[Federal Register Volume 59, Number 100 (Wednesday, May 25, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-12614]
[[Page Unknown]]
[Federal Register: May 25, 1994]
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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 May 2, 1994, through May 13, 1994. The last
biweekly notice was published on May 12, 1994 (59 FR 24745).
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, 11555 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
20555. The filing of requests for a hearing and petitions for leave to
intervene is discussed below.
By June 24, 1994, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714 which is
available at the Commission's Public Document Room, the Gelman
Building, 2120 L Street, NW., Washington, DC 20555 and at the local
public document room for the particular facility involved. If a request
for a hearing or petition for leave to intervene is filed by the above
date, the Commission or an Atomic Safety and Licensing Board,
designated by the Commission or by the Chairman of the Atomic Safety
and Licensing Board Panel, will rule on the request and/or petition;
and the Secretary or the designated Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555, Attention: Docketing and Services
Branch, or may be delivered to the Commission's Public Document Room,
the Gelman Building, 2120 L Street, NW., Washington DC 20555, 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 20555, and at the local public document
room for the particular facility involved.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of amendments request: March 25, 1994.
Description of amendments request: The proposed amendment would
make the following administrative changes to the Technical
Specifications.
Brunswick Unit 1
1. Bases Section 2.2.1: Remove references to the Rod Sequence
Control System (RSCS) in item 2 on page B 2-4.
2. Bases Section 2.2.1: Correct typographical error in acronym
for hydrogen water chemistry in item 6 on page B 2-6.
3. TS 3.1.4.1: Correct typographical errors in action d,
misspelling of preset, and action d.1, misspelling of BPWS acronym,
on page 3/4 1-14.
4. TS Table 4.3.4-1: Remove references to the RSCS in item g of
the Notes on page 3/4 3-52.
5. TS Table 3.3.5.5-1 Label each item to permit identification
consistent with the scheduling system used for surveillance testing
on pages 3/4 3-64a.
6. TS Table 4.3.5.5-1 Label each item to permit identification
consistent with the scheduling system used for surveillance testing
on page 3/4 3-64c.
7. TS 4.3.6.1.1: Correct typographical error that references
Non-existent Table 4.3.6.1.1-1 to provide correct reference of Table
4.3.6.1-1 on page 3/4 3-88.
8. TS 3.4.2: Correct typographical error indicating extraneous
second footnote on page 3/4 4-4.
Brunswick Unit 2
1. TS Table 2.2.1-1: Correct typographical error in item 2.b
under allowable values by changing 115% to 115.5% on page 2-4.
2. Bases Section 2.2.1: Remove references to the Rod Sequence
Control System (RSCS) in item 2 on page B 2-4.
3. Bases Section 2.2.1: Remove references to the Rod Sequence
Control System in item 10 and revise bases description of the Select
of the Select Rod Insertion consistent with removal of the RSCS on
pages.
4. TS 3.1.4.1: Correct typographical error in action d.1 to
correct misspelling of BPWS acronym on page 3/4 1-14.
5. TS Table 4.3.1-1: Correct grammatical omission of the word
``is'' in item e of the Notes on page 3/4 3-9.
6. TS Table 4.3.1-1: Remove references to the RSCS in item g of
the Notes on page 3/4 3-52.
7. TS Table 3.3.5.5-1: Label each item to permit identification
consistent with the scheduling system used for surveillance testing
on page 3/4 3-64a.
8. TS Table 4.3.5.5-1: Label each item to permit identification
consistent with the scheduling system used for surveillance testing
on page 3/4 3-64c.
9. TS 3.3.6.2: Eliminate footnote, revise applicability
statement and correct typographical errors in actions d and e that
references non-existent Specification on page 3/4 3-93.
10. Base Section 3/4.1.4: Correct identification of Reference
cited to reference 6 on page B 3/4 1-4.
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 because the proposed change [sic] is
administrative in nature. These changes do not alter the
configuration or operation of the facility. The Limiting Safety
Systems Settings and Safety Limits specified in the current
Technical Specifications remain unchanged.
2. The proposed amendment does not create the possibility of a
new or different kind of accident from any accident previously
evaluated. The safety analysis of the facility remains complete and
accurate. There are no physical changes to the facility and the
plant conditions for which the design basis accidents have been
evaluated are still valid. The operating procedure and emergency
procedures are unaffected with the possible exception of resolving
special notations that may have recognized the typographical errors
that are being corrected.
3. The margins of safety are established through the Limiting
Conditions of Operation, Limiting Safety Systems Settings and Safety
Limits specified in the Technical Specifications. Since there are no
changes to the physical design or operation of the facility, these
margins will not be changed.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Attorney for licensee: R. E. Jones, General Counsel, Carolina Power
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
NRC Project Director: William H. Bateman
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: January 28, 1994.
Description of amendment request: The proposed amendment will
remove an exception for the purge and vent valves from surveillance
requirement (SR) 4.6.1.2.d and remove SR 4.6.1.2.f.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve an [significant hazards
consideration] SHC because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change modifies SR 4.6.1.2.d. Currently this SR
indicates the purge supply and exhaust valves have an exception from
the 10CFR50 Appendix J, Type B and C tests. The proposed technical
specification change is consistent with current surveillance
procedures and the [Final Safety Analysis Report] FSAR. The second
proposed change, which removes SR 4.6.1.2.f, reflects current
containment leakage surveillance requirements. The present location
of SR 4.6.1.2.f could imply that containment leakage surveillance
requirements are met by performing SR 4.9.9. However, SR 4.9.9 is
applicable only during core alterations or movement of irradiated
fuel and not during the modes when Technical Specification 3.6.1.2
is applicable. These changes have no effect on actual Appendix J
testing of valves or the current plant accident analysis. Therefore,
the proposed changes cannot increase the probability or consequences
of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes do not introduce any new failure modes. The
plant will continue to operate as designed and there will be no
change to the testing of valves. The proposed changes will not
modify the plant response to the point where it can be considered a
new accident. Therefore, the proposed changes will not create the
possibility of a new or different kind of accident form any
previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes modify SR 4.6.1.2.d which, as presently
written, indicates that the purge supply and exhaust valves are an
exception to the 10CFR50 Appendix J, Type B and C test and
therefore, no exception is required. This is supported by current
surveillance procedures which include the purge supply and exhaust
valves as part of the Type B and C tests. In addition, the proposed
changes are consistent with the FSAR. FSAR Table 7.3-1 ``Containment
Penetrations,'' lists the purge supply and exhaust valves as
required to receive Type B and C tests. Therefore, these proposed
changes revise SR 4.6.1.2.d to reflect actual surveillance
procedures and offer no revisions or reductions to current
surveillance testing. Therefore, these changes will not result in a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Russell Library, 123 Broad
Street, Middletown, Connecticut 06457.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, Counselors at Law, City Place, Hartford, Connecticut 06103-
3499.
NRC Project Director: John F. Stolz.
Consolidated Edison Company of New York, Docket No. 50-247, Indian
Point Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: April 13, 1994
Description of amendment request: The proposed amendment request
would revise the Technical Specifications to amend Sections 3.1.F and
4.13 to allow the repair of steam generator tubes by sleeving as an
alternative to plugging. Additionally, a new tube acceptance criteria,
F*, is proposed which would allow tubes that are degraded in a location
not affecting structural integrity of the tube to remain in service.
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 the requirements of 10 CFR 50.92, the
proposed Technical Specification change is deemed to involve no
significant hazards considerations because operation of Indian Point
Unit No. 2 would not:
(1) Involve a significant increase in the probability or
consequences of an accident previously evaluated since the integrity
of the steam generator tubes after sleeving will be equivalent to
that of the original tubes. The sleeve, sleeve joint, and F* joint
have been analyzed and tested for design, operating, and faulted
condition loadings in accordance with NRC Regulatory [G]uide 1.121
safety factors. The potential for a tube rupture is not increased
with sleeving or F*. At worst case, a tube leak would occur,
resulting in a small primary to secondary leak. Primary to secondary
leakage occurring from within the sleeved or F* portions of the tube
is bounded by the steam generator tube rupture scenario evaluated in
the Final Safety Analysis Report. In addition, the steam generator
tube remains capable of performing its required heat transfer
function. Placing a sleeve in the steam generator tube or leaving a
tube in service with a defect in a portion of the tube that provides
no function results in a more efficient steam generator than
plugging an affected tube. Thus, the consequences of any accident
previously evaluated are not increased because the structural
integrity and the heat transfer capability of the steam generators
are not significantly altered by the proposed change.
(2) Create the possibility of a new or different kind of
accident from any accident previously evaluated because both the
structural integrity and the heat transfer capability of the steam
generators will not be significantly affected by the use of either
of the sleeving processes or the implementation of the F* criteria.
Testing and previous experience indicate that any primary to
secondary leakage would be well below technical specification
limits. In addition, in the unlikely event the defective tube failed
completely at the defect, the remaining sleeve end or F* joint would
restrain tube movement due to the sleeve end geometry or length of
expanded contact within the tubesheet bore. Therefore, there is no
threat to adjacent tubes and no other plant systems will be affected
by this change. Thus, there is no potential for a new or different
kind of accident.
(3) Involve a significant reduction in a margin of safety. The
heat transfer capabilities of Indian Point 2 Steam Generators will
be improved by utilizing the proposed sleeving process or
implementing the F* criteria rather than the currently required tube
plugging and subsequent loss of heat transfer area. The proposed
change will allow a repaired (sleeved) tube or a tube with a tube
end defect below the F* distance to remain in service, rather than
completely blocking the tube's flow with plugs. Because the
structural integrity of the tubes will be unaltered, the net effect
of implementing the proposed change, rather than the currently
required plugging procedure, will be an increase in the heat
transfer characteristics of the steam generator. Westinghouse has
done an evaluation of selected LOCA [loss-of-coolant accident] and
non-LOCA transients to verify that use of sleeves resulting in a
plugging equivalency at the current plant limit will not have an
adverse affect on the thermal-hydraulic performance of the plant.
Therefore, the margin of safety is not reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place,
New York, New York 10003.
NRC Project Director: Robert A. Capra.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: November 15, 1991, as supplemented
February 22, March 11, and April 7, 1994.
Description of amendment request: The amendment request, as
submitted November 15, 1991, proposed completely rewritten requirements
for the instrumentation and control (I&C) sections of the Palisades
Technical Specifications (TS) and was initially noticed in the Federal
Register October 28, 1992 (57 FR 48819). Since that time the licensee
has updated its submittal, providing (1) changes to pages affected by
intervening amendments, (2) clarifications suggested by NRC and
Palisades reviewers, (3) addition of two instrument channels to the
accident monitoring instruments Limiting Condition for Operation (LCO),
(4) deletion of surveillance requirements for safety injection tank
(SIT) instruments, as suggested by Generic Letter (GL) 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance
Requirements for Testing During Power Operation,'' and (5) addition of
a general ``Applicability'' LCO which appears in the Standard TS but
not in the Palisades TS. Changes (4) and (5) were not addressed in the
initial proposed no significant hazards consideration (NSH)
determination. The licensee's NSH analysis for these two changes was
provided in its April 7, 1994, letter to the NRC and is discussed
below.
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:
Consumers Power Company finds that activities associated with
the February 22, 1994 and March 11, 1994 Instrument and Control
Technical Specification change revisions include no significant
hazards; and accordingly, a no significant hazards determination in
accordance with 10CFR50.92(c) is justified. The following summary
supports the finding that the proposed change would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
Neither the deletion of instrument surveillance requirements for
the Safety Injection Tank (SIT) instrumentation nor the addition of
allowance of temporarily returning inoperable equipment to service
for maintenance or testing would affect the probability or
consequences of an accident.
The SIT instrument channels themselves have no accident
function. Their only purpose is to allow verification that the SITs
themselves are operable. Surveillance requirements for these
instruments were purposely deleted from STS during the Technical
Specification Improvement Program. Their removal from Technical
Specifications was suggested in GL 93-05.
Returning inoperable equipment to service as allowed by LCO
3.0.5 is necessary if failed channels are to be restored to operable
status. The restoration of such channels enhances the ability to
monitor for and mitigate abnormal operating conditions and
accidents.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed changes would not alter the operating conditions of
the plant systems, and would not reduce the reliability of any plant
safety equipment.
Therefore, this change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes would not affect the setpoints, capacities,
or operating limits for any equipment. Therefore, the proposed
changes do not involve a significant reduction of 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, as revised, involves no significant hazards
consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Ledyard B. Marsh.
Consumers Power Company, Docket No. 50-255, Palisades Plant, Van Buren
County, Michigan
Date of amendment request: April 7, 1994.
Description of amendment request: The proposed amendment would
change certain Technical Specifications (TS) to relocate fuel cycle-
specific parameter limits that can generally change with core reloads
to a Core Operating Limits Report (COLR) in accordance with the
guidance of Generic Letter 88-16, ``Removal of Cycle-Specific Parameter
Limits from Technical Specifications.'' Several of the TS bases would
also be revised to refer to limits relocated to the COLR. In each case
where TS limits would be relocated to the COLR, the limits placed in
the COLR would be unchanged and the appropriate bases would be revised
accordingly.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The following evaluation supports the finding that operation of the
facility in accordance with the proposed TS would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to the TS simply move the values and
parameters for fuel cycle-specific limits from the TS to a Core
Operating Limits Report (COLR). The requirements to maintain the
plant within appropriate bounds are retained in the TS. The values
of the cycle-specific parameter limits in the COLR are determined
using an NRC-approved methodology and remain consistent with all
applicable limits of the plant safety analyses that are addressed in
the Final Safety Analysis Report (FSAR). A requirements for the COLR
and identification of the approved methodology documents are added
to the TS. There are no associated changes in plant operation.
Therefore, operation of the facility in accordance with the proposed
TS would not result in a significant increase in the probability or
consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
As discussed above, the proposed changes do not remove or
alleviate any requirements to maintain the plant within the
appropriate bounds. There are no associated changes in plant
operation. Therefore, operation of the facility in accordance with
the proposed TS would not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Involve a significant reduction in a margin of safety.
The proposed changes to the TS simply move the values and
parameters for cycle-specific limits from the Specifications to a
Core Operating Limits Report (COLR). The requirements to maintain
the plant within appropriate bounds are retained in the TS. The
values of the cycle-specific parameter limits in the COLR are
determined using an NRC-approved methodology and remain consistent
with all applicable limits of the plant safety analyses that are
addressed in the Final Safety Analysis Report (FSAR). A requirement
for the COLR and identification of the approved methodology
documents are added to the TS. There are no associated changes in
plant operation. Therefore, operation of the facility in accordance
with the proposed TS 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 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Van Wylen Library, Hope
College, Holland, Michigan 49423.
Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
NRC Project Director: Ledyard B. Marsh.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: March 29, 1994, as corrected April 26,
1994.
