[Federal Register Volume 59, Number 51 (Wednesday, March 16, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10316]


[[Page Unknown]]

[Federal Register: March 16, 1994]


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UNITED STATES NUCLEAR REGULATORY COMMISSION

 

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 18, 1994, through March 4, 1994. 
The last biweekly notice was published on March 2, 1994 (59 FR 9999).

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 P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 April 15, 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.

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, Unit 
Nos. 1, 2, and 3, Maricopa County, Arizona

    Date of amendment requests: January 20, 1994
    Description of amendment requests: The proposed amendment would 
change the departure from nucleate boiling ratio (DNBR) in Safety 
Limits, Section 2.1.1.1, and the associated Bases, as well as the DNBR 
- Low Trip Setpoint in Table 2.2-1, and the associated Bases, from a 
value of 1.24 to 1.30. In addition, the amendment would add a 
methodology supplement entitled,System 80GT1TMInlet 
Flow Distribution,'' to the list of methods used to determine the core 
operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1 -- Involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The purpose of the proposed TS amendment is to provide a revised 
DNBR Safety Limit and Low DNBR Trip Setpoint to ensure that no 
anticipated operational occurrence or postulated accident will 
result in core conditions exceeding DNBR Safety Limit.
    The change in the DNBR Safety Limit from 1.24 to 1.30 can be 
accommodated directly by increasing the limit (including the DNBR 
Trip Setpoint) or by an increase in the DNBR overall uncertainty 
factors for core operating limit supervisory system (COLSS) (EPOL2 
and EPOL4) and core protection calculator (CPC) (BERR1). Using the 
1.24 DNBR Safety Limit will result in larger uncertainty factors, 
and conversely using the increased DNBR Safety Limit of 1.30 will 
result in lower uncertainty factors. Therefore, plant operation for 
COLSS and CPC are not significantly affected by the choice of the 
DNBR Safety Limit and the Trip Setpoint as long as the corresponding 
overall uncertainty factors are calculated and implemented. PVNGS 
will implement the 1.30 DNBR Safety Limit and its corresponding 
overall uncertainty factors in the reload safety analysis for Unit 3 
Cycle 5 and in subsequent reload safety analyses for Units 1 and 2.
    The proposed amendment changes only the DNBR Safety Limit and 
associated Trip Setpoint, and does not in any way impact the 
operation of the plant. Safety and setpoint analyses will be 
performed consistent with the increased DNBR limit of 1.30. The core 
power distribution during all phases of normal and anticipated 
operational occurrences will remain bounded by the initial 
conditions assumed in Chapter 15 of the PVNGS Updated Safety 
Analysis Report (UFSAR). Furthermore, the UFSAR Chapter 15 analysis 
remains bounding because the margins of safety will be maintained. 
Therefore, the proposed change to Sections 2.1.1.1 and 2.2.1 (Table 
2.2-1) will not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to Section 6.9.1.10 does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. The proposed change is administrative 
in nature and does not involve any change to the configuration or 
method of operation of any plant equipment that is used to mitigate 
the consequences of an accident. Also, the proposed change does not 
alter the conditions or assumptions in any of the UFSAR accident 
analyses. Since the FSAR accident analyses remain bounding, the 
radiological consequences previously evaluated are not adversely 
affected by the proposed change. Therefore, it can be concluded that 
the proposed change to Section 6.9.1.10 will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Standard 2 -- Create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed amendment is limited to changing the DNBR Safety 
Limit and Low DNBR Trip Setpoint and does not involve any physical 
change to plant systems or to the COLSS and the CPC algorithms. 
These changes will not affect any safety-related equipment used in 
the mitigation of anticipated operational occurrences or design 
basis accidents. Therefore, this change to Section 2.1.1.1 and 2.2.1 
(Table 2.2-1) will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change to Section 6.9.1.10 does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated. The proposed change is administrative in 
nature and does not involve any change to the configuration or 
method of operation of any plant equipment that is used to mitigate 
the consequences of an accident. Accordingly, no new failure modes 
have been defined for any plant system or component important to 
safety nor has any new limiting failure been identified as a result 
of the proposed change. Therefore, it can be concluded that the 
proposed change to Section 6.9.1.10 will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    The DNBR Safety Limit specified in TS 2.1.1.1 and the Low DNBR 
Trip Setpoint specified in TS 2.2.1 (Table 2.2-1) ensure that 
operation of the reactor is prevented from exceeding the DNBR Safety 
Limit during normal operation and design basis anticipated 
operational occurrences. Therefore, operating within the increased 
DNBR Safety Limit will ensure that no anticipated operational 
occurrence or postulated accident will result in core conditions 
exceeding the specified DNBR Safety Limit. The UFSAR Chapter 15 
analysis remains bounding because the margins of safety will be 
maintained. Additionally, the COLSS and the CPC overall uncertainty 
factors will be calculated and implemented consistent with the 
increased DNBR Safety Limit of 1.30. Therefore, this change to 
Section 2.1.1.1 and 2.2.1 (Table 2.2-1) will not result in a 
significant reduction in a margin of safety.
    The proposed change to Section 6.9.1.10 does not involve a 
significant reduction in a margin of safety. The proposed change is 
administrative in nature and does not adversely impact the plant's 
ability to meet applicable regulatory requirements. Therefore, it 
can be concluded that the proposed change to Section 6.9.1.10 does 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004Attorney for licensees: Nancy 
C. Loftin, Esq., Corporate Secretary and Counsel, Arizona Public 
Service Company, P.O. Box 53999, Mail Station 9068, Phoenix, Arizona 
85072-3999
    NRC Project Director: Theodore R. Quay

