[Federal Register Volume 60, Number 50 (Wednesday, March 15, 1995)]
[Notices]
[Pages 14015-14040]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-60315]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice
Applications and Amendments to Facility Operating Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from February 16, 1995, through March 3, 1995.
The last biweekly notice was published on March 1, 1995.
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 April 14, 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. [[Page 14016]]
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.
Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318,
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of amendments request: January 31, 1995
Description of amendments request: The proposed amendments would
revise the Technical Specifications (TSs) for Calvert Cliffs, Unit Nos.
1 and 2, to increase the amount of Trisodium Phosphate Dodecahydrate
(TSP) located in the containment sump baskets required to be verified
by TS surveillance. The requested change is the result of an reanalysis
of the amount of TSP necessary to maintain the appropriate pH in the
containment sump water subsequent to a Loss of Coolant Accident.
Specifically, the request would change the TS value of TS 4.5.2.e.3
from the existing amount of 100 ft3 to 289 ft3. TS 4.5.2.e.4
would also be changed by moving the amounts of TSP and refueling water
storage tank water to be used in the required tests to the TS Bases
Section 3/4.5.2 and 3/4.5.3. These Bases sections would also be changed
by modifying the test methods.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Would not involve a significant increase in the probability
orconsequences of an accident previously evaluated.
Trisodium Phosphate Dodecahydrate (TSP) is stored in the
containment lower level to raise the pH of the sump and spray water
following a Loss of Coolant Accident (LOCA). As the pH of the water
increases, more radioactive iodine is kept in solution and the
possibility of airborne radioactivity leakage is decreased. An
additional advantage of a higher pH is the beneficial reduction in
chloride stress corrosion cracking of metal components in the
containment following an accident.
This chemical is an accident mitigator, not an accident
initiator in that it is not used until after an accident has
occurred. At the time it goes into solution, the accident has
occurred, containment spray has been activated and water has
collected in the containment sump. Therefore, increasing the
Technical Specification minimum amount verified to be in each
containment will not involve a significant increase in the
probability of an accident previously evaluated.
Updated Final Safety Analysis Report, Chapter 14.24, ``Maximum
Hypothetical Accident'', uses an assumption of a pre-RAS minimum
containment spray pH of 5.0 for the iodine removal calculation and a
post-RAS sump pH of 7.0 for iodine retention. Raising the pH to 7.0
does not increase the consequences of an accident previously
evaluated.
The proposed change to Technical Specification 4.5.2.e.4 would
remove the amounts of chemical and water used in the test to the
Bases. This relocation will not alter the test method or acceptance
criteria, but will allow adjustments to the ratio of TSP and borated
water under the controls of 10 CFR 50.59 to reflect changes in plant
conditions. In the Bases, the amount of TSP used in the test is
changed to reflect the ratio of TSP to water that would be found in
the containment following a LOCA. The specified concentration of
boron in the test reflects the highest concentration that could be
found in the containment following a LOCA. The test temperature is
changed to 120 deg.F which is well below the temperature expected to
be found in the containment sump following a LOCA. The decanting of
the solution does not change the intent of the test method since the
dissolving period will still be conducted without agitation.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Would not create the possibility of a new or different type
of accident from any accident previously evaluated. [[Page 14017]]
The addition of more TSP does not represent a significant change
in the configuration or operation of the plant. Trisodium Phosphate
Dodecahydrate is currently present in the containment lower level.
There are no physical changes which result from the increase in
volume. The proposed change to Technical Specification 4.5.2.e.4
would move the amounts of chemical and water used in the test to the
Bases. This relocation will not alter the test method or acceptance
criteria, but will allow adjustments to the ratio of TSP and borated
water under the controls of 10 CFR 50.59 to reflect changes in plant
conditions. In the Bases, the amount of TSP used in the test is
changed to reflect the ratio of TSP to water that would be found in
the containment following a LOCA. The specified concentration of
boron in the test reflects the highest concentration that could be
found in the containment following a LOCA. The test temperature is
changed to 120 deg.F which is well below the temperature expected to
be found in the containment sump following a LOCA. The decanting of
the solution does not change the intent of the test method since the
dissolving period will still be conducted without agitation.
Therefore, this change would not create the possibility of a new
or different type of accident from any accident previously
evaluated.
3. Would not involve a significant reduction in a margin of
safety.
Trisodium Phosphate Dodecahydrate is stored in the containment
lower level to raise the pH of the sump and spray water following a
LOCA. As the pH of the water increases, more radioactive iodine is
kept in solution and the possibility of airborne radioactivity
leakage is decreased. Additionally, a higher pH has a beneficial
effect on chloride stress corrosion cracking of metal components in
the containment.
Technical Specification 4.5.2.e.3 requires verification that a
minimum volume of TSP is contained in the storage baskets in each
containment. This change proposes to increase that volume. The
increased volume will ensure the containment sump, when filled with
water, will have an acceptable pH following a LOCA.
The proposed change to Technical Specification 4.5.2.e.4 would
move the amounts of chemical and water used in the test to the
Bases. This relocation will not alter the test method or acceptance
criteria, but will allow adjustments to the ratio of TSP and borated
water under the controls of 10 CFR 50.59 to reflect changes in plant
conditions. In the Bases, the amount of TSP used in the test is
changed to reflect the ratio of TSP to water that would be found in
the containment following a LOCA. The specified concentration of
boron in the test reflects the highest concentration that could be
found in the containment following a LOCA. The test temperature is
changed to 120 deg.F which is well below the temperature expected to
be found in the containment sump following a LOCA. The decanting of
the solution does not change the intent of the test method since the
dissolving period will still be conducted without agitation.
Therefore, this change would not involve a significant reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendments request involves no significant hazards consideration.
Local Public Document Room location: Calvert County Library, Prince
Frederick, Maryland 20678.
Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Ledyard B. Marsh
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of amendment request: February 9, 1995
Description of amendment request: The proposed amendment would
increase the Reactor High Water Level Trip Level Setting for the Group
1 isolation. The change will allow an increase to the main steam
isolation valve (MSIV) high water level isolation setpoint.
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:
In accordance with 10 CFR 50.91, Boston Edison submits the
following analysis addressing the no significant hazards
consideration. The proposed changes do not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Operation of the station in accordance with the proposed Trip
Level Setting will not significantly increase the probability or
consequences of an accident previously evaluated. The MSIV high
water level isolation signal is provided to protect against rapid
depressurization due to a pressure regulator malfunction during
plant startup. The high water level isolation signal is not
functional when the mode switch is in the RUN position. A high water
level in the reactor vessel indicates that fuel is covered.
Increasing the Trip Level Setting will have minimal effect on
moisture carryover in the event of a pressure regulator failure at
low reactor power. MSIV closure (Group 1) is initiated by low
reactor pressure (810 psig) approximately 30 seconds into the event.
The resulting reactor water level swell is not sufficient to reach
the bottom elevation of the main steam lines.
The proposed Technical Specification allowable value for the
Reactor Low Level Trip Level Setting and the Reactor Low Low Water
Level Trip Level setting does not involve significant increase in
the probability or consequence of an accident.
(2) Create the possibility of a new or different kind of
accident from any previously analyzed.
The proposed change does not affect the Group 1 isolation safety
function. The change does not involve any plant hardware changes
that could introduce any new failure modes or effects; thus, the
change can not create the possibility of a new or different kind of
accident from any previously analyzed.
(3) Involve a significant reduction in a margin of safety.
The proposed change does not affect the Group 1 isolation safety
function. The proposed change is consistent with the FSAR [Final
Safety Analysis Report] and Technical Specification basis associated
with reactor vessel inventory control and main steam line flooding.
The proposed change to the instrument calibration range does not
affect the margin of safety for systems or components affected by
the change. Operating Pilgrim in accordance with the proposed Trip
Level Setting 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: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company,
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
NRC Project Director: Walter R. Butler
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties,
North Carolina
Date of amendment request: February 6, 1995
Description of amendment request: The change proposes to relocate
the cycle specific core operating limits of Figure 3.1-1, Shutdown
Margin Versus Boron Concentration, from Technical Specification (TS)
3.1.1.2, Shutdown Margins - Modes 3, 4, and 5, to the Core Operating
Limits Report (COLR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change of relocating TS Figure 3.1-1, Shutdown
Margin Versus Boron Concentration to the COLR has no influence
[[Page 14018]] or impact to the probability or consequences of an
accident. The revised TS will continue to implement the shutdown
margin limits through reference to the Shutdown Margin Curve in the
COLR. In addition, the COLR is subject to the existing controls of
TS 6.9.1.6. Given that this change is an administrative relocation
of the Shutdown Margin Curve to another TS controlled document,
there would be no increase in the probability or consequences of an
accident previously evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
No safety-related equipment, safety function, or plant operation
will be altered as a result of this proposed change. The TS will
continue to require operation within the required core operating
limits. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment does not involve a significant
reduction in the margin of safety.
Relocation of the Shutdown Margin Curve to the TS controlled
COLR has no effect on the core operating limits currently in force
in TS 3.1.1.2. Future revisions to the Shutdown Margin Curve are
governed by TS 6.9.1.6 which stipulates the specific TS that
reference the COLR limits and the methodologies utilized in
developing those limits. Given that the change is an administrative
relocation of the Shutdown Margin Curve to another TS controlled
document, 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: Cameron Village Regional
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
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: January 12, 1995
Description of amendment request: The proposed amendments would
revise and clarify portions of Technical Specification (TS) Section
6.0, ``Administrative Controls,'' for the McGuire, Catawba, and Oconee
nuclear stations. The licensee submitted a combined amendment request
covering the three Duke Power nuclear stations. The proposed changes
are described below.
1. Remove the specific assignment of responsibilities for the
review, distribution, and approval activities contained in the
Technical Review and Control Section of each station's TS. The proposed
specifications state that these activities will be performed by a
knowledgeable individual/organization. Approval of the affected
documents is to be at the appropriate manager/superintendent level as
specified in Duke administrative controls.
2. Move the requirement for the review of proposed changes in the
stations' TS and Operating Licenses by the Duke Nuclear Safety Review
Board (NSRB) to Duke administrative procedures (Selected Licensee
Commitments documents) and change the wording of the requirements
covering NSRB meeting frequency. The Oconee TS covering the NSRB are
being rewritten to be consistent with McGuire and Catawba.
3. Add Technical Review and Control Program implementation and
Plant Operations Review Committee (PORC) implementation to the list of
required procedures and programs for each nuclear station.
4. Change or clarify certain TS administrative requirements
covering technical review and control activities or records retention
requirements.
5. For Oconee only, under ``Station Operating Procedures,'' revise
the TS requirements covering the review and approval of station
procedures and temporary procedure changes such that these are now
consistent with the corresponding requirements for McGuire and Catawba.
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:
(It should be noted that the licensee submitted a combined
analysis that covers McGuire, Catawba, and Oconee nuclear stations.)
Standard 1. The proposed amendments will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The provisions of these proposed amendments concern
administrative changes in the stations' Technical Specifications
involving the Technical Review and Control, Procedures and Programs/
Station Operating Procedures, and Records Retention/Station
Operating Records portions of the Administrative Controls Section.
The requested changes primarily affect review and control
activities, but also include other administrative changes affecting
the approval of station procedures (Oconee only), records retention,
and definition of the term ODCM [offsite dose calculation manual]
(McGuire and [Catawba]). The provisions of the proposed amendment
primarily involve the relocation of existing Technical
Specifications review, distribution, or approval requirements to
internal Duke administrative controls. However, implementation of
the proposed amendment does involve changes to several review/
distribution activities. Theses review/distribution activities are
primarily for: 1) Proposed changes to the stations' Technical
Specifications, 2) Proposed tests and experiments which affect
nuclear safety and are not addressed in the stations' FSAR [Final
Safety Analysis Report] or Technical Specifications, 3)
Environmental radiological procedures, 4) Reportable events
documentation and reports of violations of Technical Specifications,
5) Reports of special reviews and investigations, and 6) Reports of
unplanned onsite releases of radiological material to the environs.
Planned implementation of the proposed Technical Specifications
amendments utilizing Selected Licensee Commitments will result in
the above items being reviewed/received by a different
organizational unit in the future. The organizational unit is to be
either the recently initiated Plant Operations Review Committee
(PORC) or the General Manager, Environmental Services. Personnel
serving on the PORC, and the General Manager, Environmental Services
will be qualified based upon education and experience to review the
operational and technical considerations involved with the
applicable items listed above. No required reviews are being
eliminated by the requested amendments, only the organizational
units responsible for performing the reviews will be changed. Future
reviews of theses items under the auspices of the PORC or the
General Manager, Environmental Services will maintain a quality
level equivalent to that being currently achieved by Duke's
Qualified Reviewer Program, the Station Managers, or the
Duke Nuclear Safety Review Board as applicable. Consequently,
merely changing the organizational units performing future reviews,
or making the additional administrative changes described above,
results in no increase in the probability or consequences of an
accident previously evaluated because the review function will
continue to be conducted in an equivalent manner.
The implementing SLC will also permit proposed amendments to the
stations' Technical Specifications and Operating Licenses to be
approved for the Station Manager by a designee. However, this
individual will occupy a position equivalent to, or higher, in the
Duke organization as the Station manager.
Additionally, the proposed changes do not directly impact the
design or operation of any plant systems or components any more so
than the review and approval processes currently being conducted in
accordance [[Page 14019]] with existing approved Technical
Specifications.