Date of amendment request: March 29, 1994, as corrected April 26,
1994.
Description of amendment request: The proposed amendment would
modify the surveillance requirements for scram discharge volume vent
and drain valves and isolation actuation instrumentation and modify the
required actions and surveillance requirements for the emergency diesel
generators to reduce testing during power operation. These changes are
in accordance with guidance contained in Generic Letter (GL) 93-05
``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation,'' dated
September 27, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes to the frequency of testing for these
components will reduce the probability of failure due to wear and
eliminate the possibility of initiating transients during testing of
these components. Therefore, the proposed changes will result in a
decrease in the probability of previously evaluated accidents.
Further, the proposed changes do not alter the design, function, or
operation of the components involved and therefore, do not affect
the consequences of any previously evaluated accident.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. As stated above, the proposed changes do not alter the
design, function, or operation of the components involved and
therefore, no new accident scenarios are created.
3. The proposed changes do not involve a significant reduction
in a margin of safety. As developed in Reference 3 [NUREG-1366,
``Improvement to Technical Specification Surveillance
Requirements,'' dated December 1992] and endorsed in Reference 2 [GL
93-05], the proposed changes to the testing frequency will increase
the margin of safety through reduced equipment wear and elimination
of opportunities to induce transients.
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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Ledyard B. Marsh.
Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County,
Michigan
Date of amendment request: April 26, 1994.
Description of amendment request: The proposed amendment would
relocate tables of instrument response time limits from the Technical
Specifications to the Updated Final Safety Analysis Report (UFSAR) in
accordance with the guidance contained in Generic Letter 93-08 dated
December 29, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes delete and subsequently relocate the details of
Technical Specification Table 3.3.1-2, ``REACTOR PROTECTION SYSTEM
RESPONSE TIMES,'' Table 3.3.2-3, ``ISOLATION ACTUATION SYSTEM
INSTRUMENTATION RESPONSE TIME,'' and Table 3.3.3-3, ``EMERGENCY CORE
COOLING SYSTEM RESPONSE TIMES,'' consistent with the guidance
provided by Generic Letter 93-08 dated, December 29, 1993, entitled,
``Relocation of Technical Specification Tables of Instrument
Response Time Limits.'' Generic Letter 93-08 recommends the removal
and subsequent relocation of various Technical Specification tables
which denote instrument and system response time limits. The
response time limits and associated footnotes are proposed to be
relocated to the Fermi 2 Updated Final Safety Analysis Report
(UFSAR). This allows Fermi 2 to administratively control subsequent
changes to the response time limit tables in accordance with 10 CFR
50.59. The procedures which contain the various response time limits
are also subject to the change control provisions in the
Administrative Controls section of the Technical Specifications. The
proposed change only relocates the existing response time limits.
The Surveillance Requirements and associated Actions are not
affected and remain in the Technical Specifications. Relocating this
information does not affect the initial conditions of a design basis
accident or transient analysis. Since any subsequent changes to the
UFSAR or procedures are evaluated in accordance with 10 CFR 50.59,
no increase in the probability or consequences of an accident
previously evaluated is allowed. Further, the proposed changes do
not alter the design, function, or operation of the components
involved and therefore, do not affect the consequences of any
previously evaluated accident.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously
evaluated. The proposed changes will not impose any different
operational or surveillance requirements. The changes propose to
relocate these response time limit tables to other plant documents
whereby adequate control of information is maintained. Further, as
stated above, the proposed changes do not alter the design,
function, or operation of the components involved and therefore, no
new accident scenarios are created.
3. The proposed changes do not involve a significant reduction
in a margin of safety. The proposed change will not reduce a margin
of safety because it has no impact on any safety analysis
assumption. The proposed change does not alter the scope of
equipment currently required to be OPERABLE or subject to
surveillance testing nor does the proposed change affect any
instrument setpoints or equipment safety functions. In addition, the
values to be transposed from the Technical Specifications to the
UFSAR are the same as the exiting Technical Specifications. Since
any future changes to these requirements in the UFSAR or procedures
will be evaluated per the requirements of 10 CFR 50.59, no reduction
in a margin of safety is allowed. Therefore, the change does not
involve a significant reduction in 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: Monroe County Library System,
3700 South Custer Road, Monroe, Michigan 48161.
Attorney for licensee: John Flynn, Esq., Detroit Edison Company,
2000 Second Avenue, Detroit, Michigan 48226.
NRC Project Director: Ledyard B. Marsh.
Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of amendment request: March 30, 1994.
Description of amendment request: The proposed amendments would
allow the analog channel operational test interval for radiation
monitoring instrumentation to be increased from monthly to quarterly.
The proposed amendments are said by the licensee to be consistent with
NRC staff recommendations and guidance contained in NUREG-1366,
``Improvements to Technical Specifications Surveillance Requirements,''
and Generic Letter 93-05, ``Line-Item Technical Specifications
Improvements to Reduce Surveillance Requirements for Testing During
Power Operation.''
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
The requested amendments will not involve a significant increase
in the probability or consequences of an accident previously
evaluated. Decreasing the frequency of the radiation monitor analog
channel operational test from monthly to quarterly will have no
impact upon the probability of any accident, since the radiation
monitors are not accident initiating equipment. Also, no credit is
taken in accident analyses for automatic actions performed by
radiation monitors contained in Catawba's Technical Specifications,
so the requested amendments will have no adverse impact upon the
consequences of any accident.
Criterion 2
The requested amendments will not create the possibility of a
new or different kind of accident from any accident previously
evaluated. As stated above, the radiation monitors are not accident
initiating equipment. No new failure modes can be created from an
accident standpoint. The plant will not be operated in a different
manner.
Criterion 3
The requested amendments will not involve a significant
reduction in a margin of safety. Plant safety margins will be
unaffected by the proposed changes. No safety equipment which is
taken credit for in accident analyses will be affected by the
requested amendments. The availability of the affected radiation
monitors will be increased as a result of the proposed amendments
because the monitors will not have to be made unavailable for
testing as frequently. In addition, radiation monitor operating
experience supports the proposed amendments. Finally, the proposed
amendments are consistent with the NRC position and guidance set
forth in NUREG-1366 and Generic Letter 93-05.
Based upon the preceding analyses, Duke Power Company concludes
that the requested amendments do not involve a significant hazards
consideration.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: York County Library, 138 East
Black Street, Rock Hill, South Carolina 29730.
Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422
South Church Street, Charlotte, North Carolina 28242.
NRC Project Director: David B. Matthews.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: April 19, 1994.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications (TS) 4.0.5 a,
``Applicability--Surveillance Requirements.'' The licensee proposes to
delete the wording ``. . . (g), except where specific written relief
has been granted by the Commission pursuant to 10 CFR, Section
50.55a(g)(6)(i)'' in TS 4.0.5 a, for the inservice inspection and
testing programs. With the revisions to the Technical Specifications,
upon finding an ASME Code requirement impractical because of
prohibitive dose rates or limitations in the design, construction, or
system configuration, the licensee may implement the relief request
once it has been submitted to the NRC provided it has been: (1)
Acceptably reviewed pursuant to 10 CFR 50.59; (2) approved by the plant
staff in accordance with the administrative process described in the
inservice inspection and testing programs administrative procedures;
and (3) reviewed and approved by the Plant Nuclear Safety Committee.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments remove the wording ``. . . (g), except
where specific written relief has been granted by the Commission
pursuant to 10 CFR, Section 50.55a(g)(6)(i)'', provided a 10 CFR
50.59 evaluation is performed. The Inservice Inspection and Testing
Programs are described in the Technical Specifications pursuant to
10 CFR 50.55a. In addition, the proposed amendments, in accordance
with NUREG 1431 and draft NUREG 1482, provide relief to the ASME
code requirement in the interim between the time of submittal of a
relief request until the NRC has issued a safety evaluation and
granted the relief. The changes being proposed are administrative in
nature and do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Any relief from the approved ASME Section XI code
requirements will require a 10 CFR 50.59 evaluation to ensure no
Technical Specification changes or unreviewed safety questions
exist. Therefore, operation of the facility in accordance with the
proposed amendments would not affect the probability or consequences
of an accident previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The changes being proposed are administrative in nature and will
not change the physical plant or the modes of operation defined in
the Facility License. The change does not involve the addition or
modification of equipment nor does it alter the design or operation
of plant systems. Any reliefs from the approved ASME Section XI code
requirements will require a 10 CFR 50.59 evaluation to ensure no
Technical Specification changes or unreviewed safety questions
exist. Therefore, operation of the facility in accordance with the
proposed amendments would not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The changes being proposed are administrative in nature and do
not alter the bases for assurance that safety-related activities are
performed correctly or the basis for any Technical Specification
that is related to the establishment of or maintenance of a safety
margin. Any reliefs from the approved ASME Section XI code
requirements will require a 10 CFR 50.59 evaluation to ensure no
Technical Specification changes or unreviewed safety questions
exist. Therefore, operation of the facility in accordance with the
proposed amendments 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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.
Florida Power and Light Company, Docket Nos. 50-250 and 50-251, Turkey
Point Plant Units 3 and 4, Dade County, Florida
Date of amendment request: April 19, 1994.
Description of amendment request: The licensee proposes to change
Turkey Point Units 3 and 4 Technical Specifications by increasing the
surveillance interval specified for air or smoke flow test through the
containment spray header from ``at least once per 5 years'' to ``at
least once per 10 years.'' The licensee stated that the proposed
surveillance interval is consistent with both Generic Letter 93-05,
``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation'' and
NUREG-1366, ``Improvements to Technical Specifications Surveillance
Requirements.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Operation of the facility in accordance with the proposed
amendments would not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed amendments extend the surveillance interval
required for performing a qualitative smoke or air flow test on the
containment spray headers. This surveillance test is not designed to
track degradation of equipment by monitoring or trending
performance. The air and smoke flow test is a test of the passive
design of the containment spray nozzles, i.e., the testing
demonstrates whether or not the nozzles are clogged. A single
failure rendering a significant number of nozzles inoperable as a
result of clogging is considered not credible. The changes being
proposed do not affect assumptions contained in plant safety
analyses, the physical design and/or operation of the plant, nor do
they affect Technical Specifications that preserve safety analysis
assumptions. Therefore, operation of the facility in accordance with
the proposed amendments would not involve a significant increase in
the probability or consequences of an accident previously analyzed.
(2) Operation of the facility in accordance with the proposed
amendments would not create the possibility of a new or different
kind of accident from any accident previously evaluated.
The proposed amendments extend the surveillance interval
required for performing a qualitative smoke or air flow test on the
containment spray headers. The changes being proposed will not
change the physical plant or the modes of plant operation defined in
the Facility License. The change does not involve the addition or
modification of equipment nor does it alter the design or operation
of plant systems. Therefore, operation of the facility in accordance
with the proposed amendments would not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
(3) Operation of the facility in accordance with the proposed
amendments would not involve a significant reduction in a margin of
safety.
The revised surveillance interval proposed by this submittal
will not change or otherwise influence the degree of operability
assumed for the containment spray system in the plant safety
analyses. The changes being proposed do not alter the bases for
assurance that safety-related activities are performed correctly or
the basis for any Technical Specification that is related to the
establishment of or maintenance of a safety margin. Therefore,
operation of the facility in accordance with the proposed amendments
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
amendment request involves no significant hazards consideration.
Local Public Document Room location: Florida International
University, University Park, Miami, Florida 33199.
Attorney for licensee: Harold F. Reis, Esquire, Newman and Holtzer,
P.C., 1615 L Street, NW., Washington, DC 20036.
NRC Project Director: Herbert N. Berkow.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: April 15, 1994.
Description of amendment request: The proposed amendment requests
the deletion of the audit program frequency requirements from Technical
Specification (TS) 6.5.3 and to utilize the Operational Quality
Assurance (OQA) Plan as the controlling document. This change will
introduce more flexibility into audit scheduling to consider plant
activities and performance. In addition, a minor editorial change has
been incorporated correcting a reference in TS 6.5.1.14 in response to
a finding in the Operational Safety Team Inspection report of December
23, 1993.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
GPU Nuclear has determined that this [Technical Specification
change request] TSCR poses no significant hazard as defined by the NRC
in 10 CFR 50.92.
1. These changes do not affect the function of any system or
component. Therefore, they do not increase the probability of
occurrence or consequence of an accident previously evaluated in the
[Safety Analysis Report] SAR.
2. These changes do not involve a physical change to plant
configuration and they do not affect the performance of any
equipment. Therefore, they do not create the possibility of an
accident or malfunction of a different type than previously
identified.
3. The shifting of the audit frequency requirements from the
Technical Specifications to the OQA Plan and the extension of the
maximum interval between audits of certain areas do not change the
activities to be audited nor the scope of individual audits.
Furthermore, audit frequencies are not associated with the margin of
safety in the bases of any Technical Specification.
Therefore, the margin of safety is not affected by this change.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: April 19, 1994.
Description of amendment request: The proposed change updates and
clarifies Technical Specification 3.4.B.1 to be consistent with
existing Specifications 1.39 and 4.3.D (ASME Code Section XI, Article
5000 requirements).
The requested change would delete reference to the ASME Code
Section XI, IS-5000 ten year hydrotest inspection interval and replace
this with references to: (1) The Technical Specification 1.39
definition for Reactor Vessel Pressure Testing, and (2) the Technical
Specification 3.3.A.(i) Reactor Vessel Pressure Testing limits (P/T and
250 deg.F maximum test temperature).
The requested change will clarify that the five electromatic relief
valves' (EMRV) pressure relief function may be inoperable or bypassed
during system pressure testing required by ASME Code Section XI,
Article IWA-5000, including system leakage and hydrostatic test, with
reactor vessel completely solid, core not critical and Technical
Specification 3.2.A (Core Reactivity limits) satisfied.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The requested change will not involve a significant increase
in the probability or consequence of any accident previously
evaluated because this change: (a) Merely updates and clarifies
Technical Specification 3.4.B.1 to be consistent with other existing
Technical Specifications, (b) contains no adverse changes to any
existing safety function necessary for the reactor vessel solid,
core not critical condition, and (c) makes no modification or
physical changes to plant equipment, performance or operation
necessary to respond to accidents for the reactor vessel solid, core
not critical condition.
2. The requested change does not create the possibility of a new
or different accident from any accident previously evaluated because
this change: (a) Merely updates and clarifies Technical
Specification 3.4.B.1 to be consistent with other existing Technical
Specifications, (b) contains no adverse changes to any existing
safety function necessary for the reactor vessel solid, core not
critical condition, and (c) over pressure protection would continue
to be provided by the code safety valves when the EMRV pressure
relief function is bypassed.
