system remains below the acceptance criteria for
the design basis, which the licensee identified as below the ASME code
limit and below the applicable TS limit. Thus the proposed TS change
does not reduce the margin to safety.
* [Increase the value of Pa (the pressure at which primary
containment is tested). Add a new definition for Pa in the
``Definition'' section of the TS that gives a specific value of
Pa, and simplify the TS by deleting the specific value of Pa
and 1.10Pa from other locations in the TS]
The pressure at which primary containment is to be tested is being
increased to reflect the analyzed results of the power uprate. The
increase will assure that the periodic verification of containment
integrity is performed at a pressure that assures that postulated
radioactive release rates remain within analyzed assumptions. This
assures that existing margins of safety are maintained.
* [Increase in the SRV setpoint tolerance]
The proposed increase in the SRV setpoint tolerance from +1/-3% to
plus or minus 3% of the setpoint has the potential to increase the
actual pressure at which the SRV would lift. This increase in pressure
is approximately 23 psig, or 2% of the design lift pressure. This
pressure is still well within the RPV design pressure that establishes
the safety margin. The proposed change would not, therefore, affect a
margin of safety.
* [Increase in the number of automatic depressurization system
(ADS) valves that are allowed out of service]
The proposed change results in an increase in the calculated peak
centerline temperature (PCT) for postulated accidents, specifically,
the loss of coolant accident (LOCA). The margin of safety for this
parameter at WNP-2 is 2200 deg.F, which is also the regulatory limit.
Maintaining PCT less than 2200 deg.F ensures cladding integrity,
including the safety margin established by regulation. The proposed
change does not, therefore, affect a margin of safety.
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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Theodore R. Quay
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington
Date of amendment request: October 31, 1994
Description of amendment request: The proposed amendment would
modify the Technical Specifications (TS) to: (1) add two action
statements that would provide allowed outage times for either one or
both of the scram discharge volume (SDV) vent or drain valves less
stringent that the current requirements of TS 3.0.3.; and (2) change
the surveillance requirements for the SDV vent and drain valves to
conduct the testing during shutdown conditions rather than at power as
currently required.
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?
[Adding action statements for allowed outage times]
The primary functions of the SDV vent and drain valves are to
isolate the SDV following a scram to stop leakage of reactor coolant
past the CRD seals and to reopen following a scram to reset to drain
the reactor coolant from the SDV to the reactor building equipment
drain sump. The SDV is sized to accept CRD over piston discharge
water from all 185 CRDs. The SDV vent and drain valves reopen when
the scram signal is reset to provide assurance that there is
sufficient SDV volume available to accept the CRD discharge in the
event of another scram. Each vent and drain line contains two
redundant valves in series, which close to isolate the SDV on a
scram signal.
Proposed Action Statement 3.1.3.1.d allows 7 days to repair an
inoperable valve when the redundant valve is still operable, or
isolate the affected line. The reliability of the isolation function
is reduced during the period of the 7 day Allowed Outage Time (AOT)
since a single failure could prevent a redundant valve from
isolating a SDV vent or drain line. As a result, the proposed 7 day
AOT introduces an increase in the risk of an unisolated path for
reactor coolant release to the reactor building floor drain or
equipment drain sump. However, the reduction in reliability can only
affect plant safety if the redundant valve should fail to close
during an accident involving core damage. The probability of two
valves in series failing to isolate the SDV upon demand is about 4.9
x 10-5. The redundant valve is designed to automatically close
in response to a scram signal or upon loss of air or electrical
power. The probability of having one valve fail a surveillance test,
combined with a subsequent scram and redundant valve failure during
the 7 day AOT following the surveillance is about 8.2 x 10-7
per year. The risk of failure to isolate the SDV is increased by
about 2% (as determined by the WNP-2 Individual Plant Evaluation
(IPE)) by proposed Action Statement 3.1.3.1.d, which is not
considered significant.