Standard 2. The proposed amendments will not create the
possibility of a new or different kind of accident from any
previously evaluated.
The proposed changes are administrative in nature and primarily
cover the review, distribution, and/or approval function performed
for items identified in existing Technical Specifications. The
quality level of the future reviews will not decrease and the
ability of Duke to identify the possibility for the concurrence of
new or different kinds of accidents prior to implementation will be
maintained. Of specific interest in the consideration of Standard
2 is the review of proposed tests and experiments which
affect station nuclear safety and are not addressed in the FSAR or
Technical Specifications. The Technical Specifications required
reviews of these tests and experiments are not being proposed for
removal by these requested amendments. Only the organizational unit
conducting the review of proposed tests and experiments is being
changed by the requested amendments. The PORC, instead of the
Station Manager, is being assigned the responsibility for conducting
the reviews of proposed tests and experiments in the future. It is
believed that the combined expertise of the PORC membership will
enhance Duke's ability to identify potential situations which could
possibly involve a new, or different, kind of accident.
Standard 3. The proposed amendments will not involve a
significant reduction in any margin of safety.
The changes contained in the requested amendments are
administrative in nature and do not impact the design capabilities
or operation of any plant structures, systems, or components. There
will be no reduction in margin of safety as a result of implementing
these requested amendments. Impact upon margin of safety is a
consideration primarily included in the 10 CFR 50.59 evaluation
process conducted for station procedures, procedure changes, and
nuclear station modifications. The 10 CFR 50.59 evaluation process
in conducted under the auspices of the Duke Qualified Reviewer
Program and is not affected by these requested amendments. The
impact on margin of safety for future Technical Specifications and
Operating License changes will be reviewed by the PORC, but these
reviews will be equivalent in quality to the reviews presently
conducted by the Qualified Reviewers.
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
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: January 13, 1995
Description of amendment request: The proposed amendments would
increase the surveillance test intervals and allowed outage times for
Reactor Trip System (RTS) and Engineered Safety Features Actuation
System (ESFAS) equipment based upon analyses by Westinghouse for the
Westinghouse Owners Group and approved by the NRC. The proposed changes
to the RTS and ESFAS instrumentation are based upon WCAP-10271, its
supplements, and the NRC's safety evaluation reports.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Criterion 1 - Operation of McGuire in accordance with the
proposed license amendment[s] [do] not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The determination that the results of the proposed changes are
within all acceptable criteria was established in the SERs prepared
for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 Supplement 2,
and WCAP-10271 Supplement 2, Revision 1 issued by letters dated
February 21, 1985, February 22, 1989, and April 30, 1990.
Implementation of the proposed changes is expected to result in an
acceptable increase in total RTS yearly unavailability. This
increase, which is primarily due to less frequent surveillance,
results in an increase of similar magnitude in the probability of an
Anticipated Transient Without Scram (ATWS) and in the probability of
core melt resulting from an ATWS and also results in a small
increase in core damage frequency (CDF) due to ESFAS unavailability.
Implementation of the proposed changes is expected to result in
a significant reduction in the probability of core melt from
inadvertent reactor trips. This is a result of a reduction in the
number of inadvertent reactor trips (0.5 fewer inadvertent reactor
trips per unit per year) occurring during testing of RTS
instrumentation. This reduction is primarily attributable to testing
in bypass and less frequent surveillance.
The reduction in core melt frequency from inadvertent reactor
trips is sufficiently large to counter the increase in ATWS core
melt probability resulting in an overall reduction in total core
melt probability.
The values determined by the WOG and presented in the WCAP for
the increase in CDF were verified by Brookhaven National Laboratory
(BNL) as part of an audit and sensitivity analysis for the NRC
staff. Based on the small value of the increase compared to the
range of uncertainty in the CDF, the increase is considered
acceptable.
Changes to surveillance test frequencies for the RTS [reactor
trip system] interlocks do not represent a significant reduction in
testing. The currently specified test interval for interlock
channels allows the surveillance requirement to be satisfied by
verifying that the permissive logic is in its required state using
the permissive annunciator window. The surveillance as currently
required only verifies the status of the permissive logic and does
not address verification of channel setpoint or operability. The
setpoint verification and channel operability are verified after a
refueling shutdown. The definition of the channel check includes
comparison of the channel status with other channels for the same
parameter. The requirement to routinely verify permissive status is
a different consideration than the availability of trip or actuation
channels which are required to change state on the occurrence of an
event and for which the function availability is more dependent on
the surveillance interval. The change in surveillance requirement to
at least once every refueling does not therefore represent a
significant change in channel surveillance and does not involve a
significant increase in unavailability of the RTS.The proposed
changes do not result in an increase in the severity or consequences
of an accident previously evaluated. Implementation of the proposed
changes affects the probability of failure of the RTS but does not
alter the manner in which protection is afforded nor the manner in
which limiting criteria are established.
Criterion 2 - The proposed license amendment[s] [do] not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
The proposed changes do not result in a change in the manner in
which the RTS provides plant protection. No change is being made
which alters the functioning of the RTS (other than in a test mode).
Rather, the likelihood or probability of the RTS functioning
properly is affected as described above. Therefore, the proposed
changes do not create the possibility of a new or different kind of
accident.
The proposed changes do not involve hardware changes except
those necessary to implement testing in bypass. Some existing
instrumentation is designed to be tested in bypass and current
Technical Specifications allow testing in bypass. Testing in bypass
is also recognized by IEEE standards. Therefore, testing in bypass
has been previously approved and implementation of the proposed
changes for testing in bypass does not create the possibility of a
new or different kind of accident from any previously evaluated.
Furthermore, since the other proposed changes do not alter the
functioning of the RTS, the possibility of a new or different kind
of accident from any previously evaluated has not been created.
Criterion 3 - The proposed license amendment[s] [do] not involve
a significant reduction in a margin of safety.
The proposed changes do not alter the manner in which safety
limits, limiting safety [[Page 14020]] system setpoints, or limiting
conditions for operation are determined. The impact of reduced
testing other than as addressed above is to allow a longer time
interval over which instrument uncertainties (e.g., drift) may act.
Experience has shown that the initial uncertainty assumptions are
valid for reduced testing.
Implementation of the proposed changes is expected to result in
an overall improvement in safety by:
1) Less frequent testing will result in fewer inadvertent
reactor trips and actuation of Engineered Safety Features Actuation
System components.
2) Higher quality repairs leading to improved equipment
reliability due to longer allowable repair times.
3) Improvements in the effectiveness of the operating staff in
monitoring and controlling plant operation. This is due to less
frequent distraction of the operator and shift supervisor to attend
to instrumentation testing.
The foregoing analysis demonstrates that the proposed
amendment[s] to McGuire's Technical Specifications [do] not involve
a significant increase in the probability or consequences of a
previously evaluated accident, [do] not create the possibility of a
new or different kind of accident, and [do] not involve a
significant reduction in a margin of safety.
Based upon the preceding analysis, Duke Power Company concludes
that the proposed amendment[s] [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
Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee
Nuclear Station, Units 1, 2 and 3, Oconee County, South Carolina
Date of amendment request: January 12, 1995
Description of amendment request: The proposed amendments would
revise and clarify portions of Technical Specification (TS) Section
6.0, ``Administrative Controls,'' for the McGuire, Catawba, and Oconee
nuclear stations. The licensee submitted a combined amendment request
covering the three Duke Power nuclear stations. The proposed changes
are described below.
1. Remove the specific assignment of responsibilities for the
review, distribution, and approval activities contained in the
Technical Review and Control Section of each station's TS. The proposed
specifications state that these activities will be performed by a
knowledgeable individual/organization. Approval of the affected
documents is to be at the appropriate manager/superintendent level as
specified in Duke administrative controls.
2. Move the requirement for the review of proposed changes in the
stations' TS and Operating Licenses by the Duke Nuclear Safety Review
Board (NSRB) to Duke administrative procedures (Selected Licensee
Commitments documents) and change the wording of the requirements
covering NSRB meeting frequency. The Oconee TS covering the NSRB are
being rewritten to be consistent with McGuire and Catawba.
3. Add Technical Review and Control Program implementation and
Plant Operations Review Committee (PORC) implementation to the list of
required procedures and programs for each nuclear station.
4. Change or clarify certain TS administrative requirements
covering technical review and control activities or records retention
requirements.
5. For Oconee only, under ``Station Operating Procedures,'' revise
the TS requirements covering the review and approval of station
procedures and temporary procedure changes such that these are now
consistent with the corresponding requirements for McGuire and Catawba.
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:
(It should be noted that the licensee submitted a combined
analysis that covers McGuire, Catawba, and Oconee nuclear stations.)
Standard 1. The proposed amendments will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The provisions of these proposed amendments concern
administrative changes in the stations' Technical Specifications
involving the Technical Review and Control, Procedures and Programs/
Station Operating Procedures, and Records Retention/Station
Operating Records portions of the Administrative Controls Section.
The requested changes primarily affect review and control
activities, but also include other administrative changes affecting
the approval of station procedures (Oconee only), records retention,
and definition of the term ODCM [offsite dose calculation manual]
(McGuire and [Catawba]). The provisions of the proposed amendment
primarily involve the relocation of existing Technical
Specifications review, distribution, or approval requirements to
internal Duke administrative controls. However, implementation of
the proposed amendment does involve changes to several review/
distribution activities. These review/distribution activities are
primarily for: 1) Proposed changes to the stations' Technical
Specifications, 2) Proposed tests and experiments which affect
nuclear safety and are not addressed in the stations' FSAR [Final
Safety Analysis Report] or Technical Specifications, 3)
Environmental radiological procedures, 4) Reportable events
documentation and reports of violations of Technical Specifications,
5) Reports of special reviews and investigations, and 6) Reports of
unplanned onsite releases of radiological material to the environs.
Planned implementation of the proposed Technical Specifications
amendments utilizing Selected Licensee Commitments will result in
the above items being reviewed/received by a different
organizational unit in the future. The organizational unit is to be
either the recently initiated Plant Operations Review Committee
(PORC) or the General Manager, Environmental Services. Personnel
serving on the PORC, and the General Manager, Environmental Services
will be qualified based upon education and experience to review the
operational and technical considerations involved with the
applicable items listed above. No required reviews are being
eliminated by the requested amendments, only the organizational
units responsible for performing the reviews will be changed. Future
reviews of these items under the auspices of the PORC or the General
Manager, Environmental Services will maintain a quality level
equivalent to that being currently achieved by Duke's Qualified
Reviewer Program, the Station Managers, or the Duke Nuclear Safety
Review Board as applicable. Consequently, merely changing the
organizational units performing future reviews, or making the
additional administrative changes described above, results in no
increase in the probability or consequences of an accident
previously evaluated because the review function will continue to be
conducted in an equivalent manner.
The implementing SLC will also permit proposed amendments to the
stations' Technical Specifications and Operating Licenses to be
approved for the Station Manager by a designee. However, this
individual will occupy a position equivalent to, or higher, in the
Duke organization as the Station Manager.
Additionally, the proposed changes do not directly impact the
design or operation of any plant systems or components any more so
than the review and approval processes currently being conducted in
accordance with existing approved Technical Specifications.
Standard 2. The proposed amendments will not create the
possibility of a new or different kind of accident from any
previously evaluated.
The proposed changes are administrative in nature and primarily
cover the review, [[Page 14021]] distribution, and/or approval
function performed for items identified in existing Technical
Specifications. The quality level of the future reviews will not
decrease and the ability of Duke to identify the possibility for the
occurrence of new or different kinds of accidents prior to
implementation will be maintained. Of specific interest in the
consideration of Standard 2 is the review of proposed tests
and experiments which affect station nuclear safety and are not
addressed in the FSAR or Technical Specifications. The Technical
Specifications required reviews of these tests and experiments are
not being proposed for removal by these requested amendments. Only
the organizational unit conducting the review of proposed tests and
experiments is being changed by the requested amendments. The PORC,
instead of the Station Manager, is being assigned the responsibility
for conducting the reviews of proposed tests and experiments in the
future. It is believed that the combined expertise of the PORC
membership will enhance Duke's ability to identify potential
situations which could possibly involve a new, or different, kind of
accident.
Standard 3. The proposed amendments will not involve a
significant reduction in any margin of safety.
The changes contained in the requested amendments are
administrative in nature and do not impact the design capabilities
or operation of any plant structures, systems, or components. There
will be no reduction in margin of safety as a result of implementing
these requested amendments. Impact upon margin of safety is a
consideration primarily included in the 10 CFR 50.59 evaluation
process conducted for station procedures, procedure changes, and
nuclear station modifications. The 10 CFR 50.59 evaluation process
is conducted under the auspices of the Duke Qualified Reviewer
Program and is not affected by these requested amendments. The
impact on margin of safety for future Technical Specifications and
Operating License changes will be reviewed by the PORC, but these
reviews will be equivalent in quality to the reviews presently
conducted by the Qualified Reviewers.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Oconee County Library, 501
West South Broad Street, Walhalla, South Carolina 29691
Attorney for licensee: J. Michael McGarry, III, Winston and Strawn,
1200 17th Street, NW., Washington, DC 20036
NRC Project Director: Herbert N. Berkow
Entergy Operations Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 27, 1995
Description of amendment request: The requested change would modify
Section 5.3.1, Fuel Assemblies, of the Waterford 3 technical
specifications. The requested change increases the maximum enrichment
for the spent fuel pool and containment temporary storage rack from 4.1
to 4.9 weight percent U-235 when fuel assemblies contain fixed poisons.