3. A significant reduction in margin of safety is not involved
because even though the EMRV pressure relief function is bypassed,
over pressure protection would continue to be provided by the code
safety valves. Elimination of this relief function does not affect
the reactor safety analysis, since credit was not taken for the EMRV
pressure relief function . . . .
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Ocean County Library,
Reference Department, 101 Washington Street, Toms River, New Jersey
08753.
Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw,
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Project Director: John F. Stolz.
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: March 15, 1994.
Description of amendment request: The proposed amendment would
revise the technical specifications (TS) by removing TS 3/4.3.8,
``Turbine Overspeed Protection System,'' from the TS and relocating it
to an administratively controlled document.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
This change request proposes deletion of Technical Specification
3/4.3.8, ``Turbine Overspeed Protection System'' and relocates this
requirement to an existing plant program. The purpose of overspeed
protection is to minimize the possible generation of turbine
fragment missiles. Excessive overspeed could potentially result in
the generation of missiles which could impact and damage safety
related components, equipment or structures, depending on the size
and trajectory of the missiles. The proposed deletion of this
specification is based on the low probability of the generation of a
damaging turbine missile and other existing performance
verifications of the overspeed protection system.
The turbine-generator orientation at RBS [River Bend Station] is
a ``favorable'' orientation for reducing the probability of damage
to safety-related equipment from turbine missiles since all safety-
related components and structures are located in the axial direction
from the turbine-generator. Turbine Overspeed Protection System is
necessary for protection of the turbine from only an operational and
economic point of view. The system is not essential to mitigating
the consequences of an accident. The system is not used in an
initial condition of a design basis accident or transient analysis.
The probability of damage to safety-related equipment based on
turbine manufacturer's turbine failure data was calculated to be
1.473 x 10-8 per year and is acceptably low based on the
probability of turbine failure data of 4.75x10-7 per year as
recommended by NUREG-0800. Therefore, this proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
The change proposes to relocate this requirement to an existing
plant program, whereby adequate control of information is
maintained. The proposed change does not necessitate a physical
alteration of the plant (no new or different type of equipment will
be installed) or changes to parameters governing normal plant
operation. The proposed change will not impose any different
operational or surveillance requirements. No new failure modes are
introduced. Therefore, this proposed change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Does the change involve a significant reduction in a margin of
safety?
The proposed change will not reduce a margin of safety because
it has no impact on any safety analysis assumption. The proposed
change does not alter the scope of equipment currently required to
be OPERABLE or subject to surveillance testing, nor does the
proposed change affect any instrument setpoints or equipment safety
functions. The favorable orientation of the turbine provides a
margin of safety such that the possibility of missile damage to
safety-related equipment is acceptably low. Therefore the change
does not involve a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Government Documents
Department, Louisiana State University, Baton Rouge, Louisiana 70803.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, D.C. 20005.
NRC Project Director: William D. Beckner.
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: April 28, 1994.
Description of amendment request: The licensee proposes to revise
Technical Specification Surveillance Requirement 4.6.1.3.e to add an
option which will allow the personnel airlock pneumatic system leak
test to be completed in 8 hours with a pressure drop of 0.50 psi. The
technical specifications currently require that the door seal pneumatic
system be demonstrated operable by verifying that the system pressure
does not decay more than 1.5 psi within 24 hours. The change to an 8-
hour test will expedite return to power following an outage since the
test is on the critical path for restart following outages.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The door pneumatic seal system pressure drop test is not altered
except for providing an option to utilize a reduced test duration. A
conservative acceptance criteria of 0.50 psi will be assigned to the
optional short duration test thus maintaining the operability of the
pneumatic seal system. The proposed change does not alter equipment
or assumptions made in previously evaluated accidents, therefore the
consequences of previously evaluated accidents are not increased.
The probability of an accident is also unaffected because the seals
are not a potential accident initiator.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
With a conservative acceptance criteria of 0.50 psi assigned to
the optional 8 hour door pneumatic seal system pressure drop test
the capability of the door pneumatic seal system to maintain 65 psig
to the airlock seals, for a minimum of 15 days upon a loss of
instrument air, is assured. Loss of plant supply air is the accident
evaluated in the UFSAR [Updated Final Safety Analysis Report]
section 3.8.2.1.2 and plant specification 2C269SS0006. The proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
To ensure the pneumatic seal system pressure drop test is not
compromised, a conservative acceptance criteria of 0.50 psi will be
assigned to the 8 hour test. With the conservative acceptance
criteria, the proposed change does not involve a significant
reduction in the margin of safety previously evaluated.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger,
P.C., 1615 L Street, N.W., Washington, D.C. 20036.
NRC Project Director: Suzanne C. Black.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1 Rockingham, New Hampshire
Date of amendment request: January 14, 1994.
Description of amendment request: The proposed amendment would
change the Technical Specifications (TS) to specify the composition of
the Station Operation Review Committee (SORC) based on experience and
expertise vice organizational position, to implement a Station
Qualified Reviewer Program (SQRP), to delete the requirement for
periodic procedure reviews, to revise the time within which the Nuclear
Safety Audit Review Committee (NSARC) must issue reports and minutes,
and to incorporate a number of editorial changes. The editorial changes
would delete certain items that are no longer applicable, would remove
inconsistencies involving the names of systems and equipment and NSARC
function, composition, and use of alternates, and would correct the
value for the reactor coolant system volume. Other editorial changes
would be made for document format consistency. The proposed amendment
would affect the following TS Sections and tables: 1.31, 3.3.3.6,
3.4.1.2, 4.6.3.2, 3.7.1.2, 3/4 10.6, 5.4.2, 6.3, and 6.4, and Table
4.3-1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below.
A. The changes do not involve a significant increase in the
probability or consequences of an accident previously evaluated (10 CFR
50.92(c)(1)).
The proposed redefinition of the composition of the SORC would not
diminish the effectiveness of the SORC and would continue to ensure
that the SORC has the desired experience and expertise to advise the
Station Manager on all matters related to nuclear safety. The proposed
change would permit operational flexibility and eliminate the need for
an amendment whenever organizational changes occur. The proposed SQRP
would not reduce the level of procedure review, since the SORC
continues to retain responsibility to review any document requiring an
evaluation pursuant to 10 CFR 50.59. The SQRP would be limited to
reviewing procedures that do not affect nuclear safety.
Deleting the requirement to periodically review procedures would
not diminish the review process for procedures since other programmatic
requirements would continue to assure procedures are reviewed and
revised when necessary.
The proposed extension of time for preparing and forwarding NSARC
meeting minutes would not affect safe operation of the facility.
Significant safety concerns or unreviewed safety questions would still
be brought to the attention of the Senior Vice President without
waiting for the release of the NSARC meeting minutes. The change would
not impede in any manner prompt communication of significant concerns
to the Senior Vice President. The proposed changes do not affect the
manner by which the facility is operated and do not change any facility
design feature or equipment. The proposed changes involve
administrative or programmatic requirements or merely involve editorial
changes, corrections, or clarifications. Since there is no change to
the facility or operating procedures, there is no effect upon the
probability or consequences of any accident previously analyzed.
B. The changes do not create the possibility of a new or different
kind of accident from any accident previously evaluated (10 CFR
50.92(c)(2)) because they do not affect the manner by which the
facility is operated and do not change any facility design feature or
equipment which affects the operational characteristics of the
facility. The proposed changes involve administrative or programmatic
requirements or merely involve editorial changes, corrections, or
clarifications.
C. The changes do not involve a significant reduction in a margin
of safety (10 CFR 50.92(c)(3)) because the proposed changes do not
affect the manner by which the facility is operated or involve
equipment or features which affect the operational characteristics of
the facility.
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: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One
International Place, Boston, Massachusetts 02110-2624.
NRC Project Director: John F. Stolz.
Northeast Nuclear Energy Company, et al., Docket No. 50-423,
Millstone Nuclear Power Station, Unit No. 3, New London County,
Connecticut
Date of amendment request: February 10, 1992, as supplemented April
14, 1994.
Description of amendment request: The proposed amendment would
remove two tables from the Technical Specifications (TS) which list
reactor trip system (RTS) instrumentation response times and engineered
safety features actuation system (ESFAS) instrumentation response
times. These tables will be placed in the Millstone 3 Technical
Requirements Manual.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve a significant hazards
consideration because the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes to remove the RTS and ESFAS response times
from the Technical Specifications will not affect the operation of
the RTS and ESFAS. Operability and surveillance requirements are
still maintained in the Technical Specifications and the response
times will be included and maintained in the plant operating
procedures. A safety evaluation and PORC [Plant Operations Review
Committee] review will be required for the limits to be changed.
Since the systems will not be affected by the proposed changes,
there is no impact on the performance of these systems or the
consequences of an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously evaluated.
There are no new failure modes associated with the proposed
changes. Since the plant will continue to operate as designed, the
proposed changes will not modify the plant response to the point
where it can be considered a new accident.
3. Involve a significant reduction in a margin of safety.
The proposed changes do not have any adverse impact on the
protective boundaries nor do they affect the consequences of any
accident previously analyzed. The Technical Specification
operability and surveillance requirements will still ensure that the
systems are tested and within the limits. Changing the limits
requires a safety evaluation and PORC review which will ensure that
the licensing basis is maintained. Therefore, the proposed changes
will not impact the margin of safety as defined in the basis of any
Technical Specification.
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, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: April 22, 1994.
Description of amendment request: The proposed amendment would
delete the requirements regarding the condenser air ejector monitor
from Tables 3.3-12 and 4.3-12 of the Millstone Unit 2 Technical
Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed technical specification change has been reviewed
against the criteria of 10 CFR 50.92, and it has been determined not to
involve a significant hazards consideration (SHC). Specifically, the
proposed change does not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
Deleting the operability and surveillance requirements for the
condenser air ejector monitor from Tables 3.3-12 and 4.3-12 of the
Millstone Unit No. 2 Technical Specifications would leave the steam
generator blowdown monitor as the primary method of monitoring and
isolating steam generator blowdown. The proposed license amendment
imposes stricter limitations on the operation of Millstone Unit No.
2, because it requires the use of a single monitor, the steam
generator blowdown monitor, to meet the requirements of Millstone
Unit No. 2 Technical Specification 3.3.3.9 (Table 3.3-12).
While NNECO [Northeast Nuclear Energy Company] is proposing to
delete the operability and surveillance requirements for the
condenser air ejector monitor from the Millstone Unit No. 2
Technical Specifications, there are no plans to change any of the
design features or functions or the condenser air ejector monitor,
or any of the specified surveillances or frequency for such
surveillances. The condenser air ejector monitor will continue to
isolate blowdown upon a high radiation alarm.
Additionally, steam generator blowdown isolation is required to
ensure compliance with 10 CFR 20. It is not required to ensure
compliance with 10 CFR 100. Therefore, the condenser air ejector
monitor does not perform any safety function. The condenser air
ejector monitor is not safety related. It is not credited in any
radiological consequence calculations presented in the Millstone
Unit No. 2 FSAR [Final Safety Analysis Report].
Based on the above, this proposed license amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
form any accident previously evaluated.
The proposed license amendment does not involve any physical
changes to plant equipment or any changes to plant procedures that
would be a precursor to an accident. NNECO has no plans to change
any of the specified surveillances or frequency for such
surveillances. The condenser air ejector monitor will continue to
isolate blowdown upon a high radiation alarm. Also, the proposed
license amendment imposes stricter limitations on the operation of
Millstone Unit No. 2 because it requires the use of a single
monitor, the steam generator blowdown monitor, to meet the
requirements of Millstone Unit No. 2 Technical Specification 3.3.3.9
(Table 3.3-12). Therefore, this proposed license amendment does not
create the possibility of a new or different kind of accident form
any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Deleting the operability and surveillance requirements for the
condenser air ejector monitor from Tables 3.3-12 and 4.3-12 of the
Millstone Unit No. 2 Technical Specifications would leave the steam
generator blowdown monitor as the primary method of monitoring and
isolating steam generator blowdown. The proposed license amendment
imposes stricter limitations on the operation of Millstone Unit No.
2, because it requires the use of a single monitor, the steam
generator blowdown monitor, to meet the requirements of Millstone
Unit No. 2 Technical Specification 3.3.3.9 (Table 3.3-12).
Therefore, this proposed license amendment does not impact or 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, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: April 22, 1994.
Description of amendment request: The proposed amendment would
modify the Millstone Unit 2 Technical Specification Table 3.3-9 by
eliminating the measurement range of 10-1-104 counts per
second (CPS) for the entry regarding the ``Wide Range Logarithmic
Neutron Flux Monitor.'' Also the amendment would correct a few
typographical and editorial errors on page V of the Index for the
Millstone Unit 2 Technical Specifications.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
NNECO [Northeast Nuclear Energy Company] has reviewed the proposed
changes in accordance with 10 CFR 50.90 and has concluded that the
changes do not involve a significant hazards consideration (SHC). The
basis for this conclusion is that the three criteria of 10 CFR 50.92(c)
are not compromised. The proposed changes do not involve an SHC because
the changes would not:
1. Involve a significant increase in the probability or
consequences of an accident previously analyzed.
NNECO's proposal to eliminate the CPS scale for the ``Wide Range
Logarithmic Neutron Flux Monitor'' entry in Millstone Unit No. 2
Technical Specification Table 3.3-9 will not affect the ability of
Millstone Unit No. 2 to meet the intent and purpose of panel C-21's
original design.
The 10-8% to 100% power scale overlaps the CPS scale. The
range of 10-8% to 100% power for the ``Wide Range Logarithmic
Neutron Flux Monitor'' is adequate to permit the operators to bring
the unit to hot shutdown from outside the control room. Also, the
instruments on C-21 are not used to provide the start-up rate signal
during start-up or refueling operations. This proposed license
amendment does not impact the performance of any safety-related
component, system, or structure.
A review of the original design drawings concluded that this
proposed change is consistent with the original plant design, and
reflects the actual as-built condition of the unit. The original
design drawings show that the wide range logarithmic neutron flux
indicators only receive a percent power signal.
NNECO's proposals to rectify a few typographical and editorial
errors on page V of the Index for the Millstone Unit No. 2 Technical
Specifications are administrative in nature. They ensure that the
Index accurately reflects the contents of the technical
specifications.
Based on the above, the proposed license amendment does not
involve a significant increase in the probability or consequences of
an accident previously analyzed.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed license amendment does not impact the performance
of any safety-related component, system, or structure. Panel C-21 is
required to permit the operators to bring the unit to a hot shutdown
condition from a location outside the control room. Deleting the CPS
range for the ``Wide Range Logarithmic Neutron Flux Monitor'' does
not affect the ability of the operators to accomplish this function.
Also, the proposed change is consistent with the original design of
the plant. The proposed license amendment cannot create the
possibility of a new or different kind of accident form any
previously analyzed.