A failure of the SDV to isolate during a scram generally does
not pose a hazard because the reactor coolant from CRD discharge and
seal leakage is routed to the reactor building drain sumps. The
release of reactor coolant can be terminated by resetting the scram
from the control room, which would close the scram outlet valves, or
by manually closing the isolation valves located in the reactor
building. Failure of the SDV to isolate can pose a risk of higher
consequences if a core damage accident occurs simultaneously. The
increased risk of activity release during a core damage accident due
to addition of the 7 day AOT is less than 1 x 10-9 per year.
Proposed Action Statement 3.1.3.1.e allows 8 hours for repair
when both valves in a line are inoperable. If a scram should occur
during this 8 hour period, both valves in a SDV vent or drain line
could fail to isolate the line, creating a path for reactor coolant
release to the reactor building floor drain or equipment drain sump.
The probability of having to enter the proposed action statement is
low, at about 4.9 x 10-5 per year. The probability of entering
the action statement, combined with a scram occurring in the
following 8 hours is about 2.6 x 10-7 per year. This represents
less than a 1% increase in the probability of SDV isolation failure
upon demand during normal operation, which is not considered
significant.
As stated above, a failure of the SDV to isolate during a scram
generally does not pose a hazard. The reactor coolant from CRD
discharge and seal leakage is routed to the reactor building
equipment drain sump. The release of reactor coolant can be
terminated by resetting the scram from the control room or by
manually closing the isolation valves located in the reactor
building. Failure of the SDV to isolate can pose a risk of higher
consequences if a core damage accident occurs simultaneously. The
increased risk of radioactivity release during a core damage
accident due to addition of the 8 hour AOT is less than 1 x 10-
9 per year. To date, there have not been any instances at WNP-2
where both valves in a SDV vent or drain line were inoperable at the
same time during plant operation. If this unlikely event were to
occur, the proposed action statement would require the affected line
to be isolated within 8 hours. With a SDV vent or drain line
isolated, normal operational leakage from the scram outlet valves
would cause the SDV instrument volume level to increase. However,
ample time would exist after receipt of a SDV high level alarm in
the control room to drain the SDV instrument volume to prevent
actuation of an automatic scram on high SDV level.
Since it is unlikely that both valves in an SDV vent or drain
line would be inoperable and isolated, the periodic opening of an
isolated line under administrative control for SDV instrument volume
venting and draining in accordance with proposed Note
is expected to be very infrequent. In addition,
Proposed Note does not adversely affect plant safety by
permitting separate action statement entry for each vent and drain
line since the probability of entering the proposed action
statements is low.
Based on the information presented above, it is concluded that
the AOTs proposed in Action Statements 3.1.3.1.d and 3.1.3.1.e and
the associated Notes do not represent changes that involve a
significant increase in the probability of an accident previously
evaluated.
As discussed above, the proposed AOTs and the associated Notes
introduce a small increase in the risk of an unisolated path for
reactor coolant release to the reactor building floor drain or
equipment drain sump through an unisolated SDV vent or drain line.
However, this event is bounded by the *NUREG-0803 evaluation of the
consequences of a postulated reactor scram and SDV rupture. Based on
the NUREG safety evaluation, the volume of coolant lost via the
bounding leakage pathway is relatively small (approximately 550 gpm
or the equivalent of a 1.008 inch break), and adequate core cooling
would be maintained such that no fuel failures are predicted for the
event. The NRC staff concluded in the NUREG safety evaluation that
resulting reactor building flooding for this event did not adversely
impact safety-related equipment. The NRC staff also concluded that
the area of the reactor building where the leak occurs will become
contaminated only to the activity level normally present in the
reactor coolant, and offsite doses would be well within the 10 CFR
Part 100 reference values for plants operating with Standard
Technical Specification (STS) coolant activity limits (0.2
microCuries/gm). The release of reactor coolant through an
unisolated SDV vent or drain line can be terminated by resetting the
scram from the control room, which would close the scram outlet
valves, or by manually closing the isolation valves located in the
reactor building. The NUREG-0803 evaluation determined that the
reactor building would be accessible to terminate leakage during a
postulated reactor scram and SDV rupture with appropriate
radiological precautions. In addition, the SDV vent and drain lines
route the release in a controlled manner to the reactor building
floor drain and equipment drain sumps. Thus, the consequences are
significantly less than those of the SDV rupture analysis.