Waterford 3 plans to use higher enriched fuel in the next fuel cycle
(Cycle 8) to meet the energy plans and maintain a reload batch size
similar to that used in Cycles 6 and 7.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed change will increase the fuel enrichment limit in
order to meetthe cycle energy requirements while maintaining fuel
batch sizes consistent with previous cycle designs. The calculated
k-effective, including uncertainties, demonstrate substantial margin
to criticality in the storage racks for both normal and accident
conditions. No changes to the facility are required. No new modes of
operating the fuel storage or transfer systems are required, except
a restriction to limit the use of the new fuel vault to fuel with a
maximum enrichment of 4.1 weight percent U-235. This restriction
will be implemented by administrative controls. Since the plant
equipment and operation are essentially the same, there is no
significant increase in the probability of a criticality accident.
Since a criticality event is demonstrated to be unfeasible, there
are no increased adverse consequences for such a postulated event.
As previously discussed, the proposed change will not result in
a physical change to the facility nor will it result in a
significant change to the operation of the facility; therefore, it
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change has been analyzed to establish a k-
effective, including uncertainties, at or below the NRC criticality
acceptance criteria of k-effective below 0.95 including
uncertainties at the 95/95 probability/confidence level; therefore,
there is no 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 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
GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania
Date of amendment request: January 16, 1995
Description of amendment request: The proposed amendment would
revise the TMI-1 Technical Specifications (TS) to incorporate certain
improvements from the Revised Standard Technical Specifications (TS)
for Babcock & Wilcox nuclear power plants (NUREG-1430). The amendment
would also change the bases incorporating the results of analyses to
support allowance for drift of the pressurizer code safety valve
setpoint. One of the proposed STS improvements involves a change to
Chapter 6, Administrative Controls, affecting both TMI-1 and TMI-2 TSs.
A separate notice of consideration of issuance of amendment to facility
operating license is being issued for the proposed TMI-2 TSs Change.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Operation of the facility in accordance with the proposed
amendment would not involve a significant increase in the
probability of occurrence or the consequences of an accident
previously evaluated.
The proposed amendments involve a) an administrative change to
both the TMI-1 and TMI-2 Technical Specifications which is
consistent with the B&W Standard Technical Specifications (STS),
NUREG-1430, and b) changes to the TMI-1 Technical Specifications
which are consistent with the STS. This change does not involve any
change to system or equipment configuration. The proposed amendment
revises certain surveillance requirements, extends certain
surveillance intervals as evaluated above, or involves changes that
are purely
administrative. The reliability of systems and components relied
upon to prevent or mitigate the consequences of accidents previously
evaluated is not degraded by the proposed changes. Assurance of
system and equipment availability is maintained. Therefore, this
change does not increase the probability of occurrence or the
consequences of an accident previously evaluated.
2. Operation of the facility in accordance with the proposed
amendment would not create the possibility of a new or different
kind of accident from any accident previously evaluated. The changes
only involve changes to surveillance requirements that are
consistent with STS and with the ASME Code. No new failure modes are
created and thus the changes are bounded by accidents previously
evaluated.
3. Operation of the facility in accordance with the proposed
amendment would not [[Page 14022]] involve a significant reduction
in a margin of safety. Each of these changes is compatible with the
STS and has been evaluated to preserve the level of safety assured
by the current TS.
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, Walnut Street and Commonwealth
Avenue, Box 1601, Harrisburg, PA 17105.
Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: Phillip F. McKee
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 20, 1995
Description of amendment request: The proposed amendment would
revise the fire hazards analysis for the River Bend Station (RBS) by
allowing a deviation from 10 CFR 50, Appendix R, Section III.G.3 with
respect to the requirement for a fixed fire suppression system in fire
area C-17. This area houses the control building heating, ventilation
and air conditioning (HVAC) systems and the loss due to a fire could
cause the loss of main control room habitability. C-17 does not have a
fixed fire suppression system but depends upon the use of the existing
remote shutdown system as described in the updated safety analysis
report (USAR).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1) The request does not involve a significant increase in the
probability or consequences of accident previously evaluated.
The event of concern is a fire in fire area C-17. The low fire
loading and sparse concentration of exposed combustible material in
fire area C-17 would limit fire spread. However, for this scenario
all equipment in fire area C-17 will be assumed lost. Fire area C-17
contains the air handling units for the main control room envelope.
The loss of both air handling units would cause the control building
chillers to stop running due to a logic tie requiring air flow
through the air handling equipment for the chilled water system to
operate during normal operation. The loss of the HVAC system in the
control building would cause the main control room and the equipment
rooms to begin heating up if exposed to design summer conditions.
Operator actions can be accomplished to minimize the heat up rates
for the rooms prior to the areas reaching equipment temperature
limits. This would allow the operators to begin the shutdown process
from the main control room. If the main control room continued to
heat up, the operators could accomplish the shutdown using the
remote shutdown system. HVAC for the remote shutdown panel is
located in fire area C-4 and would not be damaged by a fire in fire
area C-17. Operation of the control building HVAC system from the
remote shutdown panel bypasses the logic between the chilled water
system and the air handling system. This would allow restart of the
HVAC system for all areas except the main control room. The scenario
would conclude in a manner similar to that described in RBS USAR
Appendix 15A, Event 52, ``Reactor Shutdown From Outside Main Control
Room.''
In summary, the probability of a fire occurring in fire area C-
17 is not increased. However, if a fire were to occur in fire area
C-17 which caused the loss of main control room HVAC, the remote
shutdown system would provide an acceptable method of shutdown. The
low fire loading and sparse concentration of exposed combustible
material in fire area C-17 would limit fire spread. Therefore, this
request does not involve a significant increase in the probability
or consequences of an accident previously evaluated.
2) The request does not create the possibility of occurrence of
a new or different kind of accident from any accident previously
evaluated.
The event of concern is a fire in fire area C-17. Fire area C-17
does not have a fixed suppression system as required by 10 CFR 50,
Appendix R, Section III.G.3. Fire suppression systems are generally
used to limit fire spread, once the heat of the fire opens thermally
sensitive sprinklers. The low fire loading and sparse concentration
of exposed combustible material in fire area C-17 would limit fire
spread. However, for the purpose of event analysis, all equipment in
fire area C-17 is assumed lost. Thus a fire in fire area C-17 is
bounded by the same analysis with or
without a fixed suppression system in terms of equipment
availability.
The proposed method of shutdown for a fire in fire area C-17
will be changed in that the remote shutdown system will be credited.
Use of the remote shutdown system is bounded by RBS USAR Appendix
15A, Event 52, ``Reactor Shutdown From Outside Main Control Room.''
The HVAC for the remote shutdown panel is located in fire area C-4
and would be undamaged by a fire in fire area C-17. Operation of the
control building HVAC system from the remote shutdown panel bypasses
the logic between the chilled water system and the air handling
system. This would allow restart of the HVAC system for all areas
except the main control room.
In summary, if a fire were to occur in fire area C-17 which
caused the loss of main control room HVAC, the remote shutdown
system would provide an acceptable method of shutdown. Since, for
the purpose of event analysis, all equipment in fire area C-17 is
assumed lost, a fire in fire area C-17 is bounded by the same
analysis with or without a fixed suppression system in terms of
equipment availability. Therefore, this request does not create the
possibility of occurrence of a new or different kind of accident
from any accident previously evaluated.
3) The request does not involve a significant reduction in a
margin of safety.
In this case, the margin of safety is implicit rather than being
explicitly expressed as a numerical value. An implicit margin of
safety involves conditions for NRC acceptance. Since the RBS
Technical Specification Bases do not specifically address a margin
of safety for fire protection, the SAR, the NRC's Safety Evaluation
Report (SER), and appropriate other licensing basis documents were
reviewed to determine if the proposed change would result in a
reduction in a margin of safety. As stated, in part, in Attachment 4
to NPF-47:
EOI shall implement and maintain in effect all provisions of the
approved fire protection program as described in the Final Safety
Analysis Report for the facility through Amendment 22 and as
approved in the SER dated May 1984 and Supplement 3 dated August
1985 subject to provisions 2 and 3....
As discussed in the Reason for Request, SSER 3 dated August 1985
states, in part:
On the basis of its evaluation the staff finds that the
applicant's fire protection program with approved deviations is in
conformance with the guidelines of BTP CMEB 9.5-1, sections III.G,
III.J, and III.O of Appendix R to 10CFR50, and GDC 3, and is,
therefore, acceptable.
Thus, the margin of safety in this case can be defined as
conformance with the specified fire protection guidelines. 10 CFR
50, Appendix R, Section III.G.3, requires, in part, that alternative
shutdown capability be provided for areas where adequate separation
of redundant safe shutdown components cannot be provided. In
addition, fire detection and a fixed fire suppression system must be
installed in the area, room, or zone under consideration. Since fire
area C-17 does not have a fixed suppression system, use of the
remote shutdown system for a fire in this fire area would deviate
from the requirements of 10 CFR 50, Appendix R, Section III.G.3.
However, as discussed previously, the low fire loading and sparse
amount of exposed combustibles compensate for the lack of a fixed
fire suppression system. There is no adverse impact on the ability
to achieve and maintain safe shutdown. Therefore, this request does
not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
[[Page 14023]] 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
Northeast Nuclear Energy Company (NNECO), Docket No. 50-245,
Millstone Nuclear Power Station, Unit 1, New London County,
Connecticut
Date of amendment request: July 28, 1994
Description of amendment request: The proposed amendment would add
a footnote to Technical Specifcaiton 3.5.C. The footnote would state
that the operability of the feedwater coolant injection (FWCI) system
be independent of its seismic capability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO has reviewed the proposed change in accordance with
10CFR50.92 and concluded that the change does not involve a
significant hazards consideration (SHC). The basis for this
conclusion is that the three criteria of 10CFR50.92(c) are not
compromised. The proposed change does not involve an SHC because the
change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Any postulated failure in the non-seismic portion of the FWCI
subsystem may result in a loss of feedwater flow transient. However
comparing the probability of occurrence of a seismic event, any
increase in the probability of occurrence of a loss of feedwater
event would be small. The proposed change would have no impact on
the probability of occurrence of any other accident, including LOCAs
[loss of coolant accidents].
The FWCI subsystem will continue to be maintained as QA Category
1 (except for the seismic attribute). Therefore, it will remain
available for accident mitigation for most scenarios. Nevertheless,
LOCA analyses have been reevaluated to demonstrate that FWCI is not
necessary to show compliance with 10CFR50.46. Potentially limiting
LOCA scenarios have been analyzed without the FWCI subsystem using
an approved LOCA methodology. An active single failure was
postulated in addition to not taking credit for the FWCI subsystem.
Based on the results of these analyses, the current design basis
large and small break LOCAs remain bounding. Moreover, FWCI is not
credited in mitigating any of the non-LOCA transients/accidents.
Safe shutdown following a seismic event can be achieved using
the LPCI [low pressure coolant injection] and ESW [emergency service
water] systems, and the SRVs [safety relief valves], which are all
seismically qualified. Therefore, the FWCI system is not required to
mitigate a seismic event.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
Seismic reclassification of portions of FWCI does not create the
possibility of a new kind of an accident. The portion of the piping
up to the second isolation valve (from the RPV [reactor pressure
vessel]), is seismically qualified and will remain classified as
seismic. This ensures that a postulated failure in the non-seismic
portion of piping or components does not degrade containment
integrity or result in a blowdown of the RPV. Consequential and
environmental effects of a FW [feedwater] piping failure have been
analyzed in the HELB [high energy line break] program and have been
found to be acceptable.
3. Involve a significant reduction in the margin of safety.
All accidents, including LOCAs, can be mitigated without using
FWCI. FWCI is also not necessary for safe shutdown following a
seismic event. The intended function of the FWCI subsystem is to
reduce the likelihood of core uncovery during the lifetime of the
plant. The CS [core spray] and LPCI subsystems provide redundant and
diverse means of injecting water to the RPV. The FWCI subsystem
provides an additional diverse means to inject water. Since FWCI
will be maintained QA Category 1 (except for the seismic attribute),
it will continue to provide the additional diversity to the
injection systems. Considering the intended function of the
subsystem and the credit taken in the accident analysis,
reclassifying FWCI to be non-seismic does not significantly reduce
the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: January 9, 1995, as supplemented
February 7, 1995
Description of amendment requests: The proposed amendments would
revise Prairie Island Nuclear Generating Plant Technical Specification
(TS) 4.12, ``Steam Generator Tube Surveillance,'' to incorporate
revised acceptance criteria for steam generator tubes with degradation
in the tubesheet roll expansion region. These criteria for steam
generator tube acceptance were developed by Westinghouse Electric
Corporation and are known as F* (F-Star'') and L*
(L-Star''). These criteria would be utilized to avoid
unnecessary plugging and sleeving of steam generator tubes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed amendment[s] will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
F* Steam Generator Tube Repair Criteria
The supporting technical and safety evaluations of the subject
criterion demonstrate that the presence of the tubesheet will
enhance the tube integrity in the region of the hardroll by
precluding tube deformation beyond its initial expanded outside
diameter. The resistance to both tube rupture and tube collapse is
strengthened by the presence of the tubesheet in that region. The
results of hardrolling of the tube into the tubesheet is an
interference fit between the tube and the tubesheet. Tube rupture
cannot occur because the contact between the tube and tubesheet does
not permit sufficient movement of tube material. The radial preload
developed by the rolling process will secure a postulated separated
tube end within the tubesheet during all plant conditions. In a
similar manner, the tubesheet does not permit sufficient movement of
tube material to permit buckling collapse of the tube during
postulated LOCA [loss-of-coolant accident] loadings.