3. Involve a significant reduction in a margin of safety.
NNECO's proposal to eliminate the CPS scale for the wide range
logarithmic neutron flux monitors will not affect the ability of
Millstone Unit No. 2 to meet the intent and purpose of panel C-21's
original design. The 10-8% to 100% power scale overlaps the CPS
scale. The range of 10-8% to 100% power for the ``Wide Range
Logarithmic Neutron Flux Monitor'' is adequate to permit the
operators to bring the unit to hot shutdown from outside of the
control room. Also, the instruments on C-21 are not used to provide
the start-up rate signal during start-up or refueling operations.
This proposed license amendment does not impact the performance of
any safety-related component, system, or structure.
Therefore, this proposed license amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut
Date of amendment request: April 25, 1994.
Description of amendment request: The proposed amendment would
change the Technical Specifications concerning four related issues: (1)
Power-operated relief valve (PORV) and block valve reliability; (2)
low-temperature overpressure protection (LTOP); (3) boron dilution; and
(4) shutdown risk management. Specifically, the proposed amendment
would revise Technical Specifications 3.4.3 and 3.4.9.3 to address the
issues specifically raised in Generic Letter (GL) 90-06. Technical
Specifications 3.1.1.3, 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.1.2.8,
3.4.1.4, 3.4.2.1, 3.4.9.1, 3.5.3, 4.1.1.3, 4.1.2.3, 4.1.2.4, 4.4.1.4,
4.4.3.1., 4.4.3.2, 4.4.9.3.1, 4.4.9.3.2, 4.5.3.2 and 4.9.8.1 would be
revised to provide consistency with the proposed changes in GL 90-06 or
are related to the boron dilution issue or shutdown risk management
philosophies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
The proposed changes do not involve an SHC [significant hazards
consideration] because the changes would not:
1. Involve a significant increase in the probability or
consequence of an accident previously evaluated.
The proposed changes address the operability and surveillance
requirements for the charging pump, HPSI [high-pressure safety
injection] pumps, reactor coolant pumps, safety valves, PORVs, block
valves, and the LTOP, boron dilution and SDC [shutdown cooling]
systems. These changes were proposed to address four main issues: to
reflect the guidance of GL 90-06 with respect to PORV and cold
overpressure; to address boron dilution concerns; to address
shutdown risk management lessons learned; and to address recent
information on cold overpressure mitigation concerns. Generally, the
changes are more restrictive than present requirements and are
consistent with the recommendations of GL 90-06. Also, the changes
provide the operator with additional guidance that was not
previously available. Therefore, the changes will not impact the
probability of occurrence or consequences of an LTOP event, boron
dilution event, loss of shutdown cooling, or other event requiring
emergency core cooling which has been previously analyzed.
PORV Requirements
The proposed changes to Technical Specification 3.4.3 have been
made to be consistent with GL 90-06. One enhancement has been made
to the guidance contained in GL 90-06 and that was to replace the
phrase ``because of excessive seat leakage'' with the phrase ``and
capable of being manually cycled.'' Although the PORV may be
designated inoperable, it may be able to be manually opened and
closed and in this manner can be used to mitigate transients. For
example, PORV inoperability may be due to seat leakage,
instrumentation problems, automatic control problems, or other
causes that do not prevent manual use and do not create a
possibility for a small break LOCA. The wording changes are meant to
be more specific while meeting the intent of GL 90-06. The
additional enhancement to GL 90-06 includes Surveillance Requirement
4.4.3.1c whereby Millstone Unit No. 2 proposed to bench test the
PORVs at a qualified laboratory under conditions representative of
Mode 3 or 4 conditions. We believe this off site test will result in
safer plant conditions than the in situ test proposed in the generic
letter. The remaining changes to Technical Specification 3.4.3
incorporate the guidance contained in GL 90-06 and do not
significantly increase the probability or consequence of an LTOP
event or the failure of the PORV to operate as required.
Cold Overpressurization Protection
Changes are being proposed to Technical Specification sections
3.1.2.1, 3.1.2.3, 3.4.1.4, 3.4.2.1, 3.4.3, 3.4.9.1, 3.4.9.3, 3.5.3,
4.1.2.3, 4.4.1.4, 4.4.3.1, 4.4.3.2, 4.4.9.3.1, 4.4.9.3.2, and
4.5.3.2 to incorporate the guidance of GL 90-06 as well as enhance
the availability of equipment to reduce the shutdown risk while
still satisfying the cold overpressure requirements.
The proposed changes to Technical Specifications 3.1.2.1 and
3.1.2.3 will ensure only one charging pump and one HPSI pump are
operable in Mode 5 or 6 with the reactor vessel head on with an
available vent of less than 2.8 square inches. The remaining pumps
will be secured. These proposed changes have been made to ensure
Millstone Unit No. 2 does not create an LTOP condition by the
operation of too many pumps injecting fluid, thereby increasing
pressure in a low-temperature condition. These proposed
modifications are consistent with Technical Specification 3.5.3
which has also been modified and will decrease the possibility of an
LTOP condition from occurring.
The proposed change to Technical Specification 3.4.2.1 will
ensure consistency between this technical specification and
Technical Specification 3.4.9.3. The safety valves at Millstone Unit
No. 2 are not used for LTOP mitigation. The PORVs, or RCS [reactor
coolant system] vent at Millstone Unit No. 2 are used to mitigate an
LTOP condition. Safety valves are required to be operable during
operating conditions to automatically reduce system pressures. The
use of the PORV, which allows manual control, for mitigation of an
LTOP event, reduces the severity and consequence of a potential
overpressure event by giving the operators more control.
The proposed changes to Technical Specification 3/4.9.3 provide
enhanced operational flexibility through the use of a PORV or RCS
vent. The APPLICABILITY statement has been changed for clarification
purposes with no change in intent and safety implications. The
ACTION requirements for the LTOP system include a 7-day allowable
outage time (AOT) to restore an inoperable LTOP channel to operable
status before other remedial measures would have to be taken. In
addition, new Action Statement `f' states that the provisions of
Specification 3.0.4 are not applicable. Therefore, the unit may
enter the Modes for which the LCO apply, during a unit shutdown or
placement of the head on the reactor vessel following refueling,
when an LTOP channel is inoperable. In this situation, the 7-day AOT
applies for restoring the channel to operable status before other
remedial measures would have to be taken. This is the same manner in
which the ACTION requirements apply when an LTOP channel is
determined to be inoperable while the plant is in a Mode for which
the LTOP system is required to be operable.
Specifications 3.4.1.4 and 3.4.9.1 have been revised to address
concerns identified in an NRC Information Notice regarding
previously unconsidered pressure drops across the reactor. The
modifications to these two technical specifications will ensure that
unanticipated pressure rises do not occur and that there will be no
increase in the probability or consequences of the LTOP event.
Based on the evaluation done in support of resolution to GL 90-
06 regarding the LTOP system unavailability, NNECO concludes that
additional restrictions on operation with an inoperable LTOP channel
are warranted when the potential for a low-temperature overpressure
event is the highest, and especially when the unit is in a water-
solid condition. It is also concluded that these additional measures
emphasize the importance of the LTOP system, especially while
operating in a water-solid condition as the primary success path for
the mitigation of overpressure transients during low-temperature
operation. Therefore, these enhancements will not involve a
significant increase in the probability or consequence of an
accident previously evaluated.
Boron Dilution
Changes are being proposed to Technical Specifications 3.1.1.3,
3.1.2.2, 3.1.2.3, 3.1.2.4, 3.1.2.8, 4.1.1.3, 4.1.2.3, and 4.1.2.4 to
provide added assurance that the boron dilution analysis remains
bounding while allowing lower flow rates to reduce the potential of
a loss of shutdown cooling due to vortexing at mid-loop operation.
The changes to Technical Specifications 3.1.1.3, 3.1.2.2,
3.1.2.3, 3.1.2.4, 3.1.2.8, 4.1.1.3, 4.1.2.3, 4.1.2.4, and 4.9.8.1
will not significantly increase the probability or consequences of
an accident. Tagging out of a charging pump, increasing shutdown
margin, and reducing SDC flow will impact results of the boron
dilution accident, but will not increase the probability of
initiating events.
An increase in the shutdown margin requirement as was done in
Technical Specifications 3.1.2.2 and 3.1.2.8 will assure consistency
with the Core Operating Limits Report which provided additional
margin in a boron dilution event.
Shutdown Risk
The changes proposed to Technical Specifications 3.1.1.3,
3.1.2.1, 3.1.2.3, 3.5.3, 4.1.1.3, 4.1.2.3, 4.5.3.2 and 4.9.8.1 have
been optimized to take into account shutdown risk concerns. Lower
shutdown cooling flow rates are allowed to minimize the potential of
a loss of shutdown cooling due to vortexing during RCS mid-loop
operation.
The availability of injection sources in the shutdown modes have
been optimized while still meeting the cold overpressurization
requirements.
To address shutdown risk issues, the method to secure an
inoperable HPSI pump has been modified. Previously, disconnecting
the motor circuit breaker from its electrical power circuit was the
only acceptable method of isolating this pump. Additional methods of
isolating the pump have been added with the key locking of a
discharge valve downstream of the HPSI pump and the tagging the
valve. These actions from the control room will allow the operator
the ability to quickly restore water flow and reduce the risk
associated with having equipment out of service while shutdown.
Inadvertent actuation is prevented by requiring the operator to
obtain the key to open this discharge valve from the shift
supervisor. The opening of this valve would, therefore, require the
actions of two knowledgeable individuals, the operator, and the
shift supervisor. The limitation on the amount of pumps available is
as a direct result of LTOP concerns. This provides assurance that
the LTOP requirements are met while maintaining the maximum
available equipment to mitigate shutdown risk concerns.
2. Create the possibility of a new or different kind of accident
from any previously analyzed.
The proposed changes to Technical Specifications 3.1.1.3,
3.1.2.1, 3.1.2.2, 3.1.2.4, 3.1.2.8, 3.4.1.4, 3.4.2.1, 3.4.9.1,
4.1.1.3, 4.1.2.4, and 4.9.8.1 do not create the possibility of a new
or different kind of accident from any previously analyzed. The
proposed changes provide clarification or additional restrictions
for plant personnel concerning the operation of charging pumps, HPSI
pumps, PORVs, blocking valves, and the SDC, boron dilution, and LTOP
systems. The proposed technical specification changes do not
introduce significant changes in the manner in which the plant is
being operated. Therefore, no new failure modes are being
introduced, and the potential for an unanalyzed accident is not
created.
The proposed changes to Technical Specifications 3.4.3 do not
create the possibility of an accident of a different type than
previously evaluated, since there is no change to the design of the
plant. In addition, plant operations are only being altered enough
to allow a block valve and PORV to be placed in conditions which
allow them to better perform their safety functions.
The proposed changes to Technical Specification 3.4.9.3 do not
create the possibility of an accident of a different type than
previously evaluated, since there is no change to the design of the
plant and the way the plant is operated.
The proposed changes to Technical Specification 3.1.2.3 and
3.5.3 allow for the isolation of an inoperable HPSI pump by the key
lock closing of a valve at the discharge of the HPSI pump and the
safety tagging in the closed position. This isolation is required so
that a LTOP condition does not occur. This method of isolation is
required so that a LTOP condition does not occur. This method of
isolation is acceptable and will not create a new or different kind
of accident since it is not possible to inadvertently open this
valve. A deliberate action is required by the operator, with the
concurrence of the shift supervisor, to obtain the key and open the
valve.
3. Involve a significant reduction in a margin of safety.
The proposed changes will not have an adverse impact on the
protection boundaries.
With regard to the GL 90-06 modifications, there is no
degradation in the operability and surveillance requirements for the
PORVs and block valves and the LTOP systems. There will be no change
in actual practice for, or resulting performance of, these systems.
All other changes are proposed mainly to clarify each requirement.
For Modes 1, 2, and 3, safety-related overpressure protection is
provided by the pressurizer code safety relief valves. Therefore,
there will be no adverse impact on the margin of safety as defined
in the bases of any technical specification. Although any two
charging pumps are allowed to be operable in a shutdown condition,
the flow of these pumps is consistent with the assumptions of the
boron dilution analysis. Additional pumping capability is being
provided to address shutdown risk concerns, however, the limitation
on pumping is tied to the vent path that is available. This will
ensure that the margin of safety is not impacted.
The combined effects of reducing SDC flow, tagging out a
charging pump, and increasing shutdown margin is that the required
operator response times of 15 minutes in Modes 4 and 5, and 30
minutes in Mode 6 are maintained.
By reducing the allowed SDC flow rate to less than that where
vortexing can occur, the potential for a loss of SDC event is being
reduced. Therefore, there is no decrease in the margin of safety for
the boron dilution and shutdown cooling events.
The proposed changes associated with the cold overpressure
mitigation system will ensure the appropriate margin of safety is
maintained by limiting RCP operation in Mode 5 and limit RCS
cooldown rates. These actions will ensure an LTOP condition does not
occur.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
Attorney for licensee: Gerald Garfield, Esquire, Day, Berry &
Howard, City Place, Hartford, Connecticut 06103-3499.
NRC Project Director: John F. Stolz.
Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County,
Pennsylvania
Date of amendment request: April 5, 1994.
Description of amendment request: This amendment will delete the
frequency requirements for a number of audits listed under Technical
Specification (TS) 6.5.2.8 for each unit. The proposed change also
includes removing the audit requirements for the Emergency Plan and the
Security Plan from the TS and relocating these requirements to each of
the respective plans.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
I. This proposal does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
The proposed Technical Specification changes to delete
prescribed audit frequencies and remove the Emergency Plan and
Security Plan from Technical Specifications are administrative in
nature and neither directly increase or decrease the likelihood that
an accident will occur. The Technical Specification changes will not
impact the function or method of operation of plant systems,
structures, or components. Thus, the consequences of a malfunction
of equipment important to safety previously evaluated in the FSAR is
not increased by the changes. Therefore, it is concluded that the
proposed changes do not increase the probability or consequences of
an accident previously evaluated.
II. This proposal does not create the possibility of a new or
different kind of accident or from any accident previously
evaluated.
The proposed Technical Specification changes to delete
prescribed audit frequencies and remove the Emergency Plan and
Security Plan from Technical Specifications are administrative in
nature and do not involve changes to the physical plant or
operations. The proposed changes do not affect systems, structures,
or components (SSCs) or the operation of these SSCs; and therefore
do not create the possibility of a new or different kind of
accident.
III. This change does not involve a significant reduction in a
margin of safety.
The proposed Technical Specification changes to delete
prescribed audit frequencies and remove the Emergency Plan and
Security Plan from Technical Specifications do not involve any
reductions in the margin of safety. The proposed changes will enable
more effective resource utilization through performance based
scheduling of audits in the affected areas. Using performance
indicators and other measures of program effectiveness, potential
problems can be more readily identified and audit resources can be
applied to these areas to enhance performance. The proposed
performance based audit process will maintain or enhance the margin
of safety in the areas audited.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Osterhout Free Library,
Reference Department, 71 South Franklin Street, Wilkes-Barre,
Pennsylvania 18701.
Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
NRC Project Director: Charles L. Miller.
Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: March 28, 1994.
Description of amendment request: The proposed modification to
Technical Specification (TS) Section 4.8.4.3.a, would increase the
surveillance interval for the functional test of the Reactor Protection
System (RPS). The increase would be from every six (6) months to each
time the plant is in cold shutdown for a period of 24 hours, unless the
test was performed in the previous six months. This change is based on
guidance provided in Generic Letter 91-09, ``Modification Of
Surveillance Interval For The Electrical Protective Assemblies In Power
Supplies For The Reactor Protection System.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specification changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The Reactor Protection System equipment subject to the proposed
Technical Specifications changes are not accident initiators.
The Electrical Protective Assemblies (EPAs) specified by these
proposed changes are not required to actuate in order to mitigate an
accident. The functional test methodology of the RPS electrical
power monitoring channels will not be effected by the proposed
change in test frequency. The design and function of the EPAs will
not be altered and will perform as originally designed.
A review of the RPS electrical power monitoring relays
surveillance test history results was performed and supports the
proposed TS changes to extend the testing interval. Fifty-one (51)
surveillance tests were reviewed, and all the as-found channel
calibration results were within the required TS limits. There were
identified deficiencies in four (4) of the fifty-one tests
performed, however, these four deficiencies did not affect the
operability of the RPS EPAs. Based on good historical surveillance
test results, we have concluded that the reliability of the
equipment is not expected to degrade during the proposed extended
test interval. Furthermore, the proposed reduced testing will result
in a net decrease in the probability of occurrence of a malfunction
of equipment important to safety. These malfunctions would cause an
invalid inadvertent trip of the RPS which would impose unnecessary
challenges on the affected unit at power. The guidance set forth in
Generic Letter 91-09 states ``The staff concludes that the benefit
to safety of reducing the frequency of testing during power
operations more than offsets the risk to safety from relaxing the
surveillance requirement to test the EPAs during power operation.''
Since the RPS EPAs are not accident initiators, and the design
and function of the equipment will not be affected by the proposed
TS changes, and the reliability of the equipment is not expected to
degrade during the extended test interval, and the changes would
reduce the probability of unnecessary challenges to the affected
unit, we have concluded that the proposed changes do not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed TS changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The design and function of the RPS EPAs will not be affected by
the proposed TS changes. The failure modes of the existing equipment
will remain unchanged, and no new accident types will be created.
The RPS electrical power monitoring channels' functional test
methodology will not be affected by the proposed change in test
frequency. Therefore, the proposed TS changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed TS changes do not involve a significant
reduction in a margin of safety.
Based on a review of the RPS electrical power monitoring relays
surveillance test history results we have concluded that the
reliability of the equipment is not expected to degrade during the
proposed extended test interval. In addition, the benefit to safety
by reducing the frequency of testing during power operation and the
attendant possible challenges to safety systems more than offsets
any risk to safety from relaxing the surveillance requirements to
test the EPAs during power operation. Therefore, the proposed TS
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
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: Charles L. Miller.
Philadelphia Electric Company, Docket No. 50-352, Limerick Generating
Station, Unit 1, Montgomery County, Pennsylvania
Date of application for amendment: May 6, 1994.
Description of amendment request: The amendment would revise Unit 1
Technical Specifications, Section 5.5.3, ``Capacity,'' to permit an
interim increase in the spent fuel storage capacity in the Unit 1 Spent
Fuel Pool (SFP) from 2040 fuel assemblies to 2500 fuel assemblies.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed Technical Specifications (TS) change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Increasing the spent fuel storage capacity in the Unit 1 Spent
Fuel Pool (SFP) from 2040 fuel assemblies to 2500 fuel assemblies
does not increase the probability of occurrence of an accident.
Since all fuel handling activities will be performed using approved
procedures and compatible equipment, the probability of a fuel
handling accident occurring is unchanged.
Increasing the spent fuel storage capacity in the Unit 1 SFP to
2500 fuel assemblies will facilitate storing 1940 spent fuel
assemblies (including contingency) that have been discharged from
LGS, Units 1 and 2, and 560 low exposure fuel assemblies shipped to
LGS from the Shoreham Nuclear Power Station. The decay heat load
associated with the entire Shoreham fuel inventory is insignificant,
since it equates to less than 5% of the heat load generated from one
(1) recently discharged full power fuel bundle. Therefore, the
actual decay heat load to the Unit 1 SFP will be equivalent to that
which is generated from storing the 1940 spent fuel assemblies
discharged from LGS, Units 1 and 2.
Increasing the spent fuel storage capacity in the Unit 1 SFP to
accommodate the storage of 2500 fuel assemblies, as proposed in this
TS Change Request, is bounded by the existing analysis supporting
the storage of spent fuel at LGS. The existing analysis considers
design inputs for structural integrity, criticality, and thermal-
hydraulics and is based on the storage of 2862 spent fuel
assemblies. As documented in Section 9.1.3, ``Spent Fuel Pool
Cooling and Cleanup Systems,'' of Supplement 2 of the NRC's Safety
Evaluation Report, i.e., NUREG-0991, ``Safety Evaluation Report
Related to the Operation of Limerick Generating Station, Units 1 and
2,'' the NRC indicated that based on its independent analysis the
heat removal capability of the Fuel Pool Cooling and Cleanup (FPCC)
system could only support 2484 spent fuel assemblies. However, the
LGS, Unit 1 TS currently limit the storage of spent fuel to 2040
spent fuel assemblies. Since the decay heat load from the Shoreham
fuel inventory (i.e., 560 fuel assemblies) is insignificant, the
actual heat load to the Unit 1 SFP will be equivalent to that
generated from 1940 fuel assemblies discharged from LGS, Units 1 and
2, which is less than the limit currently specified [in] the TS
(i.e., 2040 fuel assemblies).
Relocating six (6) of the existing Unit 2 spent fuel storage
racks to the Unit 1 SFP will be conducted in accordance with PECO
Energy's Heavy Loads Program which was developed in order to
implement the guidance delineated in NUREG-0612, ``Control of Heavy
Loads at Nuclear Power Plants,'' such that the likelihood of a heavy
load drop is precluded. The Unit 2 spent fuel storage racks are
identical to those already in use in the Unit 1 SFP. Procedures will
be in place to ensure that the Unit 2 spent fuel storage racks are
situated in the Unit 1 SFP to insure [ensure] proper neutron poison
alignment with the existing Unit 1 racks. The existing spent fuel
storage racks are designed for rack-to-rack contact during design
basis events without the loss of structural integrity. The racks are
also designed to withstand the impact from a dropped fuel assembly
without the loss of structural integrity or be damaged in a way that
could adversely affect the criticality analysis. Increasing the
spent fuel storage capacity to accommodate the storage of 2500 spent
fuel assemblies will not affect the spent fuel storage racks since
the racks are specifically designed to safely store spent fuel.
This proposed TS change will not prevent the ability of the FPCC
system from performing its design function to adequately cool the
SFP. The FPCC system will continue to function normally and be
capable of maintaining the SFP temperature at or below 140 deg.F.
The backup cooling and makeup systems (i.e., Residual Heat Removal
(RHR), Emergency Service Water (ESW), and Residual Heat Removal
Service Water (RHRSW) systems) will continue to function as designed
to provide an alternate source of cooling and makeup water to ensure
SFP cooling is maintained. The RHR system is still capable of
maintaining the SFP temperature less than 140 deg.F as described in
LGS Updated Final Safety Analysis Report (UFSAR). Increasing the
spent fuel storage capacity in the Unit 1 SFP will not increase the
probability of a loss of fuel pool cooling accident or adversely
affect the Refuel Floor ventilation system.
The consequences of a Fuel Handling Accident as described in the
LGS UFSAR are not increased since the number of fuel assemblies
stored in a SFP is not an input to the initial conditions of the
accident evaluation. This accident evaluates the dropping of a spent
fuel assembly and the fuel grapple assembly into the reactor core
during refueling operations. A drop height of 32 feet for the spent
fuel assembly and 47 feet for the fuel grapple assembly are assumed
and will produce the largest number of failed fuel rods. Since the
maximum possible height a fuel assembly can be dropped over the SFP
does not exceed 32 feet, the consequences of a Fuel Handling
Accident will not be increased by increasing the number of fuel
storage cells.
The consequences of a loss of fuel pool cooling as described in
Section 9.1.3.6 of the LGS UFSAR will not be increased. The event
described in the UFSAR assumes that the iodine in the fuel from past
refuelings is negligible, due to the long decay time. Iodine is the
major contributor to thyroid dose. Since the iodine in the fuel from
past refuelings is negligible, due to the long decay time,
increasing the spent fuel storage capacity will not increase the
dose due to the release of iodine in the SFP water resulting from
boiling and therefore, the consequences are not increased.
Increasing the storage capacity in the Unit 1 SFP, on an interim
basis, will not increase the probability of a malfunction of the
stored spent fuel since the existing thermal-hydraulic analysis
confirms that sufficient cooling capability exists to accommodate
the storage of 2500 fuel assemblies in the Unit 1 SFP. As for fuel
criticality, the existing analysis also confirms that the stored
fuel assemblies will remain sub-critical under normal and abnormal
conditions.
Increasing the storage capacity in the Unit 1 SFP will not
increase the probability of a malfunction of the SFP structure or
SFP liner. The existing structural analysis confirms that the SFP
structure has adequate margin to prevent overstressing and meets the
code requirements. Increasing the storage capacity in the Unit 1 SFP
will not increase the probability of a malfunction of the spent fuel
storage racks during design basis events based on the existing
seismic/structural analysis.
Increasing the on-site spent fuel storage capacity will not
increase the probability of a malfunction of the FPCC system. The
FPCC system will continue to function as designed.
The probability of a malfunction of fuel handling equipment will
not be increased since increasing the storage capacity in the Unit 1
SFP, as proposed, does not affect fuel handling equipment.
Increasing the spent fuel storage capacity does not increase the
consequences of a spent fuel assembly failure since the failure of
one (1) assembly will not result in additional spent fuel assembly
failures.
Increasing the spent fuel storage capacity will not increase the
consequences of spent fuel storage rack failure, since the existing
racks have been designed/qualified to limit the consequences of a
failure. A failure of, or damage to one (1) storage rack, will not
result in failure or damage to another storage rack.
Increasing the spent fuel storage capacity will not increase the
consequences of the failure of fuel handling equipment since the
maximum expected number of fuel rods damaged by a fuel handling
equipment failure remains as evaluated in the LGS UFSAR.
Therefore, the proposed TS change does not involve an increase
in the probability or consequences of an accident previously
evaluated.
2. The proposed TS change does not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
Increasing the spent fuel storage capacity in the LGS Unit 1 SFP
to permit an interim increase from 2040 fuel assemblies to 2500 fuel
assemblies will not create the possibility of an accident of a
different type. The Unit 1 SFP has been analyzed for criticality
effects, structural effects, radiological effects, and thermal-
hydraulic effects. The increase in spent fuel storage capacity will
be achieved by relocating six (6) existing spent fuel storage racks
from the Unit 2 SFP to the Unit 1 SFP. The spent fuel storage racks
are of identical design and are passive components; therefore, the
possibility of creating a new accident does not exist.
No new operating schemes or active equipment types will be
required to store additional fuel bundles in the SFP. Therefore, the
possibility of a different type of malfunction occurring is not
created.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed TS change does not involve a significant
reduction in a margin of safety.
Since the existing TS limits for fuel handling interlocks, heavy
loads restrictions, water coverage over irradiated fuel, in-core
decay time, and fuel sub-criticality will be maintained, the margin
of safety will not be reduced.
Therefore, the proposed TS change does not involve a reduction
in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Pottstown Public Library, 500
High Street, Pottstown, Pennsylvania 19464.
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: Charles L. Miller.
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: April 15, 1994.
Description of amendment request: The proposed amendment would: (1)
revise Unit 3 Technical Specification (TS) 3.3.A.2.f to correct a
typographical error, (2) revise the license and TSs to change the
licensee's name from Philadelphia Electric Company to PECO Energy
Company, (3) revise the frequency listed in TS 4.3.A.2.a for exercising
each partially or fully withdrawn operable control rod from every 24
hours to within 24 hours when operating above the rod worth minimizer
low power setpoint if there are three or more inoperable control rods
or if there is one fully or partially withdrawn rod which cannot be
moved and for which control rod drive mechanism damage has not been
ruled out, (4) revise TS 4.4.A.2 to allow for the replacement charge on
the explosive valve for the standby liquid control system to be from
either the same manufactured batch as the one fired or another batch
which has been certified by having one of the batches successfully
fired, (5) revise the frequency in TS 4.4.B.3 to functionally test each
standby liquid control system pump loop from monthly to at least once
per 92 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated
because the proposed changes do not alter the operation of equipment
assumed to be an initiator of any analyzed event or assumed to be
available for the mitigation of accidents or transients. Proposed
changes 1 and 2 are administrative in nature. Proposed change 3 to
reduce the requirement to verify insertion capability from every 24
hours to a single verification when one or more control rods are
stuck is sufficient to verify that the problem is not generic while
providing the benefit of removing a very resource intensive
requirement and permits licensed operators to focus on other, more
safety significant actions. Proposed change 4 will continue to
provide the necessary assurance that replacement charges on the
explosive valve of the standby liquid control system will be from a
batch from which a sample charge has been tested satisfactorily.
Proposed change 5 modifies the allowable interval between
surveillance tests for the standby liquid control system without
reducing the reliability of the system while providing the benefit
of reduced wear and tear on the system. Therefore, these proposed
changes do not increase the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any accident previously evaluated
because implementation of the proposed changes do not involve any
physical changes to plant systems, structures, or components. The
proposed changes do not allow plant operation in any mode that is
not already evaluated. Therefore, the possibility of a new or
different kind of accident from any accident previously evaluated is
not created.
3. The proposed changes do not involve a significant reduction
in a margin of safety because the proposed changes do not affect the
manner in which the facility is operated or change equipment or
features which affect the operational characteristics of the
facility. Proposed changes 1 and 2 are administrative in nature.