The WNP-2 FSAR Accident Analyses, Section 15.6, ``Decrease in
Reactor Coolant Inventory,'' acceptance limits for radiological
consequences are based on the guidance set forth in 10 CFR Part 100.
Since WNP-2 operates in accordance with the STS coolant activity
limits and the offsite dose reference values of 10 CFR Part 100, the
consequences established in the NUREG-0803 safety evaluation bound
the consequences of an unisolated SDV vent or drain line event.
Therefore, the AOTs proposed in Action Statements 3.1.3.1.d and
3.1.3.1.e and the associated Notes do not represent changes that
involve a significant increase in the consequences of an accident
previously evaluated.
[Changing surveillance requirements]
The primary functions of the SDV vent and drain valves are to
isolate the SDV following a scram to stop leakage of reactor coolant
past the CRD seals and to reopen following a scram reset to drain
the reactor coolant from the SDV to the reactor building equipment
drain sump. The basis for Surveillance Requirement 4.1.3.1.4.a is to
verify SDV vent and drain valve operability so that the SDV will be
available when needed to accept CRD over piston discharge water and
so that the reactor coolant collected in the SDV will be isolated
from the secondary containment (reactor building). Performance of
Surveillance Requirement 4.1.3.1.4.a, with the proposed deletion of
the control rod configuration and density requirements and
associated Note *, will still ensure that the safety functions and
operability requirements are met.
Valve operability can be demonstrated from shutdown conditions
even though the surveillance test conditions of nearly ambient
temperature and pressure and reduced CRD discharge flow do not match
power conditions. Maximum SDV back pressure and CRD discharge flow
will not significantly affect the SDV vent and drain valves closure
rates. As verified by testing at LaSalle County Station and WNP-2,
there is only approximately a 1 second difference in SDV vent and
drain valve closing time from a scram at less than or equal to 50%
rod density versus the closure time from either the test pushbuttons
or a cold shutdown scram with all rods full in. These test results
show that the differences in temperatures, pressures, and CRD
discharge flows between power and cold shutdown conditions have a
negligible effect on SDV vent and drain valves closing times. [In
addition, the valves will be tested in the open direction at cold
shutdown conditions during the same test conducted for valve
closure.] Although the ability of the valves to open against full
reactor pressure cannot be demonstrated during shutdown conditions,
[additional verification of the valves' ability to open will be]
verified [sic] as part of the scram recovery procedure. Thus, the
ability of the valves to open against full reactor pressure will
still be demonstrated after each reactor scram during operation.
Since the operability of the SDV vent and drain valves can be
demonstrated by performing Surveillance Requirement 4.1.3.1.4.a
during shutdown conditions, the change does not represent 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?
[Adding action statements for allowed outage times]
The AOTs proposed in Action Statements 3.1.3.1.d and 3.1.3.1.e
and the associated Notes do not involve any changes to the facility
or operation of the facility as described in the WNP-2 FSAR. The
isolation function of the SDV vent and drain valves to prevent the
discharge of reactor coolant into the reactor building floor drain
and equipment drain sumps following a scram is maintained. The
release of reactor coolant through a SDV vent or drain line during a
scram can be isolated either by automatic closure of the redundant
valve in response to the scram signal or by manual isolation of the
affected line. The valves will also close automatically upon loss of
air to the valves or electrical power to the associated solenoid
pilot valves. The alarm, control rod withdrawal block, and reactor
scram functions on increasing water level in the SDV instrument
volume are unaffected by the proposed amendments. Although the
proposed AOTs will change the method of plant operation, potentially
resulting in a reduction in SDV vent or drain line isolation
reliability, the potential release to the reactor building drain
sumps has been previously evaluated in NUREG-0803, and as shown in
(1) above, the associated probabilities of the event are acceptably
low.