The F* length of roll expansion is sufficient to preclude tube
pullout from tube degradation located below the F* distance,
regardless of the extent of the tube degradation. The existing
Technical Specification leakage rate requirements and accident
analysis assumptions remain unchanged in the unlikely event that
significant leakage from this region does occur. As noted above,
tube rupture and pullout is not expected for tubes using the F*
criterion. Any leakage out of the tube from [[Page 14024]] within
the tubesheet at any elevation in the tubesheet is fully bounded by
the existing steam generator tube rupture analysis included in the
Prairie Island Plant USAR [Updated Safety Analysis Report]. For
plants with partial depth roll expansion like Prairie Island, a
postulated tube separation within the tube near the top of the roll
expansion (with subsequent limited tube axial displacement) would
not be expected to result in coolant release rates equal to those
assumed in the USAR for a steam generator tube rupture event due to
the limited gap between the tube and tubesheet. The proposed
plugging criterion does not adversely impact any other previously
evaluated design basis accident.
Leakage testing of roll expanded tubes indicates that for roll
lengths approximately equal to the F* distance, any postulated
faulted condition primary to secondary leakage from F* tubes would
be insignificant.
L* Steam Generator Tube Repair Criteria
The presence of the tubesheet enhances steam generator tube
integrity in the region of the hardroll by precluding tube
deformation beyond its initial expanded outside diameter. The
resistance to both tube rupture and tube collapse is strengthened by
the presence of the tubesheet in that region. The result of the
hardroll of the tube into the tubesheet is an interference fit
between the tube and the tubesheet. Tube rupture cannot occur
because the contact between the tube and tubesheet does not permit
sufficient movement of tube materials. In a similar manner, the
tubesheet does not permit sufficient movement of tube material to
permit buckling collapse of the tube during postulated LOCA
loadings.
The type of degradation for which the L* criteria has been
developed (cracking with an axial or near axial orientation) has
been found not to significantly reduce the axial strength of a tube.
An evaluation including analysis and testing has been done to
determine the strength reduction for the axial loads with simulated
axial and near axial cracks. This evaluation provided the basis for
the acceptance criteria for tube degradation subject to the L*
criteria.
The length of roll expansion above L* is sufficient to preclude
significant leakage from tube degradation located below the L*
distance. The existing Technical Specification leakage rate
requirements and accident analysis assumptions remain unchanged in
the unlikely event that significant leakage from this region does
occur. As noted above, tube rupture and pullout is not expected for
tubes using the alternate plugging criteria.
Any leakage out of the tube from within the tubesheet at any
elevation in the tubesheet is fully bounded by the existing steam
generator tube rupture analysis included in the Prairie Island
Updated Safety Analysis Report. The proposed alternate plugging
criteria do not adversely impact any other previously evaluated
design basis accident.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
F*
Implementation of the proposed F* criterion does not introduce
any significant changes to the plant design basis. Use of the
criterion does not provide a mechanism to initiate an accident
outside of the region of the expanded portion of the tube. Any
hypothetical accident as a result of any tube degradation in the
expanded portion of the tube would be bounded by the existing tube
rupture accident analysis. Tube bundle structural integrity will be
maintained. Tube bundle leaktightness will be maintained such that
any postulated accident leakage from F* tubes will be negligible
with regards to offsite doses.
L*
Implementation of the proposed alternate tubesheet tube plugging
criteria does not introduce changes to the plant design basis. Use
of the criteria does not provide a mechanism to result in an
accident outside of the region of the tubesheet expansion. Any
hypothetical accident as a result of any tube degradation in the
expanded portion of the tube would be bounded by the existing tube
rupture accident analysis.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety.
F*
The use of the F* criterion has been demonstrated to maintain
the integrity of the tube bundle commensurate with the requirements
of Reg Guide 1.121 [Bases for Plugging Degraded PWR Steam
Generator Tubes] (intended for indications in the free
span of tubes) and the primary to secondary pressure boundary under
normal and postulated accident conditions. Acceptable tube
degradation for the F* criterion is any degradation indication in
the tubesheet region, more than the F* distance below the bottom of
the transition between the roll expansion and the unexpanded tube.
The safety factors used in the verification of the strength of the
degraded tube are consistent with the safety factors in the ASME
Boiler and Pressure Vessel Code used in steam generator design. The
F* distance has been verified by testing to be greater than the
length of roll expansion required to preclude both tube pullout and
significant leakage during normal and postulated accident
conditions. Resistance to tube pullout is based upon the primary to
secondary pressure differential as it acts on the surface area of
the tube, which includes the tube wall cross-section, in addition to
the inner diameter based area of the tube. The leak testing
acceptance criteria are based on the primary to secondary leakage
limit in the Technical Specifications and the leakage assumptions
used in the USAR accident analysis.
Implementation of the tubesheet plugging criterion will decrease
the number of tubes which must be taken out of service with tube
plugs or repaired with sleeves. Both plugs and sleeves reduce the
RCS (reactor coolant system) flow margin; thus, implementation of
the F* criterion will maintain the margin of flow that would
otherwise be reduced in the event of increased plugging or sleeving.
Based on the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the USAR or the Technical Specification
Bases.
L*
The use of the alternate tubesheet plugging criteria has been
demonstrated to maintain the integrity of the tube bundle
commensurate with the requirements of Reg. Guide 1.121 for
indications in the free span of tubes and the primary to secondary
pressure boundary under normal and postulated accident conditions.
Acceptable tube degradation for the L* criteria is any degradation
indication with axial or nearly axial cracking in the tubesheet
region, more than the L* distance below the bottom of the transition
between the roll expansion and the unexpended tube. For tubes with
axial or nearly axial cracks the strength of the tube relative to an
axial load would not be reduced below the strength required to
resist potential axial loads. The safety factors used in the
verification of the strength of the degraded tube are consistent
with the safety factors in the ASME Boiler and Pressure Vessel Code
used in steam generator design. The L* distance has been verified by
testing to be greater than the length of roll expansion required to
preclude significant leakage during normal and postulated accident
conditions. The leak testing acceptance criteria are based on the
primary to secondary leakage limit in the Technical Specifications
and the leakage assumptions used in the USAR accident analyses.
Implementation of the proposed tubesheet plugging criteria will
decrease the number of tubes which must be taken out of service with
tube plugs or repaired with sleeves. Both plugs and sleeves reduce
the RCS flow margin, thus implementation of the alternate plugging
criteria will maintain the margin of flow that would otherwise be
reduced in the event of increased plugging or sleeving.
Based on the above, it is concluded that the proposed change
does not result in a significant reduction in margin with respect to
plant safety as defined in the Updated Safety Analysis Report or the
bases of the Technical Specifications.
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. This
notice supersedes the staff's previous notice which was published in
the Federal Register February 1, 1995 (60 FR 6307).
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: Cynthia Carpenter, Acting [[Page 14025]]
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: February 23, 1995
Description of amendment requests: The proposed amendments would
revise the wording in the Prairie Island technical specifications to
allow implementation of exemptions to the schedule requirements of 10
CFR Part 50, Appendix J. A related exemption request would grant
temporary relief from the requirements of 10 CFR Part 50, Appendix J,
Section III.D.1.(a) which requires Prairie Island Unit 2 to perform a
Type A test in the May 1995 refueling outage.
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 proposed amendment is an administrative change which allows
implementation of approved exemptions to the regulations and by
itself does not change any retest schedules.
Therefore, the probability or consequences of an accident
previously evaluated are not affected by the proposed amendment.
2. The proposed amendment[s] will not create the possibility of
a new or different kind of accident from any accident previously
analyzed.
The proposed amendment is an administrative change which allows
implementation of approved exemptions to the regulations and by
itself does not change any retest schedules.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated would not be created
by the proposed amendment.
3. The proposed amendment[s] will not involve a significant
reduction in the margin of safety
The proposed amendment is an administrative change which allows
implementation of approved exemptions to the regulations and by
itself does not change any retest schedules.
Therefore, a significant reduction in the margin of safety would
not be involved with the proposed amendment.
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: Cynthia Carpenter, Acting
Omaha Public Power District, Docket No. 50-285, Fort Calhoun
Station,Unit No. 1, Washington County, Nebraska
Date of amendment request: February 10, 1995
Description of amendment request: The proposed amendment to the
technical specifications (TSs) would relocate the requirements for the
incore instrumentation (ICI) system from the TS to the Updated Safety
Analysis Report (USAR).
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.
The Incore Instrumentation (ICI) System is used to measure core
power distribution for the purpose of Limiting Conditions for
Operation (LCO) monitoring of Technical Specification (TS) limits on
linear heat rate, unrodded planer radial peaking factor, unrodded
integrated radial peaking factor, and azimuthal power tilt. The ICI
System has no safety purpose itself; it measures parameters which
have safety significance. No change to the monitored parameters is
proposed. The proposed changes will relocate requirements on the
number and distribution of incore detectors used by the ICI System
when measuring these parameters from the TS to the Updated Safety
Analysis Report (USAR). Changes to the requirements can be made
without NRC approval when the changes meet the criteria of 10 CFR
50.59. Changes to the ICI System requirements that do not meet the
criteria of 10 CFR 50.59 must be approved by the NRC by license
amendment.
Relocation of the requirements on the ICI System from the TS to
the USAR does not increase the probability or consequences of any
accident previously analyzed because the ICI System is neither a
precursor nor a mitigator for any analyzed accident. The ICI System
is used to ensure that operation within the LCOs for linear heat
rate, unrodded planer radial peaking factor, unrodded integrated
radial peaking factor, and azimuthal power tilt is maintained.
However, its operation serves no mitigation function associated with
any USAR Section 14 accident analysis. The parameters measured by
the ICI System are important parameters in many accident analyses;
however, this proposed change does not remove or revise the limits
on these parameters.
Additionally, it is proposed to revise TS 2.10.4(1)(b) to
clarify its requirements. Currently TS 2.10.4(1) part (b) applies
while operating under the provisions of part (a) if the plant
computer incore detector alarms become inoperable. This is incorrect
in that part (a) applies when the linear heat rate is being
monitored by the ICI System and the linear heat rate is exceeding
its limits as indicated by valid detector alarms. Part (b) of this
specification should apply only if the linear heat rate is being
monitored by the ICI System, is within its limits, and the plant
computer incore detector alarms are inoperable.
Administrative changes are also proposed which correct grammar
and renumber/relocate portions of the TS and bases to other TS, to
correspond to the proposed change to relocate ICI System
requirements.
Therefore, the proposed change does 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 previously evaluated.
The ICI System will continue to be used to monitor TS limits on
core power distribution. There will be no physical alterations to
the plant configuration, changes to setpoint values, or changes to
the implementation of setpoints or limits as a result of this
proposed change.
The proposed change to TS 2.10.4(1)(b) only clarifies its
requirements. The proposed change is more restrictive in that TS
2.10.4(1)(b), as currently written, could be interpreted to allow
continued operation for up to seven days with the linear heat rate
exceeding its limits. The proposed change clarifies this
specification to ensure that TS 2.10.4(1)(a) is applied if the
linear heat rate is exceeded while being monitored by the ICI
System. TS 2.10.4(1)(a) requires that the linear heat rate be
restored within one hour or a plant shutdown initiated.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
(3) The proposed changes do not involve a significant reduction
in a
margin of safety.
The ICI System is used to measure core power distribution
parameters which are a direct measure of the margin of safety. The
limits on these parameters are not changed. Therefore, the proposed
change (i.e., relocation of the ICI System operability requirements
to the USAR and/or plant procedures) does not involve a significant
reduction in a margin of safety.
The proposed change to TS 2.10.4(1)(b) helps ensure that the
margin of safety is maintained by clarifying when the TS is
applicable. This clarification ensures that the more restrictive
actions of TS 2.10.4(1)(a) are taken if the linear heat rate is
exceeded while being monitored by the ICI System. Therefore, the
proposed change does not involve a significant reduction in a margin
of safety. [[Page 14026]]
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
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County, California
Date of amendment requests: December 30, 1994 (Reference LAR 94-12)
Description of amendment requests: The proposed amendments would
revise the combined Technical Specifications (TS) for the Diablo Canyon
Nuclear Power Plant, Unit Nos. 1 and 2, to revise TS 2.2, 3/4.3.1, 3/
4.3.2, 3/4.3.3, 3/4.4.4, 3/4.4.9, 3/4.5.2, 3/4.8.1, 3/4.8.2, 3/4.9.2,
3/4.9.9, and 3/4.10.3. The specific TS changes proposed are as follows:
(1) The TS issued in License Amendments (LAs) 84/83 would be
changed to (a) revise the value of the overpower Delta-temperature
(OPDT) constant K6 in TS 2.2.1, Table 2.2-1, Note 3; (b) revise the
reactor coolant system (RCS) loop Delta-T function; and (c) make
editorial corrections for clarification and consistency to TS 2.2.1
(and TS 2.2.1 Bases), TS 3/4.3.1, and TS 3/4.3.2.