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: January 9, 1994
    Description of amendment request: The proposed amendment would make 
changes to the Technical Specifications and License. These changes 
consist of revised wording for the license, clarify wording to aid 
operators in selecting the correct pressure/temperature curve during 
startup and shutdown operations, and removal of certain obsolete 
mechanical snubber acceptance criteria.
    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 operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The first proposed change will modify License DPR-35 to 
eliminate the need to issue a new page 3 to identify the latest 
amendment number. The second change will provide the correction of 
an error of omitting the reference to the subcritical mode of 
operation, in relation to the pressure/temperature curves. The third 
change will remove the unnecessary mechanical snubber functional 
test acceptance criterion to determine if drag force has increased 
more than 50% since the last functional test.
    Modification of License DPR-35 for Pilgrim Nuclear Power Station 
to remove the need to update page 3 whenever a new amendment is 
approved will reduce an administrative burden. This license change 
also precludes a possible administrative error if the correct 
reference is somehow missed. This change does not affect plant 
operation or design and is considered an administrative change and 
as such does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The second change corrects an error of omission made in an 
earlier amendment by inserting a reference to the subcritical 
reactor operation phase. This proposal will enhance the procedure 
changes and training already accomplished as short term corrective 
actions. This change does not affect plant operation or design and 
is considered an administrative change and therefore does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The third change removes an acceptance criterion for mechanical 
snubber testing not required by the ASME Boiler and Pressure Vessel 
Code, Section XI, Subsection IWF nor recommended by the vendor for 
mechanical snubbers in use at Pilgrim.
    This change will not result in any physical modification to 
Pilgrim. The mechanical snubbers will continue to be tested in 
accordance with existing plant procedures which reference the ASME 
Code Section XI, Subsection IWF. Therefore, this is considered an 
administrative change and as such, operation of Pilgrim will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident than previously evaluated because they 
are administrative in nature and require no physical alterations of 
plant configuration or changes to setpoints or operating parameters.
    3. The operation of Pilgrim in accordance with the proposed 
amendment will not involve a significant reduction in a margin of 
safety.
    Because these changes do not alter plant operation or design and 
are considered administrative in nature, they do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of amendments request: April 13, 1993
    Description of amendments request: The proposed amendments would 
revise design features information pertaining to the elevation at which 
the spent fuel pool is designed to prevent inadvertent draining. The 
proposed amendment would revise this elevation from 116 feet 4 inches 
to ll5 feet 11 inches based on the actual spent fuel pool design.
    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 amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The current value in Specification 5.6.2 is incorrect. No basis 
can be determined for including this value in Specification 5.6.2 
other than incorrectly characterizing the normal fuel pool water 
level as the design level to be maintained to prevent inadvertent 
draining of the fuel pool. This value was incorrectly incorporated 
into the initial standard Brunswick Technical Specifications, Water 
Level- Spent Fuel Storage Pool.
    The 115' 11'' elevation is equal to 20' 10-7/8'' above the top 
of the spent fuel rods seated in the storage racks. This level is 
still in excess of the minimum level required (20' 6'') by Technical 
Specification 3.9.9.
    The accident discussed in UFSAR [Updated Final Safety Analysis 
Report] Section 9.1.2.3.2.4.2, Loss of Spent Fuel Pool Cooling, is 
not impacted by this change since the spent fuel pool safety 
functions are not impacted and Technical Specification minimum fuel 
pool levels (Specification 3.9.9) are not changed. As such, the 
proposed amendments do not involve a significant increase in the 
probability of an accident previously evaluated.
    The radiological consequences of this accident are discussed in 
UFSAR Section 9.1.2.3.2.5. This analysis assumes spent fuel pool 
boiling. In addition, the facilities description of the spent fuel 
storage pool (Section 9.1.2.2.1), states that the surface of the 
water will be maintained at Elevation 116.3 ft, which is the normal 
water level of the pool. Therefore, the (lower) designed level to 
prevent inadvertent pool draining is not relevant within this 
analysis. As such, the proposed amendments do not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    2. The proposed amendments do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This amendment request corrects a mischaracterization of the 
design features and does not involve a change in fuel pool 
operations. Specification 5.6.2 states that the fuel pool is 
designed to prevent inadvertent draining of the pool below elevation 
116'4''. However, it is possible that the pool could drain below 
this level by draining through piping connected to the pool coupled 
with no flow into the pool. It is not possible, however, with the 
fuel pool gates installed, that the fuel pool could be inadvertently 
drained below the bottom of the pool overflows to the skimmer surge 
tanks. The elevation at the bottom of the overflows to the skimmer 
surge tanks is 115'11''. Therefore, this is the correct value to 
cite in the design features section of Technical Specifications.
    The proposed 115'11'' elevation will not result in new drain 
pathways, nor will the minimum fuel pool water level required by the 
Technical Specifications be impacted by this change. Therefore, the 
proposed amendments do not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. The proposed amendments do not involve a significant 
reduction in the margin of safety.
    The proposed amendments do not change safety limits, setpoints, 
or plant operations. The plant is actually designed to prevent 
inadvertent draining of the fuel pool below elevation 115'11'' as 
discussed above. This change is not an actual design change; it is a 
design clarification correcting the level at which inadvertent 
draining of the spent fuel pool is prevented. As such, the proposed 
amendments do not involve a significant reduction in the margin of 
safety at Brunswick.
    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: S. Singh Bajwa