Since the AOTs proposed in Action Statements 3.1.3.1.d and
3.1.3.1.e and the associated Notes do not involve a change to the
facility or the method of operation that has not been previously
evaluated, the change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
[Changing surveillance requirements]
The deletion of the control rod configuration and density
requirements and associated Note * proposed for Surveillance
Requirement 4.1.3.1.4.a only change the conditions under which the
surveillance is performed. As such, the change does not involve a
change to the facility or method of operation as describe in the
WNP-2 FSAR. As discussed in (1) above, performance of the
surveillance with the proposed changes will still demonstrate SDV
vent and drain valve operability to ensure that the SDV will perform
as evaluated in the FSAR accident analysis.
Since the performance of Surveillance Requirement 4.1.3.1.4.a
during shutdown conditions does not involve a change to the facility
or the method of operation, the change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
[Adding action statements for allowed outage times]
As discussed in (1) above, the AOTs proposed in Action
Statements 3.1.3.1.d and 3.1.3.1.e and the associated Notes
introduce a small increase in the risk of an unisolated path for
reactor coolant release to the reactor building floor drain and
equipment drain sumps through an unisolated SDV vent or drain line.
The increased risk is due to the reduction in isolation function
reliability when SDV vent and drain valves are inoperable. Since
this reduction in reliability can only affect plant safety during a
scram, it is expected that the increased risk of release due to a
scram attributed to the AOTs will be offset by the reduced risk of a
scram that will result from a reduction in the number of manual
plant shutdowns. Currently, if one or more of the SDV vent or drain
valves is discovered to be inoperable, an immediate plant shutdown
is required. Establishment of the AOTs will eliminate these
unnecessary plant shutdowns that limit plant operational flexibility
and increase the risk of a plant scram and challenges to safety
systems. Moreover, there is only a small probability of a scram
occurring during the AOTs coincident with the failure of a SDV vent
or drain line to isolate. The release of reactor coolant to the
reactor building floor drain and equipment drain sumps through an
unisolated line can be terminated by resetting the scram from the
control room or by manually closing the isolation valves.
Furthermore, the consequences of such an event are bounded by the
consequences of the postulated reactor scram and SDV rupture event
evaluated in NUREG-0803 and would be well within the 10 CFR Part 100
reference values.
Since the AOTs proposed in Action Statement 3.1.3.1.d and
3.1.3.1.e and the associated Notes do not change the assumptions or
increase the consequences of the bounding NUREG-0803 accident
analysis, the margin to the 10 CFR Part 100 reference values is not
changed. Therefore, the change does not involve a significant
reduction in the margin of safety.
[Changing surveillance requirements]
Performance of Surveillance Requirement 4.1.3.1.4.a with the
proposed deletion of the control rod configuration and density
requirements and associated Note * will still ensure that the SDV
vent and drain valve safety functions and operability requirements
are met. Valve operability can be demonstrated from shutdown
conditions even though the surveillance test conditions of nearly
ambient temperature and pressure at shutdown and reduced CRD
discharge flow do not match power conditions. As discussed in (1)
above, the difference in test conditions represents only
approximately a 1 second difference in the 30 second (as specified
in Surveillance Requirement 4.1.3.1.4.a.1) SDV vent and drain valve
closing times.
The potential reduction in safety margin is related to the
reliability of the SDV vent and drain valves to close within the
required time to contain the reactor coolant leakage past the CRD
seals following a scram. The consequences of the valves failing to
close to isolate a line are bounded by the postulated reactor scram
and SDV rupture event evaluated in NUREG-0803, which assumes a
constant leakage rate for 4 hours. Since the NUREG evaluation
concluded that the consequences of such an event would be well
within the 10 CFR Part 100 reference values, the potential one
second difference in valve closing time is relatively insignificant.