In revising the RCS loop Delta-T function, the licensee would (a)
incorporate the 0.99 multiplying factor listed in TS 2.2.1, Table 2.2-
1, Note 5, and TS 3/4.3.2, Table 3.3-4, Note 2, into constants B1
through B4; (b) change ``Steam Generator (SG) Water Level Low-Low'' in
TS 3/4.3.2, Table 3.3-3 and Table 4.3-2, Functional Unit 6.c,
``Auxiliary Feedwater'' (AFW), by deleting the Mode 3 applicability of
the RCS loop Delta-T function and by adding a footnote to the Mode 3
applicability of the SG water level low-low function requiring that the
trip time delay (TTD) associated with the SG water level low-low
channel be less than or equal to 464.1 seconds; (c) change TS 3/4.3.1,
Table 3.3-1, Action 27, and TS 3/4.3.2, Table 3.3-3, Action 29, by
allowing up to four RCS loop Delta-T channels to be inoperable with the
TTD threshold power level for zero seconds time adjusted to 0-percent
rated thermal power (RTP) and by allowing the affected SG water level
low-low channels to be placed in the tripped condition, with one
inoperable RCS loop Delta-T channel; and (d) change the Table 3.3-1 and
Table 3.3-3 ``Channels to Trip'' and ``Minimum Channels Operable''
columns to not applicable (N.A.).
(2) The TS issued in LAs 70/69 would be changed to (a) delete
references to the plant vent noble gas activity monitors (RM-14A and
RM-14B) and footnote references to applicability of the containment
ventilation exhaust radiation monitors (RM-44A and RM-44B) in TS Tables
3.3-3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3 and TS 4.9.9; and (b)
revise the ``Trip Setpoint and Allowable Values'' column in TS Table
3.3-4, Functional Unit 3.c.4), to reference the offsite dose
calculation procedure (ODCP).
(3) Cycle-specific information in TS 4.3.2.1, TS 3.3.3.6, TS
4.4.4.1, TS 4.5.2, TS 3.8.1.1, TS 3.8.2.1, and TS 3.8.2.2 that is no
longer necessary would be deleted.
(4) The word ``analog'' would be deleted from TS 4.4.9.3.1, TS
4.9.2, and TS 4.10.3.2.
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:
a. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed change to the OPDT constant K6 is conservative and
will not cause any design or analysis acceptance criteria to be
exceeded. There is no effect on the structural and functional
integrity of any plant system. The OPDT function is part of the
accident mitigation response and is not itself an initiator for any
transient. This change does not affect the integrity of the fission
product barriers for mitigation of radiological dose consequences as
a result of an accident.
The proposed change to incorporate the 0.99 multiplier into the
TTD constants is an administrative change and has no effect on plant
operation. The proposed change to delete Mode 3 applicability of the
RCS Loop Delta-T function does not affect any design or analysis
results. Allowing up to 4 RCS Loop Delta-T channels to be inoperable
with the TTD threshold power level for zero seconds time delay
adjusted to 0% RTP is conservative with respect to ESFs [engineered
safety features] and reactor trip actuation time. Allowing the SG
[steam generator] water level low-low channels affected by the
inoperable RCS Loop Delta-T channels to be placed in the tripped
condition is also conservative with respect to reactor trip and AFW
pumps start. The change to the Channels to Trip and Minimum Channels
Operable columns is a clarifying change to reflect the proposed
changes to the action statements and identifies that the RCS Loop
Delta-T does not provide a reactor trip function. Therefore, the
proposed changes to the RCS Loop Delta-T function do not affect any
of the accident analysis results.
The proposed changes to revise Table 3.3-4, Functional Unit
3.c.4), and to delete cycle-specific TS, TS references to RM-14A and
RM-14B, and the word ``analog'' from the analog channel operation
test are administrative and have no effect on plant operation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
b. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The proposed change to the OPDT constant K6 does not affect the
assumed accident initiation sequences. No new operating
configuration is being imposed by the change to K6 that would create
a new failure scenario. No new failure modes are being created for
any plant equipment.
The proposed changes to the RCS Loop Delta-T function do not
involve any physical modification to any plant system or change the
methodology by which any safety-related system performs its
function.
1The proposed changes to revise Table 3.3-4, Functional Unit
3.c.4), and to delete cycle-specific TS, TS references to RM-14A and
RM-14B, and the word ``analog'' from the analog channel operation
test are administrative, would not result in any physical alteration
to any plant system, and would not be a change in the method by
which any safety-related system performs its function.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
c. Does the change involve a significant reduction in a margin
of safety?
The proposed change to the OPDT constant K6 will not affect any
accident analysis assumptions, initial conditions, or results.
The proposed changes to the RCS Loop Delta-T function do not
affect any accident analysis assumptions, initial conditions, or
results.
The proposed changes to revise Table 3.3-4, Functional Unit
3.c.4), and to delete cycle-specific TS, TS references to RM-14A and
RM-14B, and the word ``analog'' from the analog channel operation
test are administrative and clarify the TS. These proposed changes
have no effect on current operating methodologies or actions that
govern plant performance.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests [[Page 14027]] involve no significant hazards
consideration.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120
NRC Project Director: Theodore R. Quay
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: September 26, 1994
Description of amendment request: The proposed TS changes extend
surveillance test intervals and allowable out-of-service times for the
testing and/or repair of instrumentation that actuate the Reactor
Protection System, Primary Containment Isolation, Core and Containment
Cooling systems, Control Rod Blocks, Radiation Monitoring systems, and
Alternate Rod Insertion/Recirculation Pump Trip.
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.
The proposed TS changes increase the STIs and AOTs for actuation
instrumentation based on analyses described and justified in
Licensing Topical Reports (References 2 through 8) [see licensee's
September 26, 1994 application for reference information] which have
been evaluated in associated Safety Evaluation Reports. These
changes were incorporated into PBAPS Technical Specifications
consistent with NUREG-1433. TS requirements that govern Operability
or routine testing of plant instruments are not assumed to be
initiators of any analyzed event because these instruments are
intended to prevent, detect or mitigate accidents. Therefore, these
changes will not involve an increase in the probability of
occurrence of an accident previously evaluated. Additionally, these
changes will not increase the consequences of an accident previously
evaluated because the proposed change will not involve any physical
changes to plant systems, structures, or components (SSC), or the
manner in which these SSC are operated, maintained, modified, or
inspected. The changes will not alter the operation of equipment
assumed to be available for the mitigation of accidents or
transients by the plant safety analysis or licensing basis. As
justified in References 1 through 8, the proposed changes establish
or maintain adequate assurance that components are operable when
necessary for the prevention or mitigation of accidents or
transients and that plant variables are maintained within limits
necessary to satisfy the assumptions for initial conditions in the
safety analyses. These changes establish or modify time limits
allowed for operation with inoperable instrument channels based on
the analyses in References 1 through 8 and will not allow continuous
plant operation with plant conditions such that a single failure
will result in a loss of any safety function. Therefore, these
changes will not increase the 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.
These proposed changes will not involve any physical changes to
SSC, or the manner in which these SSC are operated, maintained,
modified, tested, or inspected. Therefore, these changes will not
create the possibility of a new or different kind of accident from
any accident previously evaluated. The changes in methods governing
normal plant operation are consistent with the current safety
analysis assumptions. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3) The proposed changes do not involve a significant reduction
in a margin of safety.
The proposed TS changes increase the STIs and AOTs for actuation
instrumentation based on analyses described and justified in
Licensing Topical Reports (References 2 through 8) which have been
evaluated in associated Safety Evaluation Reports. These changes
were incorporated into PBAPS Technical Specifications consistent
with NUREG-1433. These changes can be classified into one of the
following three categories:
a. Changes to the minimum STIs and AOTs for the testing and/or
repair of instrumentation based on the results of generic analyses
in References 1 through 8;
b. Changes to conditions, required actions, and completion times
needed to make PBAPS TS requirements consistent with the assumptions
used in the analyses in References 1 through 8; and,
c. Changes that reformat, renumber, and/or reword existing
requirements to incorporate the changes above.
All of the proposed changes will be incorporated into the PBAPS
custom Technical Specifications using the same approach and specific
requirements used in Reference 12.
There is no significant reduction in the margin of safety
resulting from changes to the STIs and AOTs for the testing and/or
repair of instrumentation based on the results of the analyses in
References 1 through 8. These analyses determined that there is no
significant change in the availability and/or reliability of
instrumentation as a result of this change in STIs and AOTs. PECO
Energy performed reviews that confirmed these analyses are
applicable to PBAPS and that there would be no effect on the
identification of excessive instrument setpoint drift as a result of
increasing from monthly to quarterly the minimum interval between
instrument functional tests. The proposed required actions ensure
that actions to mitigate loss of single failure tolerance is
initiated within 24 hours (12 hours for RPS) in accordance with the
results of the analyses in References 1 through 8 and action to
mitigate a loss of instrument function is initiated within 1 hour.
The proposed changes which replace the shutdown actions
associated with inoperable instrumentation with actions to declare
the supported system inoperable does not involve a reduction in a
margin of safety. The proposed changes ensure that appropriate
compensatory measures are taken commensurate with approved TS
Actions for the affected systems and the safety analyses. In
addition, the proposed changes provide the benefit of avoiding an
unnecessary shutdown transient when appropriate measures are
available to compensate for the inoperable instrumentation.
There is no significant reduction in the margin of safety
resulting from changes that reformat, renumber, and/or reword
existing requirements to incorporate the changes above.
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
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 17, 1994
Description of amendment request: The proposed changes to the
Technical Specifications (TS) are being requested to support
modifications 5384 and 5386 which upgrade the Main Stack and Vent Stack
Radiation Monitoring Systems. [[Page 14028]]
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.
Neither the Main Stack nor the Vent Stack Radiation Monitoring
Systems serve as an initiator or contributor to any accidents
previously evaluated. The systems provide indication and detection
of radioactivity and effluent release in the main and vent stacks.
The new systems perform the same function as the old, and have equal
or better performance characteristics. Installation and operation of
the new radiation monitoring systems do not degrade any active or
passive equipment that responds to an accident.
The proposed increase in the surveillance test interval of the
subject radiation monitoring systems from 12 to 18 months is
consistent with vendor recommendations, and is based on operating
experience with instrumentation of a similar design.
Therefore, the proposed changes do 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 previously evaluated.
Both modifications replace obsolete radiation monitoring
equipment and have the same failure modes as the existing equipment.
The upgraded systems are considered enhancements to the existing
systems and are considered neither a contributor nor initiator of
any accidents previously evaluated.
Based on the above, the proposed changes do not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed changes do not involve a significant reduction
in a margin of safety.
Neither the accuracy nor the responsiveness of the existing
radiation monitoring equipment will be degraded as a result of the
installation of modifications 5384 and 5386. Revisions to the
calibration and surveillance frequencies are based on vendor
information and experience with instrumentation of similar design.
The changes associated with setpoints and the lower limit of
detection are in the conservative direction. The upgraded main stack
system continues to provide a non-safety related trip signal to
Group III isolation valves during purging of the containment through
the SBGTS [standby gas treatment system]. The revisions to parameter
descriptions and instrument designation are considered
administrative.
Therefore, based on the above, 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: 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
Public Service Electric & Gas Company, Docket No. 50-311, Salem
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey
Date of amendment request: February 3, 1994, supplemented September
19, 1994, and November 23, 1994
Description of amendment request: The proposed amendment revises
the Technical Specifications to reflect a reduction in the Reactor
Coolant System flow.
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 analyzed.
No component modification, system realignment, or change in
operations will occur which could affect the probability of any
accident or transient. The proposed reduction in RCS loop and total
flow rates will not change the probability of a challenge to any
Engineered Safeguard Feature or other device. The consequences of
previously analyzed accidents have been found to remain within
acceptable licensing basis limits when the reduced flow rates are
assumed. The system transient response is not affected by the
initial RCS flow assumption, unless the initial assumption is so low
as to impair the steady-state core cooling capability or steam
generator heat transfer capability. This is clearly not the case
with a 1% reduction in RCS flow. The proposed change to the wording
of the parameter title on Table 3.2-1 is editorial for clarity.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously analyzed.
2. Create the possibility of a new or different kind of
accident.
No component modification, system realignment, or change in
operating procedure will occur which could create the possibility of
a new event not previously considered. The proposed reduction in RCS
loop and total flow rates will not initiate any new events.
Therefore, the proposed changes would not create the possibility of
a different or new kind of accident.
3. Involve a significant reduction in a margin of safety.
The proposed decrease in RCS loop and total flow rates has been
analyzed and found to have an insignificant effect on the applicable
transient analyses found in the FSAR. The proposed change to the
wording of the parameter title on Table 3.2-1 is editorial for
clarity. Therefore, the proposed changes would not involve a
significant reduction in any margin of safety.
Therefore, based on the information presented above, PSE&G has
concluded there is no 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: Salem Free Public library, 112
West Broadway, Salem, New Jersey 08079
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
NRC Project Director: John F. Stolz
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: January 30, 1995
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 4.6.1.2.a and associated Bases for
3/4.6.1.2 to state that Type A tests for overall integrated containment
leakage rate shall be conducted in accordance with the requirements
specified in Appendix J of 10 CFR 50, as modified by NRC-approved
exemptions. Additionally, TS 4.6.1.2.b would be revised to eliminate
the reference to the schedule contained in TS 4.6.1.2.a.
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:
Toledo Edison has reviewed the proposed change and determined
that a significant hazards consideration does not exist because
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in
accordance with these changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no accident initiators,
[[Page 14029]]
conditions or assumptions are significantly affected by the
proposed changes.