Entergy Operations Inc., Docket No. 50-382, Waterford Steam 
Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: February 14, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) in response to Generic Letter 
93-08 issued by the NRC and dated December 29, 1993, by relocating the 
reactor trip system (RTS) and engineered safety feature activation 
system (EFAS) response time limits to the updated Final Safety Analysis 
Report (FSAR).
    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 are administrative in nature and do not 
involve any change to the configuration or method of operation of 
any plant equipment used to mitigate the consequences of an 
accident.
    The proposed changes do not alter the conditions or assumptions 
in any accident previously evaluated.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The proposed changes are administrative in nature and do not 
involve any change to the configuration or method of operation of 
any plant equipment used to mitigate the consequences of an 
accident. No new
    accident initiators or failure modes are created by relocating 
the RTS and ESFAS instrumentation response time limits.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The proposed changes are administrative in nature and will in no 
way affect the TS adequacy in ensuring the response times for the 
RTS and ESFAS instrumentation do not exceed the limits assumed in 
the accident analyses. The proposed changes will have no impact on 
the protective boundaries, safety limits, or margin of safety.
    Therefore, the proposed changes will not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: July 22, 1993, as supplemented February 
4, 1994
    Description of amendment request: The proposed amendment would 
modify the Facility Operating License (OL) and Technical Specifications 
(TSs) to permit uprated power operation. The plant is currently 
licensed for operation at 3323 megawatts thermal (MWt), although many 
of the original analyses were performed at a design power level of 3467 
MWt. The proposed changes would redefine rated thermal power to be 3467 
MWt, which represents an approximately 4.3 percent increase over the 
currently licensed power level. Implementation of the power uprate 
would require minor modifications, such as, resetting of the low set 
safety relief setpoints, as well as the recalibration of plant 
instrumentation to reflect the uprated power. The proposed changes 
follow the generic guidelines for boiling water reactor power uprate 
described in General Electric Topical Report, NEDC-31897P-1, ``Generic 
Guidelines for General Electric Boiling Water Reactor Power Uprate,'' 
June 1991.
    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:
    Will the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
     OL2C(1), TS 1.34 - Increase in Rated Thermal Power to 
3467 MWt.
    The changes in the OL and TS were evaluated and it was 
determined that the probability (frequency) of a DBA [design-basis 
accident] or other licensing event occurring is not a significant 
function of the power level because the design and regulatory 
criteria originally established for plant equipment (ASME [American 
Society of Mechanical Engineers] code, IEEE [Institute of Electrical 
and Electronics Engineers] standards, NEMA [National Electrical 
Manufacturers Association] standards, Regulatory Guide criteria, 
etc.) are still imposed for the uprated power level. Scram setpoints 
are established such that there will be no significant increase in 
scram frequency due to power uprate.
    The consequences of hypothetical accidents which would occur 
from 102% of the uprated power, as opposed to that previously 
evaluated from 102% of the original power, are in all 
cases insignificant, since the accident evaluations from 102% of 
uprated power do not result in exceeding the NRC-approved acceptance 
limits. A spectrum of hypothetical accidents and transients has been 
investigated for uprated conditions and the bounding events have 
been shown to meet the same regulatory criteria to which they are 
currently licensed. In the area of core design, for example, the 
fuel operating limits such as Maximum Average Planar Linear Heat 
Generation Rate (MAPLHGR) and Safety Limit Minimum Criteria Power 
Ratio (SLMCPR) are still met at the uprated power level.