In addition, the proposed surveillance requirement would eliminate
approximately 20 scrams at power over the remaining life of the
plant and prevent the concomitant transients and challenges to
safety systems. This would be expected to increase the reliability
of the SDV vent and drain valves and mitigate any reduction in
safety margin. Although performance of the proposed surveillance
requirement during shutdown conditions will not demonstrate the
ability of the valves to open against a back pressure equal to full
reactor pressure, the valves are verified to be open as part of [the
test conducted at shutdown conditions as well] the scram recovery
procedure. Thus, the ability of the valves to open against full
reactor pressure will still be demonstrated after each reactor scram
during operation.
Since the performance of Surveillance Requirement 4.1.3.1.4.a
during shutdown conditions does not change the assumptions or
increase the consequences of the bounding NUREG-0803 accident
analysis, the margin to the 10 CFR Part 100 reference values is not
changed. Therefore, the change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, N.W., Washington, D.C. 20005-3502
NRC Project Director: Theodore R. Quay
Washington Public Power Supply System, Docket No. 50-397, Nuclear
Project No. 2, Benton County, Washington Date of application for
amendment: October 31, 1994
Brief description of amendment: The proposed amendment would revise
the technical specifications to remove the requirements related to
operability and surveillance testing of the safety/relief valve (SRV)
position indication 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. The NRC staff has reviewed the licensee's analysis
against the standards of 50.92(c). The NRC staff's review is presented
below.
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
The proposed change would relocate the SRV position indication
instrumentation TS requirements to other licensee-controlled documents.
This is an administrative change which does not involve any
modification to plant equipment or plant operation as described in the
WNP-2 Final Safety Analysis Report (FSAR). The instrumentation would
continue to be available to provide SRV position indication to the
operators. Therefore, the proposed amendment does not involve an
increase in the probability or consequences of a previously evaluated
accident.
2. Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed change does not involve any physical alteration to
plant equipment and results in no changes in the performance of any
safety-related system. Therefore, the proposed amendment does not
create the possibility of a new or different kind of accident from any
previously evaluated accident.
3. Does the proposed amendment involve a significant reduction in
the margin of safety?
As noted above, the proposed change does not involve any
modification to plant equipment or plant operation. The proposed change
does not affect any accident analyses contained in the WNP-2 FSAR.
Therefore, the proposed amendment does not involve a significant
reduction in the 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: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502
NRC Project Director: Theodore R. Quay
Notice Of Issuance Of Amendments To Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, the Gelman Building, 2120 L
Street, NW., Washington, DC, and at the local public document rooms for
the particular facilities involved.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: August 2, 1994
Brief description of amendments: The amendments revise Technical
Specifications 3.9.1 and 3.1.2.7 to clarify the requirements when the
required boron concentration is greater than 2300 ppm.
Date of issuance: November 29, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: Unit 1 - 201 - Unit 2 - 179
Facility Operating License Nos. DPR-53 and DPR-69: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47164). The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated November 29, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of application for amendment: August 25, 1994
Brief description of amendment: The amendment will revise the Total
Allowance, Z, S and Allowable Values for the steam generator (SG) level
instrument calculations. However, the trip setpoints relating to SG
level will not be changed.
Date of issuance: December 9, 1994
Effective date: December 9, 1994
Amendment No. 52
Facility Operating License No. NPF-63. Amendment revises the
Technical Specifications
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49425). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 9, 1994. No significant
hazards consideration comments received: No
Local Public Document Room location: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
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: September 19, 1994
Brief description of amendments: The amendments revise the
Technical Specifications by reducing the frequency for testing the
containment spray system spray nozzles.