The proposed change would revise Technical Specification (TS)
Surveillance Requirement (SR) 4.6.1.2.a to allow overall integrated
containment leakage rate (Type A) testing to be scheduled in
accordance with 10 CFR 50 Appendix J, as modified by approved
exemptions, and would make associated administrative changes to TS
SR 4.6.1.2.b and to TS Bases 3/4.6.1.2. As stated above, none of
these proposed changes involve accident initiators, conditions, or
assumptions.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because no accident conditions or
assumptions are affected by the proposed changes.
The results of the previous Type A testing demonstrate a high
degree of containment integrity. The Type B and C testing performed
since the last Type A test provides confidence that the high degree
of containment integrity will be maintained during the interval to
the next Type A test. Therefore, the proposed changes do not alter
the source term, containment isolation, or allowable releases, and
will not increase the radiological consequences of a previously
evaluated accident.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because no new or
different accident initiators or assumptions are introduced by the
proposed changes. The proposed changes do not affect the design or
operation of any plant system, structure, or component. The proposed
changes do not affect any accident initiators and are not initiators
themselves. The proposed changes do not alter any accident
scenarios.
3. Not involve a significant reduction in a margin of safety.
The initial conditions and methodologies used in the accident
analyses remain unchanged. As described above, the proposed changes
do not significantly reduce or adversely affect the confidence that
the present high degree of containment integrity will be maintained.
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 Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: Leif J. Norrholm
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of amendment request: January 30, 1995
Description of amendment request: The proposed amendment would
provide new Reactor Coolant Pressure Boundary (RCPB) pressure-
temperature limit curves that are applicable up to 21 effective full
power years (EFPY).
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:
Toledo Edison had reviewed the proposed change and determined
that a significant hazards consideration does not exist because
operation of Davis-Besse Nuclear Power Station, Unit 1, in
accordance with this change would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because: (1) revision of the pressure-
temperature curves and the extended applicability of the pressurizer
level/RCS pressure limit curves for periods when relief valve DH4849
is inoperable will continue to provide the same level of protection
of the RCPB as was previously evaluated, and (2) the revision to
License Condition 2.C(3)(d) is administrative to reflect the
validity of the present analyses to 21 EFPY and (3) the revision to
the Technical Specification Bases
to reflect the extension to 21 EFPY is administrative and does
not affect any previously analyzed accidents.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because: (1) revision of the pressure-
temperature curves and the extended applicability of the pressurizer
level/RCS pressure limit curves for periods when relief valve DH4849
is inoperable will continue to provide the same level of protection
of the RCPB as was previously evaluated, and (2) the revision to
License Condition 2.C(3)(d) is administrative to reflect the
validity of the present analyses to 21 EFPY and (3) the revision to
the Technical Specification Bases to reflect the extension to 21
EFPY is administrative and does not affect any previously analyzed
accidents.
2. Not create the possibility of a new or different kind of
accident from any accident previously evaluated because: (1)
revision of the pressure-temperature curves and the extended
applicability of the pressurizer level/RCS pressure limit curves
will continue to provide protection against reactor vessel failure
due to brittle fracture concerns under all postulated circumstances,
and (2) the revision to License Condition 2.C(3)(d) is
administrative to reflect the validity of the present analyses to 21
EFPY and (3) the revision to the Technical Specification Bases to
reflect the extension to 21 EFPY is an administrative change and
does not affect any activities or equipment in plant operation.
3. Not involve a significant reduction in a margin of safety
because: (1) revision of the pressure-temperature curves and the
extended applicability of the pressurizer level/RCS pressure limit
curves maintains the present margin of safety from reactor vessel
brittle fracture as required by 10 CFR 50, Appendix G, and (2) the
revision to License Condition 2.C(3)(d) and the Bases revision are
administrative and do not affect any analyses which provide the
basis for the Technical Specifications.
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 Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman,
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
NRC Project Director: Leif J. Norrholm
Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281,
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia
Date of amendment request: February 14, 1995
Description of amendment request: The proposed change revises
Technical Specification 4.4.D to reference the testing requirements of
10 CFR Part 50, Appendix J, and to state that the Nuclear Regulatory
Commission-approved exemptions to the applicable regulatory
requirements are permitted.
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:
Virginia Electric and Power Company has performed an evaluation
of ... the proposed administrative Technical Specification change,
in accordance with 10 CFR 50.91(a)(1) regarding no significant
hazards considerations using the standards in 10 CFR 50.92(c). A
discussion of these standards as they relate to this ... amendment
request follows.
Criterion 1 - Does Not Involve a Significant Increase in the
Probability or Consequences of an Accident Previously Evaluated.
The proposed change ... revises Technical Specification 4.4.D to
reference the testing frequency requirements of 10 CFR 50 Appendix J
and to state that NRC approved exemptions to the applicable
regulatory [[Page 14030]] requirements are permitted. The current
Technical Specification requires retests in accordance with Section
III.D.1(a) of Appendix J. The proposed administrative change simply
includes the statement ``as modified by NRC approved exemptions.''
No new requirements are added, nor are any existing requirements
deleted. Any specific changes to the requirements of Section
III.D.1(a) will require a submittal from Virginia Electric and Power
Company under 10 CFR 50.12 and subsequent review and approval by the
NRC prior to implementation. The proposed change is stated
generically to avoid the need for further Technical Specification
changes if different exemptions are approved in the future.
The proposed change, in itself, does not affect reactor
operations or accident analyses and has no radiological
consequences. The change provides clarification so that future
Technical Specifications changes will not be necessary to correspond
to applicable NRC approved exemptions from the requirements of
Appendix J.
Therefore, this proposed change does not involve a significant
increase in the probability or consequences of any accident
previously evaluated.
Criterion 2 - Does Not Create the Possibility of a New or
Different Kind of Accident from any Previously Evaluated.
The proposed Technical Specification amendment provides
clarification to a specification that paraphrases a codified
requirement.
Since the proposed change would not change the design,
configuration or method of operation of the plant, it would not
create the possibility of a new or different kind of accident from
any previously evaluated.
Criterion 3 - Does Not Involve a Significant Reduction in the
Margin of Safety.
The proposed Technical Specification change is administrative
and clarifies the relationship between the requirements of TS 4.4.D,
Appendix J, and any approved exemptions to Appendix J. It does not,
in itself, change a safety limit or [a] Limiting Condition for
Operation. The NRC will directly approve any proposed change or
exemption to III.D.1(a) of Appendix J prior to implementation.
Therefore, this change does not involve a significant reduction
in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, 951 E. Byrd Street, Richmond, Virginia
23219.
NRC Project Director: David B. Matthews
Previously Published Notices Of Consideration Of Issuance Of
Amendments To Facility Operating Licenses, Proposed No Significant
Hazards Consideration Determination, And Opportunity For A Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of amendment request: February 14, 1995
Brief description of amendment request: The amendment request
proposes changes to Technical Specification 3.8.2, ``AC Sources-
Shutdown;'' 3.8.5, ``DC Sources-Shutdown;'' and 3.8.8, ``Inverters-
Shutdown.'' The proposed changes would revise the operability
requirements for the Division 3 diesel generator and the Division 3 and
4 batteries, battery chargers, and inverters to apply only when the
high pressure core spray system is required to be operable.Date of
publication of individual notice in Federal Register: February 17, 1995
(60 FR 9412).
Expiration date of individual notice: March 20, 1995
Local Public Document Room location: Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: September 8, 1994
Brief description of amendment request: The amendment request
proposes changes to Technical Specification Section 3/4.9.1 to
establish administrative controls to address a possible boron dilution
event directly from the reactor makeup water system.
Date of publication of individual notice in Federal Register: March
1, 1995 (60 FR 11151).
Expiration date of individual notice: March 31, 1995
Local Public Document Room location: Callaway County Public
Library, 710 Court Street, Fulton, Missouri 65251.
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.
Arizona Public Service Company, et al., Docket No. STN 50-529, Palo
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona
Date of application for amendment: November 30, 1994, as
supplemented by letter dated January 27, 1995
Brief description of amendment: The amendment changed the
pressurizer code safety valve lift setting from 2500
[[Page 14031]] psia to 2475 psia. The lift setting is being changed to
permit Unit 2 to operate with up to 1500 plugged tubes in each steam
generator.
Date of issuance: March 1, 1995
Effective date: March 1, 1995
Amendment No.: 78
Facility Operating License No. NPF-74: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
496) The additional information contained in the January 27, 1995,
supplemental letter was clarifying in nature and thus within the scope
of the initial notice and did not affect the NRC staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated March 1, 1995.No significant hazards consideration comments
received: No.
Local Public Document Room location: Phoenix Public Library, 12
East McDowell Road, Phoenix, Arizona 85004
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station,Plymouth County, Massachusetts
Date of application for amendment: September 6, 1994
Brief description of amendment: The proposed amendment relocates
the alarms for the drywell to suppression chamber vacuum breaker to a
different annunicator panel.
Date of issuance: February 16, 1995 Effective date: To be
implemented prior to startup from refueling outage 10.
Amendment No.: 158
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 1994 (59 FR
53839) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated No significant hazards
consideration comments received: No
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457,
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois
Date of application for amendments: January 5, 1994, as
supplemented by letters dated April 26, 1994, September 30, 1994, and
January 12, 1995.
Brief description of amendments: The amendments change the
Braidwood Technical Specifications to remove the requirement to verify,
every 18 months, that the control room ventilation can be manually
isolated.
Date of issuance: February 28, 1995
Effective date: February 28, 1995
Amendment Nos.: 60 and 60
Facility Operating License Nos. NPF-72 and NPF-77: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 25, 1995 (60 FR
4930). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 28, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Wilmington Township Public
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.
Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos.
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2,
Will County, Illinois
Date of application for amendments: August 31, 1993, as
supplemented July 19, 1994.
Brief description of amendments: The amendments revise the
technical specifications by increasing the allowed outage time for an
inoperable chiller only in MODES 1 through 4, adding an optional ACTION
statement in MODES 5 and 6, and adding a surveillance requirement for
the control room ventilation system.
Date of issuance: March 2, 1995
Effective date: March 2, 1995
Amendment Nos.: 70, 70, 61 and 61
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: January 25, 1995 (60 FR
4932). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 2, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: For Byron, the Byron Public
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee
Street, Wilmington, Illinois 60481.Commonwealth Edison Company, Docket
Nos. 50-237 and 50-249, Dresden Nuclear Power Station, Units 2 and 3,
Grundy County, Illinois; Docket Nos. 50-254 and 50-265, Quad Cities
Nuclear Power Station, Units 1 and 2, Rock Island County, IllinoisDate
of application for amendments: July 29, 1992, as supplemented January
14, 1993, and February 16, 1993
Brief description of amendments: Dresden and Quad Cities Technical
Specification Upgrade Program. Date of issuance: February 16,
1995Effective date: Immediately, to be implemented by December 31,
1995.
Amendment Nos.: 131, 125, 152, and 148
Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30.
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34071) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 16, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: For Dresden, The Morris Public
Library, 604 Liberty Street, Morris, Illinois 60450; For Quad Cities,
The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.
Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois
Date of application for amendments: July 29, 1992, as supplemented
January 14, 1993, February 16, 1993 and January 27, 1995
Brief description of amendments: The July 29, 1992, application, is
one of twelve applications which have been submitted by Commonwealth
Edison Company (ComEd) in an effort to upgrade the existing custom
Technical Specifications (TS) to the Boiling Water Reactor (BWR)
Standard Technical Specifications (STS). Dresden has recently
rescheduled the Unit 2 refueling outage from March 4, 1995, until June
1995. Currently, the surveillance frequency for certain Inservice
Testing (IST) requirements expires on February 21, 1995. The current
TSs do not make provisions for a grace period for surveillance
frequencies of the IST program. In accordance with BWR STS guidance,
the TSs regarding IST proposed in the July 29, 1992, application, allow
the flexibility to perform these tests appropriately during refueling
outages (where applicable) by providing a 25 percent extension to IST
surveillance intervals. The January 27, 1995, supplement requested the
staff to review and approve just that portion of the July 29, 1992,
application dealing with the implementation of the IST program in
Section 3.0/4.0 of the proposed TS. [[Page 14032]]
Date of issuance: February 22, 1995Effective date: February 22,
1995
Amendment Nos.: 132 and 126
Facility Operating License Nos. DPR-19 and DPR-25: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 23, 1993 (58 FR
34071) The January 27, 1995, letter did not change the initial proposed
no significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation datedFebruary 22, 1995.No significant hazards consideration
comments received: No
Local Public Document Room location: Morris Public Library, 604
Liberty Street, Morris, Illinois 60450.
Connecticut Yankee Atomic Power Company and Northeast Nuclear
Energy Company, Docket Nos. 50-213 and 50-245, Haddam Neck Plant
and Millstone Nuclear Power Station, Unit 1, Middlesex County and
New London County, Connecticut
Date of application for amendments: October 31, 1994, as
supplemented February 14, 1995.
Brief description of amendments: The amendments renew the existing
license conditions for both plants to implement and maintain Integrated
Implementation Schedule Program Plans (the Program Plans). The Program
Plans provide a methodology to be followed for scheduling plant
modifications and engineering evaluations.
Date of issuance: February 23, 1995
Effective date: February 23, 1995
Amendment Nos.: 183 for Haddam Neck, 80 for Millstone 1
Facility Operating License Nos. DPR-61 and DPR-21. Amendments
revise the Licenses.
Date of initial notice in Federal Register: December 7, 1994 (59 FR
63117)The February 14, 1995, letter provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination. The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated February 23,
1995.No significant hazards consideration comments received: No.