Proposed change 3 maintains the assurance that when a scram is
required that, at a minimum, the assumptions used in the accident
analysis will be met. Additionally, if the initial check of control
rod insertion is satisfactory, the subsequent checks are not likely
to identify similar problems because operating experience shows that
a [stuck] rod is rare. Once it has been determined that the same
problem is not occurring in other control rods the normal
surveillance frequency is sufficient to verify that scram capability
is maintained. Proposed change 4 provides added flexibility for
providing replacement [charges] from any batch that has had a charge
successfully fired. Proposed change 4 adds flexibility while
maintaining the firing reliability in excess of 99.99% for the
explosive valves on the standby liquid control system. Proposed
change 5 does not impact any safety analysis assumptions because the
frequency of testing is not assumed in any safety analysis and
standby liquid control system operability is maintained. In
addition, the test frequency reduction provides reduced wear and
tear on the system and increased system reliability. Therefore, the
proposed changes do not involve a significant reduction in a margin
of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: 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: Charles L. Miller.
Power Authority of the State of New York, Docket No. 50-333, James A.
FitzPatrick Nuclear Power Plant, Oswego County, New York
Date of amendment request: December 20, 1989, as supplemented
January 16, 1990, January 3, 1992, January 30, 1992, May 5, 1993, May
26, 1993, and March 2, 1994.
Description of amendment request: This application for an amendment
to the James A. FitzPatrick Technical Specifications proposes new
Safety/Relief Valve (SRV) performance limits to take credit for the
currently installed SRV capacity. Specifically, three changes to the
existing SRV performance limits are proposed:
The first permits continued plant operation with two SRVs
out-ofservice. Since 7 of the 11 SRVs at FitzPatrick are also automatic
depressurization system (ADS) valves, this reduces the number of ADS
valves required to be operable to 5. Current specifications permit only
one SRV out-of-service for 30 days.
Secondly, the setpoints for all 11 SRVs are changed to a
single nominal setpoint. Current specifications stagger the setpoints
from 1090 to 1140 psig.
The third change increases the maximum permissible
setpoint tolerance from one to three percent.
The new Limiting Safety System Setting (LSSS) for reactor coolant
system overpressurization protection (TS 2.2.1.B), as a result of these
changes, now requires that 9 of 11 SRVs be operable at a common
setpoint of 1110 psig plus or minus 3 percent.
Safety analyses were performed, using a conservative SRV setpoint
of 1195 psig, which demonstrate that these proposed changes are
acceptable.
Other changes, not associated with SRV performance, clarify
selected portions of the Technical Specifications and correct minor
typographical and editorial errors.
This ``Notice of Consideration of Issuance of Amendment to Facility
Operating License and Opportunity for Hearing'' (Notice) supersedes the
related Notice which was published in the Federal Register on May 15,
1990 (55 FR 20228).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Operation of the James A. FitzPatrick Nuclear Power Plant in
accordance with the proposed amendment would not involve a significant
hazards consideration as defined in 10 CFR 50.92, since it would not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated. A bounding
analysis (NEDC-31697P, ``Updated SRV Performance Requirements for
the James A. FitzPatrick Nuclear Power Plant'') of the revised SRV
performance requirements considered plant operation with 9 of 11
SRVs operable and with a common valve actuation pressure of 1195
psig. The analysis demonstrates that a 50 psi margin exists between
the maximum anticipated pressure and the American Society of
Mechanical Engineers (ASME) Code upset reactor vessel pressure limit
of 1375 psig. The analyses of NEDC-31697P also demonstrate that the
new SRV performance limits have no significant impact on thermal
limits, ECCS/LOCA performance, HPCI/RCIC operability, containment
response, containment integrity, or 10 CFR [Part] 50 Appendix R
alternate shutdown capability. The analyses also considered simmer
margin and downward setpoint drift.
The five miscellaneous changes clarify terminology, correct
typographical errors, remove a surveillance requirement which should
have been deleted as part of Amendment 130, clarify when SRV manual
actuation is performed, and delete a duplicate specification. These
changes are purely administrative in nature and, as such, do not
impact previously evaluated accidents or equipment malfunctions.
2. Ccreate the possibility of a new or different kind of
accident from those previously evaluated. The new SRV performance
limits are primarily administrative changes. The only physical
changes involve recalibration of SRV setpoints and operation with 2
SRVs/ADS valves out-of-service. The operation and function of the
pressure relief system and [are] unaffected. No new failure modes
are introduced.
The proposed miscellaneous changes are purely administrative in
nature and, as such, do not create the possibility of an accident or
malfunction.
3. Involve a significant reduction in the margin of safety. The
new SRV performance limits slightly reduce the existing margin to
vessel overpressure and the margin to the 125% mechanical overspeed
trip for the HPCI and RCIC turbines. However, the reduction in the
overpressure margin is insignificant (approximately 25 psi) and the
plant's response to transients and accidents remains well within the
limits established in General Design Criteria (GDC) 15, Standard
Review Plan Section 5.2.2, and FSAR Section 4.4. The reduction in
turbine overspeed margin is negligible (less than 1%), because it is
within the allowable tolerance of the trip settings.
The proposed miscellaneous changes are purely administrative in
nature and do not involve a reduction in safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Reference and Documents
Department, Penfield Library, State University of New York, Oswego, New
York 13126.
Attorney for licensee: Mr. Charles M. Pratt, 1633 Broadway, New
York, New York 10019.
NRC Project Director: Robert A. Capra
Power Authority of The State of New York, Docket No. 50-286, Indian
Point Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: April 18, 1994.
Description of amendment request: The proposed amendment would
relocate the fire protection requirements of Technical Specifications
(TSs) 3.14 and 4.12, and fire brigade staffing and training
requirements of TSs 6.2.2(f) and 6.4.2 from the TSs to
administratively-controlled operational specifications. Specifically,
the proposed changes would add the NRC standard fire protection license
condition to the Operating License, update the Final Safety Analysis
Report (FSAR) to include the Fire Protection Program by reference, and
relocate the fire protection requirements from the TSs to the Indian
Point 3 Operational Specifications Manual. The proposed changes have
been developed in accordance with the guidance contained in NRC Generic
Letter (GL) 86-10, ``Implementation of Fire Protection Requirements,''
and GL 88-12, ``Removal of Fire Protection Requirements from the
Technical Specifications.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Consistent with the criteria of 10 CFR 50.92, the enclosed
application is judged to involve no significant hazards based on the
following information:
(1) Does the proposed license amendment involve a significant
increase in the probability or consequences of any accident
previously evaluated?
Response
This change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
This proposed amendment merely relocates the fire protection
program elements from the Technical Specifications to the
Operational Specifications and the FSAR [Final Safety Analysis
Report]. No reduction in content is being made to the Technical
Specification requirements that are being relocated. Operating
limitations will continue to be imposed, and required surveillances
will continue to be performed in accordance with written procedures
and instructions auditable by the NRC.
Although future proposed changes to the fire protection program
elements previously located in the Technical Specifications will no
longer be controlled by 10 CFR 50.90, proposed changes to the Fire
Protection requirements relocated to the Operational Specifications
will be evaluated by plant administrative procedures.
Thus, programmatic controls will continue to assure that future
proposed fire protection program changes will not create an
unreviewed safety question.
(2) Does the proposed license amendment create the possibility
of a new or different kind of accident from any previously
evaluated?
Response
The possibility of an accident or malfunction of a different
type than evaluated previously in the safety analysis report is not
created.
This proposed amendment merely relocates the fire protection
Technical Specification requirements from the Technical
Specifications to the Operational Specifications. No reduction to
the fire protection Technical Specification requirements is being
made and thus the change does not create the possibility of a new or
different accident from those previously evaluated.
As noted above, future changes to the requirements in the
Operational Specifications will be evaluated by plant administrative
procedures.
(3) Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The margin of safety as defined in the bases for any technical
specification is not reduced.
This proposed amendment does not involve a reduction to the
approved fire protection program or Fire Protection Technical
Specification requirements. The Technical Specification fire
protection requirements are being relocated, with no reduction in
content, to the Operational Specifications. Since there is no
reduction in the requirements, there is no reduction in the margin
of safety.
As noted above, proposed changes to the Fire Protection
Technical Specification requirements relocated to the Operational
Specifications will be evaluated by plant administrative procedures.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10601.
Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle,
New York, New York 10019.
NRC Project Director: Robert A. Capra.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of amendment request: April 12, 1994.
Description of amendment request: This amendment request would
revise the Emergency Diesel Generator hot restart test by separating it
from the 24-hour endurance run and from the load sequence testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Do not involve a significant increase in the probability or
consequences of an accident previously analyzed.
The proposed changes would revise the Salem Emergency Diesel
Generator (EDG) surveillance criteria to allow the hot restart test
to be performed independent of the Engineered Safety Features (ESF)
load sequencing test and the 24-hour endurance run. The proposed
surveillance requirements would continue to demonstrate that the
objectives of each of these tests are met. Specifically, the EDG's
are shown to be capable of starting the ESF loads in the required
sequence, operating at full load for an extended period of time, and
restarting from a full load temperature condition. Therefore, the
proposed changes would not adversely affect the EDG's ability to
support mitigation of the consequences of any previously evaluated
accident. The proposed changes to the surveillance requirements do
not affect the initiation or progression of any accident sequence.
(2) Do not create the possibility of a new or different kind of
accident from any accident previously evaluated.
The proposed change affects surveillance test criteria such that
increased scheduling flexibility is allowed while the test
objectives associated with demonstrating EDG operability continue to
be met. The proposed changes do not allow any plant configurations
that are presently prohibited by the Salem Technical Specifications.
(3) Do not involve a significant reduction in a margin of
safety.
Surveillance testing per the proposed Technical Specifications
would continue to demonstrate the ability of the EDG's to perform
their intended function of providing electrical power to ESF systems
needed to mitigate design basis transients, consistent with the
plant safety analyses. The margin of safety demonstrated by the
plant safety analyses is therefore not affected by the proposed
change.
Therefore, [Public Service Electric and Gas Company] PSE&G has
concluded that the changes proposed herein 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: 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: Charles L. Miller.
Southern California Edison Company, et al., Docket No. 50-206, San
Onofre Nuclear Generating Station, Unit No. 1, San Diego County,
California
Date of amendment request: April 18, 1994.
Description of amendment request: The proposed amendment will
revise Sections 2.C and 2.D of the San Onofre Nuclear Generating
Station, Unit 1 (SONGS 1) Operating License. Section 2.C will be
revised to modify or delete several licensing conditions which either
no longer apply or require revision to apply to SONGS 1 in its
permanently shutdown and defueled condition. Section 2.D will be
revised to exempt Fire Protection reporting from the reporting
requirements of Section 2.D.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Will operation of the facility according to this proposed
change involve a significant increase in the probability or
consequences of an accident previously evaluated?
No. SONGS 1 has been permanently shut down and all fuel has been
taken out of the reactor and stored in the SONGS 1 spent fuel pool.
The proposed change will not modify any of the existing plant
configurations, controls, procedures, or technical specification
requirements necessary to assure the integrity and safe operation of
the spent fuel pool.
The technical basis for deleting the four license conditions,
which relate to Integrated Implementation Schedule, Cycle 11 Thermal
Shield Monitoring Program, Plant Modification to Eliminate Single
Failure Susceptibility of Vital Bus Automatic Transfer Function, and
the NRC's Confirmatory Order of January 2, 1990, is that these
license conditions were intended to assure the continued safe
operation of SONGS 1 as a power producing plant. With the permanent
shutdown of SONGS 1 and the issuance of its Permanently Defueled
Technical Specifications (PDTS) on December 28, 1993, the plant
modifications and safety programs associated with the four license
conditions are no longer necessary.
The technical basis for modifying the license condition on fuel
transshipment is that this license condition was intended to ensure
the safety of the operating plant by putting restrictions on
operation of the turbine building gantry crane. These restrictions
are no longer necessary, in light of the permanent shutdown of SONGS
1.
The technical basis for modifying the license condition on
physical protection is that this is necessary to update the
information contained in the license condition.
The technical basis for exempting the Fire Protection Program
from the reporting requirements of Section 2.D is that the
applicable requirements are adequately covered in 10 CFR 50.72 and
50.73, as stated in Generic Letters 86-10 and 88-12.
2. Will operation of the facility according to this proposed
change create the possibility of a new or different kind of accident
from any accident previously evaluated?
No. No safety-related equipment will be impacted by this
proposed change. Thus, there is no credible likelihood that a new or
different kind of accident from any accident previously evaluated
would occur as a result of this proposed change.
3. Will operation of the facility according to this proposed
change involve a significant reduction in a margin of safety?
No. As explained earlier, the plant modifications and safety
programs associated with the license conditions being deleted are no
longer necessary. The safety-related equipment concerns that led to
restrictions on operation of the turbine building gantry crane no
longer exist. The modification to the license condition on physical
protection will update the information contained in this license
condition.
The revision to Section 2.D will make the reporting requirements
regarding deficiencies in the Fire Protection Program consistent
with the NRC's generic guidance on this subject.
Thus operation of the facility in accordance with this proposed
change will not significantly reduce a margin of safety.
The NRC staff has reviewed the analysis of the licensee 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: Main Library, University of
California, P.O. Box 19557, Irvine, California 92713.
Attorney for licensee: James A. Beoletto, Esquire, Southern
California Edison Company, P.O. Box 800, Rosemead, California 91770.
NRC Project Director: Seymour H. Weiss.
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 amendment request: December 23, 1993 (TS346).
Description of amendment request: The proposed amendment would
revise the BFN Units 1, 2, and 3 Technical Specifications (TS) by
providing an alternate visual inspection schedule for safety-related
snubbers. The licensee has stated that the amendment follows the
recommendations of NRC Generic Letter (GL) 90-09, ``Alternative
Requirements for Snubber Visual Inspection Intervals and Corrective
Actions'' dated December 11, 1990. GL 90-09 describes a TS line item
improvement acceptable to the NRC staff. The purpose of the line item
improvement is to provide a means for reducing resource demands and
unnecessary occupational radiological exposure attributable to snubber
inspections while continuing to provide an acceptable level of
confidence in snubber operability.
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.
Implementing the guidance specified in GL 90-09 will not
introduce any new failure mode and will not alter any assumptions
previously made in evaluating the consequences of an accident. The
proposed alternate schedule for visual inspections will maintain the
same operability confidence level as the existing schedule. Also,
the surveillance requirement and schedule for snubber functional
testing remains the same providing a 95 percent confidence level
that 90 percent to 100 percent of the snubbers operate within the
specified acceptance limits. The proposed visual inspection schedule
is separate from functional testing and provides additional
confidence that the installed snubbers will serve their design
function and are being maintained operable. The proposed changes do
not affect limiting safety system settings or operating parameters,
and do not modify or add any accident initiating events or
parameters. Therefore, the proposed change does not significantly
increase 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.