Date of issuance: November 28, 1994
Effective date: November 28, 1994
Amendment Nos.: 159 and 147
Facility Operating License Nos. DPR-39 and DPR-48. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51618). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 28, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Waukegan Public Library, 128
N. County Street, Waukegan, Illinois 60085.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: August 25, 1994
Brief description of amendments: The amendments revise the testing
interval for auxiliary feedwater (AFW) system pumps from monthly to
quarterly on a staggered test basis. The amendments are consistent with
the guidance in NUREG-1366, ``Improvements to Technical Specifications
Surveillance Requirements'' and Generic Letter 93-05, ``Line-Item
Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.'' In addition, a note
is incorporated from NUREG-1431, ``Revised Standard Technical
Specifications, Westinghouse Plants'' into the TS clarifying that the
turbine-driven AFW pump cannot be tested until the required pressure
exists in the secondary side of the steam generator.
Date of issuance: December 8, 1994
Effective date: To be implemented within 30 days from the date of
issuance.
Amendment Nos.: 126 and 120
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: 59 FR 51619 dated
October 12, 1994. The Commission's related evaluation of the amendments
is contained in a Safety Evaluation dated December 8, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: September 26, 1994
Brief description of amendment: The amendment revises the license
condition regarding the ``Plan for the Long Range Planning Program.''
The amendment revises the Plan by changing the semi-annual reporting
period to annual, and to reflect refined evaluation criteria and
assessment methodology, and to incorporate necessary changes to the
license condition wording.
Date of issuance: November 28, 1994
Effective date: As of the date of issuance.
Amendment No.: 173
Facility Operating License No. DPR-16. Amendment revised the
license.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53840). The Commission's related evaluation of this amendment is
contained in a Safety Evaluation dated November 28, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, NJ 08753.
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of application for amendment: September 26, 1994
Brief description of amendment: The amendment revises the plant
operating license for Three Mile Island, Unit 1 (TMI-1) by changing the
license condition regarding the ``Plan for the Long Range Planning
Program.'' The amendment revises the Plan by changing the semi-annual
reporting frequency to annual, reflects refined evaluation criteria and
assessment methodology, and incorporates necessary changes to the
license condition wording.
Date of issuance: November 28, 1994
Effective date: As of its date of issuance.
Amendment No.: 192
Facility Operating License No. DPR-50. Amendment revised License
Condition No. 2.C.9.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53841) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 28, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, PA 17105.
Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook,
Nuclear Plant, Unit No. 2, Berrien County, Michigan
Date of application for amendment: February 22, 1994
Brief description of amendment: The amendment revises the Technical
Specifications (TS) for pressure vessel heatup and cooldown curves and
extends the applicability from 12 effective full power years (EFPYs) of
operation to 15 EFPYs. The TS changes are based on an analysis of the
Cook Unit 2 surveillance capsule U which was removed after exposure of
8.65 EFPYs.
Date of issuance: November 25, 1994
Effective date: November 25, 1994
Amendment No.: 171
Facility Operating License No. DPR-74. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14891) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 25, 1994.No significant
hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile
Point Nuclear Station, Unit 2, Oswego County, New York
Date of application for amendment: August 26, 1994
Brief description of amendment: The amendment revises the Action
statements for TS 3.6.1.3, ``Primary Containment Air Locks,'' to allow
continued plant operation if the containment air lock interlock
mechanism becomes inoperable, provided an operable door of the air lock
is locked shut and is verified locked shut at least once per 31 days.
Date of issuance: November 29, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 59
Facility Operating License No. NPF-69: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49431) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 29, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Northeast Nuclear Energy Company, Docket No. 50-245, Millstone
Nuclear Power Station, Unit 1, New London County, Connecticut
Date of application for amendment: September 9, 1994
Brief description of amendment: The amendment to the Technical
Specifications deletes the requirement for a special test of the
alternate train when one train is inoperable.