Local Public Document Room locations: Russell Library, 123 Broad
Street, Middletown, CT 06457, for the Haddam Neck Plant, and the
Learning Resource Center, Three Rivers Community-Technical College,
Thames Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for
Millstone Unit 1.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van
Buren County, Michigan
Date of application for amendment: October 5, 1994, as supplemented
February 10, 20, and 22, 1995.
Brief description of amendment: The amendment revises primary
coolant system (PCS) pressure-temperature limits, power-operated relief
valve setting limits, and primary coolant pump starting limits to
accommodate reactor vessel fluence for an additional 4 effective full
power years. The amendment also revises the emergency core cooling
system technical specifications to render two high-pressure safety
injection pumps incapable of injecting into the PCS when the PCS is
below 300 deg.F rather than rendering both inoperable below 260 deg.F.
In addition, it revises the pressurizer heatup to achieve consistency
between design assumptions and technical specifications limits.
Date of issuance: March 2, 1995
Effective date: March 2, 1995
Amendment No.: 163
Facility Operating License No. DPR-20. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
501) The February 10, 20, and 22, 1995, submittals provided
clarifyinginformation which was within the scope of the initial
application and did not affect the staff's initial proposed no
significant hazards consideration findings. The Commission's related
evaluation of the amendment is contained in a Safety Evaluation dated
March 2, 1995.No significant hazards consideration comments received:
No.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
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: January 10, 1994, as
supplemented March 21 and September 15, 1994, and January 5, 1995
Brief description of amendments: The amendments revised Technical
Specification Table 2.2-1 and TS 4.2.5 to allow a change in the method
for measuring reactor coolant system (RCS) flow rate from the
calorimetric heat balance method to a method based on a one-time
calibration of the RCS cold leg elbow differential pressure taps.
Date of issuance: February 17, 1995
Effective date: To be implemented within 30 days from the date of
issuance
Amendment Nos.: 128 and 122
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 26, 1994 (59 FR
3743) for Unit 1; and March 1, 1994 (59 FR 9785) for Unit 2
The March 21 and September 15, 1994, and January 5, 1995, letters
provided additional information that did not change the initial scope
of the January 10, 1994, application and the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated February 17, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear
One,Unit No. 1, Pope County, Arkansas
Date of amendment request: August 30, 1994
Brief description of amendment: The amendment revised the Technical
Specifications to address the installation of two battery chargers on
each 125 vdc power train in lieu of the ``swing'' battery charger that
is currently used.
Date of issuance: February 17, 1995
Effective date: February 17, 1995
Amendment No.: 176
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 17, 1995 (60 FR
3439) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 17, 1995.No significant hazards
consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear
One,Unit No. 1, Pope County, Arkansas
Date of amendment request: June 22, 1994.
Brief description of amendment: The amendment extends the allowable
outage time for one inoperable train of emergency feedwater from 36
hours to 72 hours, clarifies the specifications and their associated
bases, and relocates information within the specifications.
Date of issuance: March 1, 1995
Effective date: 30 days following the date of
issuance. [[Page 14033]]
Amendment No.: 177
Facility Operating License No. DPR-51. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 17, 1994, (59 FR
42339) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 1, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Tomlinson Library, Arkansas
Tech University, Russellville, Arkansas 72801
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: January 19, 1995
Brief description of amendment: The amendment changed the Appendix
A technical specifications (TSs) by adding TS 3.0.5 and its associated
Bases. This new specification will allow equipment removed from service
or declared inoperable to comply with ACTIONS to be returned to service
under administrative controls soley to perform testing required to
demonstrate its OPERABILITY or the OPERABILITY of other equipment.
Date of issuance: March 1, 1995
Effective date: March 1, 1995
Amendment No.: 101
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 27, 1995 (60 FR
5441) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated March 1, 1995.No significant hazards
consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
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 by
letter dated December 2, 1994.
Brief description of amendment: The amendment revised the Technical
Specifications for the Waterford Steam Electric Station, Unit 3, by
modifying the specifications having cycle-specific parameter limits by
replacing the values of those limits with a reference to a core
operating limits report for the values of those limits. These changes
are in accordance with the requirements of Generic Letter 88-16.
Date of issuance: March 1, 1995
Effective date: March 1, 1995
Amendment No.: 102
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65812) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 1, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam
ElectricStation, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: August 19, 1994, as supplemented by
letter dated October 14, 1994.
Brief description of amendment: The amendment changed the Appendix
A technical specification (TSs) by removing the Limiting Condition For
Operation (LCO) 3/4.3.4, the associated surveillance requirements, and
Bases information from the TSs. This information and requirements will
be incorporated into the Waterford 3 Updated Final Safety Analysis
Report (UFSAR) and maintained under the provisions of 10 CFR 50.59.
Date of issuance: March 2, 1995
Effective date: March 2, 1995
Amendment No.: 103
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: August 31, 1994 (59 FR
45023) The additional information contained in the supplemental letter
dated October 14, 1994, was clarifying in nature and thus, within the
scope of the initial notice and did not affect the staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendment is contained in a Safety Evaluation
dated March 2, 1995.No significant hazards consideration comments
received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: April 21, 1993
Brief description of amendment: The amendment revised the
requirement for control rod testing to increase the ``notch testing''
surveillance interval for partially withdrawn control rods from once
per 7 days to once per 31 days. The change is consistent with the
format and content of the Improved Standard Technical Specifications
(NUREG-1434, Revision 0).
Date of issuance: February 16, 1995
Effective date: February 16, 1995
Amendment No: 115
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: May 12, 1993 (58 FR
28055) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 16, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: July 14, 1993
Brief description of amendment: The amendment revised technical
specification requirements for the hydrogen ignition system (HIS). The
amendment also removed several tables related to the HIS in accordance
with guidance contained in Generic Letter 91-08, ``Removal of Component
Lists From Technical Specifications.''
Date of issuance: February 16, 1995
Effective date: February 16, 1995
Amendment No: 116
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46232) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 16, 1995. No
significant hazards consideration comments received: No [[Page 14034]]
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce at Washington, Natchez, Mississippi 39120.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: August 11, 1993
Brief description of amendment: The amendment deleted the
requirements of Limiting Condition for Operation (LCO) 3.3.3.9 and
Surveillance Requirement 4.3.3.9 related to loose-part detection
instrumentation. The deleted requirements will be relocated to
documents that are controlled by the licensee under the provisions of
10 CFR 50.59. The change is consistent with the format and content of
the Improved Standard Technical Specifications (NUREG-1434, Revision
0).
Date of issuance: February 16, 1995
Effective date: February 16, 1995
Amendment No: 117
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46232) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 16, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: August 11, 1993
Brief description of amendment: The amendment deleted certain
accident monitoring instruments from Technical Specification Table
3.3.7.5-1 ``Accident Monitoring Instrumentation'' and deleted the
corresponding Surveillance Requirements from Table 4.3.7.5-1,
``Accident Monitoring Instrumentation Surveillance Requirements.'' The
deleted requirements will be relocated to documents that are controlled
by the licensee under the provisions of 10 CFR 50.59. The change is
consistent with the format and content of the Improved Standard
Technical Specifications (NUREG-1434, Revision 0).
Date of issuance: February 16, 1995
Effective date: February 16, 1995
Amendment No: 118
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 1, 1993 (58
FR 46234) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 16, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Mississippi Power &
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit
1, Claiborne County, Mississippi
Date of application for amendment: October 22, 1993, as
supplemented by letters dated February 10, and 14, 1995.
Brief description of amendment: The amendment modified the testing
frequencies for the drywell bypass test and the airlock test, relocated
certain drywell airlock tests from the technical specifications to
administrative procedures, and incorporates various improvements from
the Improved Standard Technical Specifications (NUREG-1434, Revision
0).
Date of issuance: February 16, 1995
Effective date: February 16, 1995
Amendment No: 119
Facility Operating License No. NPF-29. Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64607) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 16, 1995. No
significant hazards consideration comments received: No
Local Public Document Room location: Judge George W. Armstrong
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.
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: May 17, 1993, as supplemented on
December 23, 1994
Brief description of amendment: The amendment changes the action
statement for inoperable degraded grid and loss of voltage relays and
their associated auxiliary relays and timers.
Date of issuance: January 31, 1995
Effective date: January 31, 1995
Amendment No.: 193
Facility Operating License No. DPR-50. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 10, 1993 (58
FR 59750). The December 23, 1994, letter provided additional
information that did not change the initial proposed no significant
hazards consideration determination. The Commission's related
evaluation of this amendment is contained in a Safety Evaluation dated
January 31, 1995. 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. The above Notice was to be
published in the Federal Register of February 15, 1995. The notice that
was inadvertently published at 60 FR 8762 relates to a licensing action
which has not been completed.
Houston Lighting & Power Company, City Public Service Board of San
Antonio, Central Power and Light Company, City of Austin, Texas,
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2,
Matagorda County, Texas
Date of amendment request: November 7, 1994, as supplemented by
letters dated December 20, 1994, and January 23, 1995.
Brief description of amendments: The amendments changed the number
of standby diesel generators (SDGs) (emergency power source) required
to be operable during Mode 6 with greater than or equal to 23 feet of
water above the reactor vessel flange, from two to one. The amendment
also allows limited substitution of an alternate onsite emergency power
source for one of the two required SDGs, in Mode 5, and in Mode 6 with
less than 23 feet of water. In addition, certain system specifications
that are affected by the changes for the emergency power source were
also changed.
Date of issuance: February 14, 1995
Effective date: February 14, 1995, to be implemented within 31 days
of issuance.
Amendment Nos.: Unit 1 - Amendment No. 34; Unit 2 - Amendment No.
20
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical [[Page 14035]] Specifications.Public comments
requested as to proposed no significant hazards consideration: Yes (60
FR 5739, dated January 30, 1995). The notice provided an opportunity to
submit comments on the Commission's proposed no significant hazards
consideration determination. No comments have been received. The notice
also provided for an opportunity to request a hearing by March 1, 1995,
but stated that, if the Commission makes a final no significant hazards
consideration determination, any such hearing would take place after
issuance of the amendments.
The Commission's related evaluation of the amendments, finding of
exigent circumstances, and final determination of no significant
hazards consideration is contained in a Safety Evaluation dated
February 14, 1995.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488
IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center,
Linn County, Iowa
Date of application for amendment: August 15, 1994, as supplemented
on December 21, 1994, and January 20, 1995. The licensee's submittals
of December 21, 1994, and January 20, 1995, provided clarification and
did not change the original no significant hazards consideration.
Brief description of amendment: The proposed amendment would revise
the Technical Specifications by increasing the allowable main steam
isolation valve (MSIV) leakage and deleting the requirements applicable
to the MSIV leakage control system.
Date of issuance: February 22, 1995
Effective date: February 22, 1995 and to be implemented within 90
days.
Amendment No.: 207
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47169) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 22, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, S. E., Cedar Rapids, Iowa 52401.
Illinois Power Company and Soyland Power Cooperative, Inc., Docket
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County,
Illinois
Date of application for amendment: August 12, 1994, as supplemented
on October 14, 1994 and February 6, 1995.
Brief description of amendment: The amendment modifies Clinton
Power Station Technical Specification 3.6.5.1, ``Drywell,'' to permit a
one-time only change to forego performance of the drywell bypass
leakage rate test during the fifth refueling outage scheduled to begin
in March 1995.
Date of issuance: March 1, 1995
Effective date: March 1, 1995
Amendment No.: 96
Facility Operating License No. NPF-62. The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49427). The October 14, 1994, and February 6, 1995, submittals
consisted of revisions and clarifications which did not change the
staff's initial proposed no significant hazards consideration
determination or expand the scope of the original notice.The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated March 1, 1995. No significant hazards
consideration comments received: No
Local Public Document Room location: The Vespasian Warner Public
Library, 120 West Johnson Street, Clinton, Illinois 61727.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316,
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County,
Michigan
Date of application for amendments: November 18, 1994
Brief description of amendments: The amendments revise Technical
Specification 4.0.5 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 allows the licensee to
implement certain 10 CFR 50.55a relief requests while the relief
requests are being reviewed by the NRC at the beginning of an updated
interval.
Date of issuance: February 23, 1995
Effective date: February 23, 1995
Amendment Nos.: 190/176
Facility Operating License Nos. DPR-58 and DPR-74. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65817) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 23, 1995. No
significant hazards consideration comments received: No.
Local Public Document Room location: Maud Preston Palenske Memorial
Library, 500 Market Street, St. Joseph, Michigan 49085.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
MillstoneNuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of application for amendment: May 18, 1994
Brief description of amendment: The amendment modifies the
operability requirements for the fuel building exhaust filter system.
The amendment will result in modifications to the applicability,
surveillance requirement, and bases sections of Technical Specification
3/4.9.12, ``Fuel Building Exhaust Filter System.''
Date of issuance: February 22, 1995
Effective date: As of the date of issuance to be implemented
within30 days.
Amendment No.: 105
Facility Operating License No. NPF-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: June 22, 1994 (59 FR
32234) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 22, 1995.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.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323,
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis
Obispo County,California
Date of application for amendments: July 9, 1992
Brief description of amendments: The amendments extend the
operating licenses for the Diablo Canyon Nuclear Power Plant, Units 1
and 2 to recover or recapture the construction period of the reactors.
Specifically, the amendments extend the expiration date of the Unit 1
license from April 23, 2008, to September 22, 2021, and the expiration
date of the Unit 2 license from December 9, 2010, to April 26, 2025.