Implementing the recommendations specified in GL 90-09 does not
involve any physical alterations to plant equipment, changes to
setpoints or operating parameters, nor does it involve any potential
accident initiating event. As stated in the generic letter, the
alternate schedule for snubber visual inspections maintains the same
confidence level as the existing schedule. Additionally, functional
testing of snubbers provides a 95 percent confidence level that 90
percent to 100 percent of the snubbers operate within specified
acceptance limits. Since this TS change does not physically alter
the plant equipment and the snubber confidence level remains the
same there will not be any new or different accident resulting from
snubber failure from any accident previously evaluated.
3. The proposed amendment does not involve a significant
reduction in a margin of safety.
The proposed change incorporates the surveillance requirements
for snubber visual inspection intervals following the guidance
provided in GL 90-09. As stated in the generic letter, the proposed
snubber visual inspection interval maintains the same confidence
level as the existing snubber visual inspection interval. This
surveillance requirement does not alter the current Limiting
Condition for Operation or the accompanying actions for the
snubber(s). The requirement for functional testing of safety-related
snubbers is unchanged and remains the basis for the established
margin of safety and assures a 95 percent confidence level that 90
percent to 100 percent of the snubbers operate within the specified
acceptance limits. This functional testing along with the proposed
visual inspection intervals provides adequate assurance that the
snubber will perform its intended function. Therefore, the proposed
changes do not involve a significant reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Athens Public Library, South
Street, Athens, Alabama 35611.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902.
NRC Project Director: Frederick J. Hebdon.
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: March 30, 1994.
Description of amendment request: The proposed amendment would
revise the TS 3/4.1.1.1 (Reactivity Control Systems--Boration Control
Systems--Boration Control--Shutdown Margin), TS 3/4.1.2.8 (Reactivity
Control Systems--Borated Water Sources--Shutdown), TS 3/4.1.2.9
(Reactivity Control Systems--Borated Water Sources--Operating), Bases
3/4.1.2 (Boration Systems), TS 3.4.5.1 (Emergency Core Cooling Systems,
ECCS--Core Cooling Tanks), TS 3/4.5.2 (ECCS--ECCS Subsystems), TS 3/
4.5.4 (ECCS--Borated Water Storage Tank), Bases 3/4.5 (ECCS), and TS 3/
4.10.4 (Special Test Exceptions--Shutdown Margin). This amendment
would: (a) Increase the required boration flowrate in the event the
required shutdown margin is not met, (b) increase the applicable
minimum boron concentration and/or volume requirements, (c) revise the
applicable Action statements and Surveillance Requirements, and (d)
propose several administrative and editorial changes.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below, indicating that the proposed
changes would:
1a. Not involve a significance increase in the probability of an
accident previously evaluated because no accident initiators,
conditions or assumptions are significantly affected by the proposed
changes.
The proposed changes would increase the required boration
flowrate in the event the required SHUTDOWN MARGIN is not met,
increase the minimum required volume for the Boric Acid Addition
System (BAAS) and increase the minimum required boron concentration
for the Borated Water Storage Tank (BWST) and the Core Flooding
Tanks (CFT). The proposed changes would also revise the Technical
Specification (TS) Action Statements for the BWST and the CFT,
revise the TS Surveillance Requirement relating to boron
concentration sampling of the CFT, and would revise the TS
Surveillance Requirements involving trisodium phosphate chemistry.
In addition, various administrative and editorial changes, including
changes to the TS Bases, are proposed. 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 proposed changes for the minimum required boron
concentrations and volumes for the BAAS, BWST, and CPT comply with
existing requirements to maintain a 1% delta k/k shutdown margin
(SDM) at all times, and are consistent with reload and LOCA
analysis. Therefore, the accident condition assumption of 1% delta
k/k SDM at the initiation of an accident will still be met and the
radiological consequences will be as previously evaluated.
The proposed changes do not alter the source term, containment
isolation, or allowable releases. The proposed changes, therefore,
will not increase the radiological consequences of a previously
evaluated accident.
2a. Not create the possibility of a new kind of accident from
any accident previously evaluated because no new accident initiators
or assumptions are introduced by the proposed changes. As stated in
1a, the proposed changes do not affect any accident initiators and
are not initiators themselves. The proposed changes do not alter any
accident scenarios.
2b. Not create the possibility of a different kind of accident
from any accident previously evaluated because the proposed changes
only affect existing components, systems, and functions and do not
introduce any new requirements that cannot be met with the existing
components, systems, and functions. The proposed changes do not
alter any accident scenarios.
3. Not involve a significant reduction in a margin of safety.
The proposed changes to the minimum required boron concentration and
volumes for the BAAS, BWST, and CFT would ensure the margin of
safety for reactor subcriticality is maintained at all times for
anticipated future core designs.
The proposed change to the TS Action statement to increase the
required boration flowrate in the event the SHUTDOWN MARGIN
requirement is not met, would ensure that the boration rate is
adequate for restoring the required SHUTDOWN MARGIN for anticipated
future core design.
The proposed changes to the TS Action statements for the BWST
and the CFT ensure that the plant is maneuvered in a timely and
conservative manner, without challenging any plant systems, while
minimizing the time the plant would be exposed to a LOCA with
assumptions not being met.
The proposed changes to the TS Surveillance Requirements
associated with trisodium phosphate chemistry would clarify the
requirements, make it easier to perform testing, minimize radwaste
generation, and reduce the consequences of a potential radioactive
spill. The proposed changes would also make the requirements
consistent with the DBNPS Updated Safety Analysis Report.
The proposed change to the TS Surveillance Requirement
associated with the boron concentration sampling of the CFT would
eliminate an unnecessary requirement and make the Surveillance
Requirement consistent with NUREG-1430.
None of these changes would adversely affect 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 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, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
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: April 5, 1994.
Description of amendment request: The proposed amendment would
revise the TS 3/4.7.1.2, Auxiliary Feedwater System, TS 3/4.7.1.7,
Motor Driven Feedwater Pump System, and their applicable Bases. This
amendment would: (a) Clarify the requirements for operation of the
Auxiliary Feedwater System and Motor Driven Feedwater Pump System, (b)
increase the surveillance intervals for testing the steam turbine
driven auxiliary feedwater pumps and the electric motor driven pump,
and (c) modify requirements relative to stationing an individual
locally, during associated surveillance testing.
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, indicating that the proposed
changes would:
1a. Not involve a significant increase in the probability of an
accident previously evaluated because no change is being made to any
accident initiator. The proposed changes are clarifications and the
incorporations of either the recommendations of Generic Letter 93-05
or the guidance provided by NUREG-1430. Therefore, it can be
concluded that the proposed changes do not involve a significant
increase in the probability of an accident previously evaluated.
1b. Not involve a significant increase in the consequences of an
accident previously evaluated because the proposed changes do not
invalidate accident conditions or assumptions used in evaluating the
radiological consequences of an accident.
2a. Not create the possibility of a new kind of accident from
any accident previously evaluated because the proposed changes do
not change the way the plant is operated. No new types of failures
or accident initiators are introduced by the proposed changes.
2b. Not create the possibility of a different kind of accident
from any accident previously evaluated because no new failure modes
have been defined for any plant system or component important to
safety, nor has any limiting single failure been identified as a
result of the proposed changes. No different accident initiators or
failure mechanisms are introduced by the proposed changes.
3. Not involve a significant reduction in a margin of safety
because the proposed changes continue to ensure the availability of
the Auxiliary Feedwater System and the Motor Driven Feedwater System
when called upon to perform their functions and will not adversely
impact any safety analysis assumptions.
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, NW., Washington, DC 20037.
NRC Project Director: John N. Hannon.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: April 15, 1994.
Description of amendment request: The proposed change would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed changes
would modify the pressure/temperature operating limitations during
heatup and cooldown and the Low Temperature Overpressure Protection
System (LTOPS) pressure setpoints and temperatures for NA-1&2. Also,
the proposed changes include revised Limiting Conditions for Operation,
Action Statements, and Surveillance Requirements for the Power-Operated
Relief Valves (PORVs) and block valves to address the concerns
discussed in NRC Generic Letter 90-06. Additionally, the proposed
changes include several editorial/administrative changes.
The NA-1&2 Reactor Coolant Systems (RCS) are protected from
material failure by the imposition of restrictions on allowable
pressure and temperature, and on heatup and cooldown rate. The LTOPS
ensures that material integrity limits are not exceeded during the
design basis overpressurization accidents. Equipment operability
requirements are imposed to ensure that the assumptions of the accident
analyses remain valid. The operating restrictions, setpoints, and
equipment operability requirements must be revised to extend their
applicability to a higher cumulative burnup, and to improve operational
flexibility.
The current pressure/temperature operating limits and LTOPS
setpoints are valid to 12 Effective Full-Power Years (EFPY) and 17 EFPY
for NA-1&2, respectively. According to the most recent estimates, the
burnup applicability limits will be exceeded by NA-1 in the spring of
1996. The NA-2 pressure/temperature operating limits and LTOPS
setpoints remain valid well into the year 2002. The proposed NA-1 TS
include revised pressure/temperature operating limits valid to end-of-
license. Although the NA-2 pressure/temperature operating limits are
not being changed, the NA-2 LTOPS setpoints and associated reactor
vessel integrity protection philosophy are being changed. The reactor
vessel integrity protection philosophy which supports the proposed TS
changes provides improved operational flexibility while maintaining an
adequate margin of safety as demonstrated by the safety analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of [North Anna] Power Station in accordance
with the [proposed] Technical Specification changes will not:
[1] involve a significant increase in the probability or
consequences of an accident previously evaluated. The safety
analysis demonstrates that the proposed reactor vessel protection
philosophy, and the associated pressure/temperature limits, LTOPS
setpoints, and component operability requirements, ensure that
reactor vessel integrity will be maintained during normaloperation
and design basis accident conditions. Specifically, adherence to the
heatup/cooldown rate-dependent pressure/temperature operating limits
ensures that the assumed design basis flaw will not propagate during
normal operation. Below the LTOPS enabling temperature, automatic
actuation of the PORVs ensures that the assumed design basis flaw
will not propagate under design basis low-temperature
overpressurization accident conditions. Two pressurizer safety
valve[s] are sufficient to relieve the overpressurization due to the
inadvertent startup of two charging pumps at water solid conditions
without propagation of the assumed design basis flaw. The proposed
changes to address the concerns of Generic Letter 90-06 (Generic
Issues 70 and 94) improve LTOPS availability and reliability by
instituting requirements for PORV, block valve, and control system
testing and allowed outage times for these components. Although
these changes do not reduce the probability of occurrence or the
consequences of the LTOPS design basis (mass and heat addition)
transients, the changes provide increased assurance that pressure
relieving devices will perform their design function when required.
[2] create the possibility of a new or different kind of
accident from any accident previously evaluated. The proposed
Technical Specifications modify pressure/temperature operating
limits, LTOPS setpoints and enabling temperatures, and component
operability requirements. The revised pressure/temperature operating
limits, and LTOPS setpoints and enabling temperatures are only
slightly different than those currently in the Technical
Specifications. No operating limits or setpoints are added or
deleted by the proposed changes. Therefore, it may be concluded that
the operating limits and setpoint changes do not create the
possibility of a new or different kind of accident. With regard to
component operability requirements, restrictions on the number of
charging pumps which may be operable, the number of PORVs which must
be operable, and the allowable temperature difference between the
steam generator primary and secondary remain unchanged. Only the
setpoint temperature at which these restrictions apply have been
modified. The proposed changes are entirely consistent with the
reactor vessel integrity protection philosophy which ensures that
the design basis reactor vessel flaw will not propagate under normal
operation or postulated accident conditions. Further, the proposed
changes do not invalidate . . . any component design criteria or the
assumptions of any UFSAR [Updated Final Safety Analysis Report]
Chapter 15 accident analyses. In addition, modifications have been
made to the Technical Specifications to improve availability and
reliability of PORVs and associated block valves. These changes have
been made in accordance with NRC guidance in Generic Letter 90-06.
It may be concluded that none of the proposed changes creates the
possibility of a new or different kind of accident from any
previously evaluated.
[3] involve a significant reduction in a margin of safety. As
described above, the reactor vessel integrity protection philosophy
ensures that the design basis assumed flaw will not propagate under
normal operation or design basis accident conditions. Adherence to
the Technical Specification pressure/temperature operating limits
ensures that the margin to vessel fracture provided by the ASME
Section XI methodology is maintained. With regard to LTOPS
protection, the safety analysis demonstrates that the proposed LTOPS
design ensures margins consistent with those provided by ASME
Section XI Appendix G methods. This conclusion is based on industry
experience with LTOPS events and engineering evaluation.
Specifically, both industry experience and engineering evaluation
demonstrate that LTOPS design basis events may be expected to occur
at essentially isothermal conditions. Engineering evaluation
demonstrates that any reduction in allowable pressure due to thermal
stresses which may be expected to occur during low temperature
operation is insignificant when compared to margins provided by the
ASME Section XI Appendix G methods for calculating pressure/
temperature operating limits. Use of the isothermal pressure/
temperature limit curve as the design limit for establishing low
temperature PORV lift setpoints has been approved for other
utilities by the NRC. This design maximizes the operating margin
above the minimum RCS pressure for reactor coolant pump (RCP)
operation, thereby minimizing the probability of undesired PORV
lifts during RCP startup. The proposed changes to address the
concerns of Generic Letter 90-06 (Generic Issues 70 and 94) improve
LTOPS availability and reliability by instituting requirements for
PORV, block valve, and control system testing and allowed outage
times for these components. Although these changes do not increase
the margin of safety demonstrated by the analysis of the LTOPS
design basis (mass and heat addition) transients, the changes
provide increased assurance that pressure relieving devices will
perform their design function when required.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: April 19, 1994.
Description of amendment request: The proposed changes would revise
the Technical Specifications (TS) for the North Anna Power Station,
Units No. 1 and No. 2 (NA-1&2). Specifically, the proposed changes
would modify the surveillance frequency of the control rod motion
testing from monthly to quarterly in accordance with NRC Generic Letter
(GL) 93-05, ``Line-Item Technical Specifications Improvements to Reduce
Surveillance Requirements for Testing During Power Operation'' dated
September 27, 1993.
The proposed changes to the surveillance requirements for the
control rods at NA-1&2 are consistent with the intent of GL 93-05,
which is to improve safety, decrease equipment degradation, and reduce
unnecessary burden on personnel resources by reducing testing
requirements that are marginal to safety.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of North Anna Power Station in accordance
with the proposed Technical Specifications changes will not:
1. Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
The proposed change to the surveillance frequency for control
rods does not increase the probability of an accident occurrence.
Surveillance testing is a means of determining control rod
operability and does not of itself contribute to control rod
inoperability. Although reduced testing also implies a less frequent
confirmation of mechanical operability, operational experience has
established that the reduced testing does not decrease plant safety.
Furthermore, reduced frequency testing reduces the probability of an
inadvertent operational transient or misaligned control rod. There
are other means available (e.g., Individual Rod Position Indicators,
flux distributions anomalies) to detect a misaligned control rod.
