[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