Date of issuance: November 28, 1994
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 76
Facility Operating License No. DPR-21. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53842) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 28, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
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 30, 1994
Brief description of amendment: The amendment revises the Technical
Specifications (TS) to change the trip setpoint for the 4kV bus
undervoltage relay (for the grid degraded voltage) from its current
value of greater than or equal to 3710 volts to its new setting of
greater than or equal to 3730 volts.
Date of issuance: November 30, 1994
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 98
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39594) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 30, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room location: Learning Resources Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, CT 06360.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: January 14, 1994, as
supplemented by letters dated March 22, July 14, September 1, October
21, and November 22, 1994
Brief description of amendments: The amendments revise TS sections
5.5.1.1 and 5.5.3, to permit a modification to install new high density
spent fuel storage racks in each of the spent fuel pools at Limerick.
The new high density spent fuel storage racks will increase the spent
fuel pool storage capacity in each spent fuel pool to 4117 fuel
assemblies.
Date of issuance: November 29, 1994
Effective date: November 29, 1994
Amendment Nos. 82 and 43
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 8, 1994 (59 FR
40376) The supplemental letters provided clarifying information that
did not change the initial no significant hazards consideration
determination.The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 29, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Philadelphia Electric Company, Public Service Electric and Gas
Company Delmarva Power and Light Company, and Atlantic City
Electric Company, Docket Nos. 50-277 and 50-278, Peach Bottom
Atomic Power Station, Unit Nos. 2 and 3, York County, Pennsylvania
Date of application for amendments: April 15, 1994
Brief description of amendments: These amendments 1) correct a
typographical error in the Unit 3 TS, 2) reflect the name change of
Philadelphia Electric Company to PECO Energy Company, and 3) implement
line-item TS improvements recommended by Generic Letter 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation.''
Date of issuance: November 29, 1994
Effective date: November 29, 1994
Amendments Nos.: 199 and 201
Facility Operating License Nos. DPR-44 and DPR-56: Amendments
revised the Facility Operating License, Technical Specifications (TS),
and Environmental TS.
Date of initial notice in Federal Register: May 25, 1994 (59 FR
27064) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated November 29, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Government Publications
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg,
Pennsylvania 17105.
South Carolina Electric & Gas Company, South Carolina Public
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear
Station, Unit No. 1, Fairfield County, South Carolina
Date of application for amendment: July 20, 1994, as supplemented
September 20, 1994
Brief description of amendment: The amendment changes the TS to to
allow alternative, equivalent testing of diesel fuel used in the
emergency diesel generators (EDG). These alternative methods are
necessary due to recent changes in Environmental Protection Agency
(EPA) Regulations that are designed to limit the use of high sulfur
fuels.
The licensee also proposes to modify the TS by changing the
revision level of WCAP-10216-P-A, ``Relaxation of Constant Axial Offset
Control - FQ Surveillance Technical Specification,'' referenced in TS
6.9.1.11. This pertains to the FQ(z) TS (TS 3.2.1 and 3.2.2) and is
necessary since Westinghouse revised their methodology in determining
FQ(z).
Date of issuance: November 29, 1994
Effective date: November 29, 1994
Amendment No.: 121
Facility Operating License No. NPF-12. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: October 12, 1994 (59 FR
51626) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 29, 1994. No
significant hazards consideration comments received: No
Local Public Document Room location: Fairfield County Library, 300
Washington Street, Winnsboro, SC 29180
Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296,
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County,
Alabama
Date of application for amendments: March 18, 1993 (TS 331)
Brief description of amendments: The proposed changes consist of
administrative changes to the Technical Specifications for the Browns
Ferry Nuclear Plant (BFN), Units 1, 2, and 3. The changes include
deletion of requirements applicable only to BFN Unit 2 Cycle 6
operation, various administrative error corrections, eliminating
discrepancies between the Technical Specification Bases and the BFN
Final Safety Analysis Report, and clarification of certain requirements
to ensure consistency in application.