Date of issuance: March 1, 1995
Effective date: March 1, 1995
Amendment Nos.: 97 and 96
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the license.
Date of initial notice in Federal Register: July 22, 1992 (57 FR
32575) [[Page 14036]] The Commission's related evaluation of the
amendments is contained in a Safety Evaluation dated March 1, 1995.No
significant hazards consideration comments received: Yes. Comments from
the San Luis Obispo Mothers for Peace (MFP) and their contentions were
admitted into this proceeding. These contentions concern the adequacy
of the licensee's maintenance and surveillance program and interim
corrective actions in lieu of Thermo-Lag. The Atomic Safety and
Licensing Board, in its initial decision dated November 4, 1994 (LBP-
94-35), authorized the staff to extend the DCPP operating license
expiration dates. Because a hearing was held prior to license issuance,
the staff does not need to make a final no significant hazards
consideration determination.
Local Public Document Room location: California Polytechnic State
University, Robert E. Kennedy Library, Government Documents and Maps
Department, San Luis Obispo, California 93407
Pennsylvania Power and Light Company, Docket No. 50-387,
Susquehanna Steam Electric Station, Unit 1, Luzerne County,
Pennsylvania
Date of application for amendment: July 27, 1994, as supplemented
October 27, 1994 and February 3, 1995
Brief description of amendment: The amendment raises the authorized
Power Level from 3293 MWt to a new limit of 3441 MWt.
Date of issuance: February 22, 1995
Effective date: As of date of issuance and is to be implemented
prior to startup in Cycle 9, currently scheduled to occur in May 1995.
Amendment No.: 143
Facility Operating License No. NPF-14: This amendment revised the
Technical Specifications and license.
Date of initial notice in Federal Register: September 14, 1994 (59
FR 47171) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 22, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: June 23, 1994
Brief description of amendments: The amendment revises Technical
Specification 4.0.5, which provides the requirements for inservice
inspection and testing of ASME Code components, to conform to Standard
Technical Specifications (NUREG-1433).
Date of issuance: February 28, 1995
Effective date: February 28, 1995
Amendment Nos.: 144 and 113
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 3, 1994 (59 FR
39595) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 28, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of application for amendments: October 28, 1994, and
supplemented by letter dated December 29, 1994
Brief description of amendments: These amendments change the
Technical Specifications (TS) for the two units by adding 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 additions
reflect changes to the methodology that the licensee is using to
perform its nuclear fuel reload analysis for the two units.
Date of issuance: February 28, 1995
Effective date: February 28, 1995
Amendment Nos.: 145 and 114
Facility Operating License Nos. NPF-14 and NPF-22. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65819) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated February 28, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353,
Limerick Generating Station, Units 1 and 2, Montgomery County,
Pennsylvania
Date of application for amendments: August 31, 1994
Brief description of amendments: These amendments address Section
5, ``Remove Temperature Requirement for Operational Condition 5 (TSCR
94-44-0), by revising TS Table 1.2 and TS Bases 3/4.9.11 to remove the
average reactor coolant temperature requirement in Operational
Condition (OPCON) 5, Refueling.
Date of issuance: January 27, 1995
Effective date: January 27, 1995Amendment Nos. 88 and 50
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55884) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated January 27, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of application for amendment: November 16, 1994
Brief description of amendment: The amendment revises Technical
Specifications Section 3.10.8 and the associated Bases, to reduce the
maximum allowable control rod drop time from 2.4 to 1.8 seconds.
Date of issuance: February 21, 1995
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 160
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 20, 1995 (60 FR
4203) The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated February 21, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610. [[Page 14037]]
Saxton Nuclear Experimental Corporation, Docket No. 50-146, Saxton
Nuclear Reactor Facility
Date of application for amendment: August 8, 1994, as supplemented
on October 28, 1994, and January 12, 1995.
Brief description of amendment: The amendment adds characterization
as an authorized activity at Saxton and improves the wording of the
technical specifications.
Date of issuance: February 22, 1995
Effective date: February 22, 1995
Amendment No.: 12Amended Facility License No. DPR-4: Amendment
changed the Technical Specifications
Date of initial notice in Federal Register: November 9, 1994. The
Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated February 22, 1995.No significant hazards
consideration comments received: No
Local Public Document Room location: Saxton Community Library, 911
Church Street, Saxton, Pennsylvania 16678
Southern California Edison Company, et al., Docket Nos. 50-361 and
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3,
San Diego County, California
Date of application for amendments: December 30, 1993, as
supplemented by letters dated June 3, 1994, August 25, 1994, and
January 3, 19, and 30, 1995.
Brief description of amendments: These amendments revise TS 3.9.4,
``Containment Building Penetrations,'' and the associated bases to
allow both doors of the containment personnel airlock to be open at the
same time during refueling operations provided certain conditions are
met.
Date of issuance: February 28, 1995
Effective date: February 28, 1995
Amendment Nos.: 117 and 106
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49434). The additional information contained in the January 3, 19,
and 30, 1995, letters were clarifying in nature, within the scope of
the initial notice and did not affect the NRC staff's proposed no
significant hazards consideration determination. The Commission's
related evaluation of the amendments is contained in a Safety
Evaluation dated February 28, 1995.No significant hazards consideration
comments received: No.
Local Public Document Room location: Main Library, University of
California, P. O. Box 19557, Irvine, California 92713
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston
County, Alabama.
Date of amendments request: December 19, 1994
Brief description of amendments: The amendments to Technical
Specifications include: (1) a revision in Table 3.7-3 to the main steam
safety valve (MSSV) setpoint tolerance from plus or minus 1 percent to
plus or minus 3 percent, (2) modification of the bases to 3/4.7.1.1 to
increase the relieving capacity of the MSSVs to at least 12,984,660
pounds per hour which corresponds to approximately 112 percent of total
secondary steam flow at 100 percent rated thermal power, (3)
modifications to Table 3.7-1 to reduce the allowable power range
neutron flux high setpoints for multiple inoperable steam generator
safety valves, and (4) an editorial correction to Bases 3/4.7.1.2 to
indicate required auxiliary feedwater flow at ``1133 psia'' rather than
``1133 psig.''
Date of issuance March 1, 1995
Effective date: March 1, 1995
Amendment Nos.: 112 and 103
Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: January 4, 1995 (60 FR
505) The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated March 1, 1995No significant hazards
consideration comments received: No
Local Public Document Room location: Houston-Love Memorial Library,
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302The
Commission's related evaluation of the amendments is contained in a
Safety Evaluation dated March 1, 1995No significant hazards
consideration comments received: No
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: September 29, 1993
Brief description of amendment: The proposed changes increase the
amount of boron required in the standby liquid control system.
Date of issuance: February 28, 1995
Effective date: February 28, 1995
Amendment Nos.: 217, 233 and 191
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 8, 1994 (59 FR
29635) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 28, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
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: September 30, 1993 (TS 336)
Brief description of amendment: The proposed changes revise and
clarify the spent fuel pool water level, temperature, sampling, and
analysis surveillance requirements.
Date of issuance: March 2, 1995
Effective date: March 2, 1995
Amendment Nos.: 218, 334 and 192
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67862) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated March 2, 1995.No significant
hazards consideration comments received: None
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
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 31, 1994
Brief description of amendment: For Browns Ferry Units 1 and 3, the
proposed changes provide for operation in the extended load line limit
region and revised rod block monitor operability requirements. For all
three Browns Ferry units, the changes delete a obsolete value for rated
loop recirculation flow rate, relocate cycle-specific equations to the
Core Operating Limits report, and provide other miscellaneous changes.
Date of issuance: February 24, 1995
Effective date: February 24, 1995
Amendment Nos.: 216, 232, 190
Facility Operating License Nos. DPR-33, DPR-52 and DPR-68:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: September 28, 1994 (59
FR 49437) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 24, 1995.No significant
hazards consideration comments received: None [[Page 14038]]
Local Public Document Room location: Athens Public library, South
Street, Athens, Alabama 35611
Toledo Edison Company, Centerior Service Company, and The Cleveland
Electric Illuminating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio
Date of application for amendment: October 7, 1994
Brief description of amendment: Eliminates redundancy in system
leakage test requirements by revising TS 3/4.5.2 and its associated
basis for the Emergency Core Cooling System and TS 3/4.6.2 and its
associated basis for the Containment Spray System.
Date of issuance: February 27, 1995
Effective date: February 27, 1995 and to be implemented within 90
days.
Amendment No. 195
Facility Operating License No. NPF-3. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: November 9, 1994 (59 FR
55893) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 27, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: University of Toledo Library,
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: November 18, 1994 (published in Federal
Register as November 11, 1994)
Brief description of amendments: The proposed amendments would
provide for cycle-specific allowances to account for increases in the
Heat Flux Hot Channel Factor between monthly surveillances.
Date of issuance: March 1, 1995
Effective date: March 1, 1995, to be implemented within 30 days of
issuance.
Amendment Nos.: Unit 1 - Amendment No. 34; Unit 2 - Amendment No.
20
Facility Operating License Nos. NPF-87 and NPF-89. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 7, 1994 (59 FR
63127) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated March 1, 1995.No significant
hazards consideration comments received: No.
Local Public Document Room location: University of Texas at
Arlington Library, Government Publications/Maps, 702 College, P.O. Box
19497, Arlington, Texas 76019.
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 amendment modifies the Techical
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 than 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.
Date of issuance: February 27, 1995
Effective date: February 27, 1995
Amendment No.: 134
Facility Operating License No. NPF-21: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 21, 1994 (59
FR 65828) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 27, 1995.No significant
hazards consideration comments received: No
Local Public Document Room location: Richland Public Library, 955
Northgate Street, Richland, Washington 99352
Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee
NuclearPower Plant, Kewaunee County, Wisconsin
Date of application for amendment: February 23, 1994
Brief description of amendment: The amendment revises Kewaunee
Nuclear Power Plant (KNPP) Technical Specification (TS) 6.8.c by
removing the requirement to conduct a biennial review of plant
procedures in accordance with American National Standards Institute
(ANSI) N18.7-1976, Section 5.2.15. Alternate programs that are
described in the KNPP Operational Quality Assurance Program Description
(OQAPD) will be used to ensure that procedures are reviewed and
maintained current.
Date of issuance: February 23, 1995
Effective date: February 23, 1995 and to be implemented within 30
days.
Amendment No.: 115
Facility Operating License No. DPR-43. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (59 FR
14903) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated February 23, 1995.No significant
hazards consideration comments received: None.
Local Public Document Room location: University of Wisconsin
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin
54301.
Notice Of Issuance Of Amendments To Facility Operating Licenses And
Final Determination Of No Significant Hazards Consideration And
Opportunity For A Hearing (Exigent Public Announcement Or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the standards and requirements of the Atomic
Energy Act of 1954, as amended (the Act), and the Commission's rules
and regulations. The Commission has made appropriate findings as
required by the Act and the Commission's rules and regulations in 10
CFR Chapter I, which are set forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual 30-day Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for [[Page 14039]] example, in derating or shutdown of a
nuclear power plant or in prevention of either resumption of operation
or of increase in power output up to the plant's licensed power level,
the Commission may not have had an opportunity to provide for public
comment on its no significant hazards consideration determination. In
such case, the license amendment has been issued without opportunity
for comment. If there has been some time for public comment but less
than 30 days, the Commission may provide an opportunity for public
comment. If comments have been requested, it is so stated. In either
event, the State has been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at
the local public document room for the particular facility involved.
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. By April 14, 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. Since the Commission has made a final determination
that the amendment involves no significant hazards consideration, if a
hearing is requested, it will not stay the effectiveness of the
amendment. Any hearing held would take place while the amendment is in
effect.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above
date. Where petitions are filed during the last 10 days of the notice
period, it is requested that the petitioner promptly so inform the
Commission by a toll-free telephone call to Western Union at 1-(800)
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator
should be given Datagram Identification Number N1023 and the following
message addressed to (Project Director): petitioner's name and
telephone number, date petition was mailed, plant name, and publication
date and page number of this Federal Register notice. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear [[Page 14040]] 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 the
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
Pennsylvania Power and Light Company, Docket No. 50-388,
SusquehannaSteam Electric Station, Unit 2, Luzerne County,
Pennsylvania
Date of application for amendment: February 7, 1995
Brief description of amendment: The amendment changed the Technical
Specifications to allow continued operation with one neutron flux
monitor system channel (B'' channel) inoperable and should
the remaining channel become inoperable to allow continued plant
operation for 7 days to restore one of the two inoperable channels.
Date of issuance: March 1, 1995
Effective date: March 1, 1995
Amendment No.: 115
Facility Operating License No. NPF-22: Amendment revised the
Technical Specifications. Public comments requested as to proposed no
significant hazards consideration: No. On February 8, 1995, the staff
issued a Notice of Enforcement Discretion, which was immediately
effective and remained in effect until this amendment was issued.
The Commission's related evaluation of the amendments, finding of
emergency circumstances, consultation with the Commonwealth of
Pennsylvania and final no significant hazards considerations
determination are contained in a Safety Evaluation dated March 1, 1995.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts &
Trowbridge 2300 N Street NW., Washington, D.C. 20037
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18071.
NRC Project Director: John F. Stolz
Dated at Rockville, Maryland, this 8th day of March 1995.
For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of
Nuclear Reactor Regulation
[Doc. 95-6207 Filed 3-14-95; 8:45 am]
BILLING CODE 7590-01-F