Reducing the frequency of surveillance testing will decrease the
possibility of finding an inoperable control rod. Industry
experience has shown that most inoperable (stuck) control rods are
identified during rod drop testing and unit startup after refueling
outages. Therefore, the NRC has determined that a reduced frequency
surveillance test during power is acceptable to determine control
rod operability (trippable).
The control rods will continue to be operated in the same manner
during the surveillance testing and will be available to shutdown
the reactor if a Reactor Protection System trip setpoint is reached.
The operability requirements, alignment and insertion limits for the
control rods remain unchanged. Since the control rods remain
available (trippable) to perform their intended safety function,
testing of the control rods at the proposed reduced frequency will
not increase the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed reduced frequency testing of the control rods does
not change the way the Control Rod Drive System or the control rods
are operated. The reduced frequency of testing of the control rods
does not alter the operation of the Control Rod Drive System or the
control rods ability to perform their intended safety function.
Therefore, the reduced frequency testing of the control rods does
not generate any new accident precursors. In fact, industry
experience has shown that this surveillance testing may result in
inadvertent reactor trips, dropped control rods, or unnecessary
challenges to safety systems. Therefore, the possibility of a new or
different kind of accident than previously evaluated is not created
by the proposed changes in surveillance frequency of the control
rods.
3. Involve a significant reduction in a margin of safety.
The proposed reduced frequency testing of the control rods does
not change the control rod operability requirement or the way the
Control Rod Drive System is operated. NUREG-1366, concluded that
most stuck control rods are discovered during plant startup after
refueling or during control rod drop testing. Therefore, routine
surveillance testing of the control rods at the proposed reduced
frequency is considered adequate to identify inoperable (stuck)
control rods during operation. The reduced surveillance requirements
do not affect the margin of safety in that the operability
requirements remained unchanged and the existing safety analysis,
which assumes the most reactive control rod sticks out of the core
during accident scenarios, remains bounding. Therefore, no margins
of safety are adversely affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: The Alderman Library, Special
Collections Department, University of Virginia, Charlottesville,
Virginia 22903-2498.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow.
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: April 19, 1994.
Description of amendment request: The proposed changes will modify
the surveillance frequency of the control rod motion testing from
monthly to quarterly.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
Specifically, operation of Surry Power Station in accordance with
the proposed Technical Specifications changes will not:
1. Involve a significant increase in the probability of
occurrence or consequences of an accident previously evaluated.
The proposed change to the surveillance frequency for control
rods does not increase the probability of an accident occurrence.
Surveillance testing is a means of determining control rod
operability and does not of itself contribute to control rod
inoperability. Although reduced testing also implies a less frequent
confirmation of mechanical operability, operational experience has
established that the reduced testing does not decrease plant safety.
Furthermore, reduced frequency testing reduces the probability of an
inadvertent operational transient or misaligned control rod. There
are other means available (e.g., Individual Rod Position Indicators,
flux distributions anomalies) to detect a misaligned control rod.
Reducing the frequency of surveillance testing will decrease the
possibility of finding an inoperable control rod. Industry
experience has shown that most inoperable (stuck) control rods are
identified during rod drop testing and unit startup after refueling
outages. Therefore, the NRC has determined that a reduced frequency
surveillance test during power is acceptable to determine control
rod operability (trippable).
The control rods will continue to be operated in the same manner
during the surveillance testing and will be available to shutdown
the reactor if a Reactor Protection System trip setpoint is reached.
The operability requirements, alignment and insertion limits for the
control rods remain unchanged. Since the control rods remain
available (trippable) to perform their intended safety function,
testing of the control rods at the proposed reduced frequency will
not increase the consequences of an accident previously evaluated.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed reduced frequency testing of the control rods does
not change the way the Control Rod Drive System or the control rods
are operated. The reduced frequency of testing of the control rods
does not alter the operation of the Control Rod Drive System or the
control rods ability to perform their intended safety function.
Therefore, the reduced frequency testing of the control rods does
not generate any new accident precursors. In fact, industry
experience has shown that this surveillance testing may result in
inadvertent reactor trips, dropped control rods, or unnecessary
challenges to safety systems. Therefore, the possibility of a new or
different kind of accident than previously evaluated is not created
by the proposed changes in surveillance frequency of the control
rods.
3. Involve a significant reduction in a margin of safety.
The proposed reduced frequency testing of the control rods does
not change the control rod operability requirement or the way the
Control Rod Drive System is operated. NUREG-1366, concluded that
most stuck control rods are discovered during plant startup after
refueling or during control rod drop testing. Therefore, routine
surveillance testing of the control rods at the proposed reduced
frequency is considered adequate to identify inoperable (stuck)
control rods during operation. The reduced surveillance requirements
do not affect the margin of safety in that the operability
requirements remained unchanged and the existing safety analysis,
which assumes the most reactive control rod sticks out of the core
during accident scenarios, remains bounding. Therefore, no margins
of safety are adversely affected.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Local Public Document Room location: Swem Library, College of
William and Mary, Williamsburg, Virginia 23185.
Attorney for licensee: Michael W. Maupin, Esq., Hunton and
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond,
Virginia 23219.
NRC Project Director: Herbert N. Berkow.
Previously Published Notices of Consideration of Issuance of
Amendments to Facility Operating Licenses, Proposed no Signficant
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.
Northeast Nuclear Energy Company, et al., Docket No. 50-336, Millstone
Nuclear Power Station, Unit No. 2, New London County, Connecticut Date
of amendment request: April 14, 1994, as supplemented April 20, 1994.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TS) to change the laboratory
testing protocol for the charcoal absorbers for the Control Room
Emergency Ventilation System (TS 3.7.6.1) and the Enclosure Building
Filtration System (TS 3.6.5.1).
Date of publication of individual notice in Federal Register: May
4, 1994 (59 FR 23085).
Expiration date of individual notice: June 4, 1994.
Local Public Document Room location: Learning Resource Center,
Three Rivers Community-Technical College, Thames Valley Campus, 574 New
London Turnpike, Norwich, Connecticut 06360.
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 20555, and at the local public document
rooms for the particular facilities involved.
Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power
Station, Plymouth County, Massachusetts
Date of application for amendment: October 19, 1993.
Brief description of amendment: This amendment removes the low
condenser vacuum scram and reduces the turbine first stage setpoint at
which it is permissible to bypass the turbine control valve fast
closure and the turbine stop valve closure trip (scram) signals.
Date of issuance: May 5, 1994.
Effective date: May 5, 1994.
Amendment No.: 152.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 8, 1993 (58 FR
64603). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 5, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Plymouth Public Library, 11
North Street, Plymouth, Massachusetts 02360.
Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324,
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North
Carolina
Date of application for amendments: April 13, 1993.
Brief Description of amendments: The amendments change the
Technical Specifications to revise the design features information
pertaining to the elevation at which the spent fuel storage pool is
designed to prevent inadvertent draining. The amendments revise this
elevation from 116 feet 4 inches to 15 feet 11 inches based on the
actual spent fuel pool design.
Date of issuance: May 2, 1994.
Effective date: May 2, 1994.
Amendment Nos.: 170 and 201.
Facility Operating License Nos. DPR-71 and DPR-62. Amendments
revise the Technical Specifications.
Date of initial notice in Federal Register: March 16, 1994 (59 FR
12359). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 2, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of North Carolina
at Wilmington, William Madison Randall Library, 601 S. College Road,
Wilmington, North Carolina 28403-3297.
Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: May 6, 1993.
Brief description of amendments: The amendments correct an error in
Technical Specification Table 3.3-2 that was made with License
Amendments 128 and 110.
Date of issuance: May 11, 1994.
Effective date: May 11, 1994.
Amendment Nos.: 142 and 124.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 4, 1993 (58 FR
41503). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 11, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Atkins Library, University of
North Carolina, Charlotte (UNCC Station), North Carolina 28223.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: September 16, 1993.
Brief description of amendment: The amendment changed the Appendix
A Technical Specifications for the ultimate heat sink (UHS) to clarify
the requirements for the wet cooling tower fan covers, increased the
test interval for starting the dry and wet tower fans from 7 days to 31
days, increased the wet bulb temperature to 80 degrees F for
determining Operability, and made other editorial and clarifying
changes.
Date of issuance: May 9, 1994.
Effective date: May 9, 1994.
Amendment No.: 95.
Facility Operating License No. NPF-38. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 27, 1993 (58 FR
57851). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 9, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: University of New Orleans
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.
Georgia Power Company, Oglethorpe Power Corporation, Municipal Electric
Authority of Georgia, City of Dalton, Georgia, Docket Nos. 50-424 and
50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: November 19, 1993, as revised
March 31, 1994.
Brief description of amendments: The amendments revise surveillance
requirements for station batteries based on draft IEEE Standard 450-
1992, ``Recommended Practice for Maintenance, Testing, and Replacement
of Large Lead Storage Batteries for Generating Stations and
Substations.''
Date of issuance: May 2, 1994.
Effective date: May 2, 1994.
Amendment Nos.: 71/50.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 22, 1993 (58
FR 67847).
The March 31, 1994, letter, changed the initial request to provide
increased conformance to an associated draft IEEE Standard 450
maintenance and testing practice. The revision imposes restrictions on
cell replacements for degraded batteries that are in late stages of
service life. These restrictions were requested by the NRC staff and do
not affect the NRC staff's conclusions of no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated May 2, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Burke County Library, 412
Fourth Street, Waynesboro, Georgia 30830.
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: March 14, 1994.
Brief description of amendments: The amendments change the
technical specifications by adding a new Limiting Condition for
Operation (LCO), 3.0.6. LCO 3.0.6 will allow equipment removed from
service or declared inoperable to comply with actions to be returned to
service, under administrative controls, solely to perform testing. The
new LCO will provide temporary relief from the applicable action
statements to perform surveillance testing required to demonstrate
operability of the equipment being returned to service or the
operability of other equipment.
Date of issuance: April 29, 1994.
Effective date: April 29, 1994 to be implemented within 31 days of
issuance.
Amendment Nos.: Unit 1--Amendment No. 60; Unit 2--Amendment No. 49.
Facility Operating License Nos. NPF-76 and NPF-80. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 30, 1994 (58 FR
14889). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated April 29, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Wharton County Junior College,
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
Iowa Electric Light and Power Company, Docket No. 50-331, Duane Arnold
Energy, Center, Linn County, Iowa
Date of application for amendment: March 24, 1993.
Brief description of amendment: The amendment revised the Technical
Specifications by improving organization and clarity of Section 3.8/
4.8. The amendment changes the testing requirements of the operable
emergency diesel generator in Section 4.5.G.1 when the other diesel is
inoperable. Also, the testing requirements of the Emergency Service
Water pump and loop changed when the other pump or loop is inoperable.
The amendment also makes several editorial changes.
Date of issuance: May 12, 1994.
Effective date: May 12, 1994.
Amendment No.: 197.
Facility Operating License No. DPR-49. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 21, 1993 (59 FR
39051) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 12, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Cedar Rapids Public Library,
500 First Street, SE., Cedar Rapids, Iowa 52401.
North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook
Station, Unit No. 1, Rockingham, New Hampshire
Date of amendment request: September 13, 1993.
Description of amendment request: This amendment revises the
Appendix A Technical Specifications relating to certain sensor errors
stated in Table 2.2-1, Reactor Trip System Instrumentation Trip
Setpoints. The sensor errors specified for the Power Range, Neutron
Flux High Setpoint (Functional Unit 2. a.) and the Power Range, Neutron
Flux Low Setpoint (Functional Unit 2. b.) are changed to incorporate
the Nuclear Instrumentation System cabinet percent-full-power meter
accuracy and readout error.
Date of issuance: May 9, 1994.
Effective date: As of the date of issuance, to be implemented
within 60 days of issuance.
Amendment No.: 31.
Facility Operating License No. NPF-86. Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 13, 1993 (58 FR
52991). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 9, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Exeter Public Library, 47
Front Street, Exeter, New Hampshire 03833.
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: March 24, 1994.
Brief description of amendment: The Technical Specifications
amendment revised the plant staff requirement (specified in TS Section
6.2.2.i) to temporarily allow the Operations Manager to have held a
senior reactor operator (SRO) license at a pressurized water reactor
other than Indian Point 3. This temporary allowance is in effect for
the period ending 3 years after restart from the 1993/1994 Performance
Improvement Outage and is needed to support management changes at the
facility in an effort to improve overall performance.
Date of issuance: May 3, 1994.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment No.: 147.
Facility Operating License No. DPR-64: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 1, 1994 (59 FR
15464) The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated May 3, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: White Plains Public Library,
100 Martine Avenue, White Plains, New York 10610.
Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-311,
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New
Jersey
Date of application for amendments: April 28, 1993, as supplemented
by letters dated August 12, 1993, November 17, 1993, February 2, 1994,
and April 7, 1994.
Brief description of amendments: These amendments increase the
spent fuel pool capacities for Salem, Units 1 and 2 from the current
1170 fuel assemblies to 1632 fuel assemblies. Also, the decay time for
refueling operations is extended from 100 hours to 168 hours.
Date of issuance: May 4, 1994.
Effective date: May 4, 1994.
Amendment Nos. 151 and 131.
Facility Operating License Nos. DPR-70 and DPR-75. These amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 4, 1994 (59 FR
10440) The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated May 4, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Salem Free Public Library, 112
West Broadway, Salem, New Jersey 08079.
Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301 Point
Beach Nuclear Plant, Unit Nos. 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: February 26, 1993, as
supplemented on November 30, 1993, and February 8, 1994.
Brief description of amendments: These amendments revise Technical
Specifications (TS) Section 15.3.7, Section 15.4.6, and Table 15.4.1-2.
The revisions incorporate items that were identified during a
comparison of the accident analyses in the PBNP Safety Analysis Report
(FSAR) and the Limiting Conditions for Operation and surveillance
sections of the PBNP TS. The changes add systems or equipment required
by the accident analyses. Testing requirements for the diesel
generators are also revised to eliminate the daily testing requirement
when one diesel generator is inoperable.
Date of issuance: May 11, 1994.
Effective date: May 11, 1994.
Amendment Nos.: 148 and 152.
Facility Operating License Nos. DPR-24 and DPR-27. Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 18, 1993 (58 FR
43939) The November 30, 1993, and February 8, 1994, submittal provided
additional supplemental 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 May 11, 1994.
No significant hazards consideration comments received: No.
Local Public Document Room location: Joseph P. Mann Library, 1516
Sixteenth Street, Two Rivers, Wisconsin 54241.
Dated at Rockville, Maryland, this 18th day of May 1994.
For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II Office of Nuclear Reactor
Regulation.
[FR Doc. 94-12614 Filed 5-24-94; 8:45 am]
BILLING CODE 7590-01-P