Date of issuance: December 7, 1994
Effective date: December 7, 1994
Amendment Nos.: 213, 229 and 186
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: April 1, 1993 (58 FR
17296) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 7, 1994. No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611
Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear
Plant, Unit 2, Limestone County, Alabama
Date of application for amendment: May 11, 1994 (TS 347T)
Brief description of amendment: The amendment provides a temporary
extension of the allowed outage time from 5 to 45 days for 250 volt dc
power supplies which provide control power to the plant shutdown
boards. This extension will be in effect from January 1, 1995 to
December 31, 1995 to permit the licensee to upgrade the control power
supplies with new, higher capacity components.
Date of issuance: December 7, 1994
Effective date: December 7, 1994
Amendment No.: 228
Facility Operating License No. DPR-52: Amendments revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994 (59 FR
42347) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated December 7, 1994. No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301
Point Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks
Manitowoc County, Wisconsin
Date of application for amendments: December 10, 1992, as
supplemented on March 8, 1994.
Brief description of amendments: These amendments revised Technical
Specifications (TS) Section 15.3.5, ``Instrumentation System,'' and
Section 15.4.1, ``Operational Safety Review.'' Specifically, extensive
additions and modifications were made to various tables which specify
requirements for the instrumentation and safety circuits necessary to
ensure reactor safety and provide for the automatic initiation of the
engineered safety features.
Date of issuance: December 8, 1994
Effective date: December 8, 1994
Amendment Nos.: 157 & 161
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 25, 1993 (58 FR
16236) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated December 8, 1994. The March 8,
1994, submittal, provided additional supplemental information that did
not change the initial proposed no significant hazards consideration
determination. No significant hazards consideration comments received:
No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
Nuclear Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: September 17, 1993, as
supplemented on August 31, 1994.
Brief description of amendment: The amendment revises the Kewaunee
Nuclear Power Plant (KNPP) Technical Specifications (TS) by
incorporating technical and administrative changes to TS 4.5, Emergency
Core Cooling System and Containment Air Cooling System Tests; TS 4.7,
Main Steam Isolation Valves; and Table TS 4.1-3, Minimum Frequencies
for Equipment Tests. Changes have been made to the safety injection
(SI) system automatic initiation test; the internal containment spray
system (ICS) flow blockage test; the SI, ICS and residual heat removal
pumps' periodic tests; the main steam isolation valves' test; and the
periodic control rod functional test.
Date of issuance: November 30, 1994
Effective date: November 30, 1994
Amendment No.: 114
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 19, 1994 (59 FR
2874) The August 31, 1994, submittal, changed the wording of TS
4.5.a.2.B to specify the number of spray nozzles required for the
system to function properly. This modification was not outside the
scope of the original notice and did not change the initial proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated November 30, 1994. No significant hazards consideration comments
received: No.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf
Creek Generating Station, Coffey County, Kansas
Date of amendment request: July 15, 1994
Brief description of amendment: This amendment revises Technical
Specification Table 4.3-3, Radiation Monitoring Instrumentation for
Plant Operations Surveillance Requirements, to change the analog
channel operational test interval for selected radiation monitors from
monthly to quarterly.
Date of issuance: November 28, 1994
Effective date: November 28, 1994, to be implemented with 30 days
of the date of issuance
Amendment No.: 80
Facility Operating License No. NPF-42. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53845) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated November 28, 1994. No
significant hazards consideration comments received: No.
Local Public Document Room locations: Emporia State University,
William Allen White Library, 1200 Commercial Street, Emporia, Kansas
66801 and Washburn University School of Law Library, Topeka, Kansas
66621
Dated at Rockville, Maryland, this 14th day of December 1994.
For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear
Reactor Regulation
[FR Doc. 94-31196 Filed 12-20-94; 8:45 am]
BILLING CODE 7590-01-F