    The analysis of all limiting events (Section 9) [of General 
Electric Topical Report, NEDC-31994P, ``Power Uprate Licensing 
Evaluation for Nine Mile Point Nuclear Power Station Unit 2,'' 
Revision 1, May 1993] and cycle specific reload analyses will show 
plant transients meet the criteria accepted by the NRC as specified 
in NEDO-24011, GESTAR II. Challenges to fuel or ECCS [emergency core 
cooling system] performance have been evaluated (Section 9.2) [of 
NEDC-31994P] and shown to still meet the criteria of 10CFR50.46 
using the methodology defined by Appendix K (Regulatory Guide 1.70, 
USAR [Updated Safety Analysis Report] Section 6.3). Challenges to 
the containment have been evaluated for uprated power (Section 4.1) 
[of NEDC-31994P] and still meet 10CFR[Part]50 Appendix A Criterion 
38, Long Term Cooling and Criterion 50, Containment. Radiological 
Release events have been evaluated [Sections 8.4 and 8.5] [of NEDC-
31994P] and shown to be a small fraction of the criteria of 
10CFR[Part]100 (Regulatory Guide 1.70 USAR Chapter 15).
    The results of these analyses as discussed above demonstrate 
that operation of [at] the power uprate level does not significantly 
increase the probability or consequences of any accident previously 
evaluated.
     OL2C(7) - Change in Allowable Feedwater Temperature
    This change is made to maintain an equivalent 20 deg.F allowable 
operating range of final feedwater temperature for uprated power 
(405 to 425 deg.F) as compared to the presently licensed range (400 
to 420 deg.F). No change is made to the current method and criteria 
for operation of the feedwater heating systems. The limiting 
transient (Feedwater Controller Failure - Maximum Demand) has been 
evaluated at a feedwater temperature of 405 deg.F (Section 9.1.3) 
[of NEDC-31994P] to demonstrate compliance with all current thermal 
limits criteria. Previous feedwater nozzle evaluations have shown 
that operation with the proposed feedwater temperature operating 
range is acceptable. Therefore, there is no significant increase in 
the probability or consequences of an accident previously evaluated.
     TS Table 2.2.1-1 - Reactor Protection System Instrument 
Setpoints
    The increases in steam dome high pressure scram instrument 
setpoints are made to ensure that there is no significant increase 
in the frequency of scrams due to operation at the higher pressure. 
The increase of the high pressure scram setpoint by the same amount 
as the increase in the planned operating pressure maintains the same 
level of trip avoidance for scram as originally provided. The high 
pressure scram is used as a backup to other scram signals. It has 
been shown that this role is still adequate for uprated operation 
with the revised setpoints (e.g., vessel overpressure protection). 
Since the backup protectionfunctions and the current margins to trip 
avoidance are maintained with the revised setpoints, there is no 
significant increase in the probability or consequences of an 
accident previously evaluated.
     TS Bases Table B2.1.2-2 - Add footnote to applicability 
of table to uprated operation.
    The parameters listed in TS Bases Table B2.1.2-1 come from the 
original statistical analysis performed for BWR4/5 core designs 
(including NMP2) [Nine Mile Point Nuclear Station, Unit 2]. As 
discussed in Section 3.2 of LTR2 (Reference 11-1) [General Electric 
Topical Report NEDC-31984P, ``Generic Evaluation of General Electric 
Boiling Water Reactor Power Uprate,'' July 1991], the uprated 
average bundle power is used to determine the applicability of the 
generic Safety Limit Minimum Critical Power Ratio (SLMCPR) basis for 
each plant. The average bundle power for NMP2 after power uprate is 
4.538 MWt per bundle. This value is acceptable for application of 
the generic SLMCR statistical analysis to uprated NMP2. The generic 
analysis is documented through NEDC-24011-P-A (GESTAR II) and NEDC-
31152P (GE Fuel Bundle Design). Therefore, there is no significant 
increase in the probability or consequences of an accident 
previously evaluated.
     TS 4.1.5.c and TS 4.1.5.d2 - Increases in SLCS [standby 
liquid control system] Surveillance Test Pressure and SLCS pump 
discharge relief valve setpoint.
    The standby liquid control system test pressure increase ensures 
continued ability of the system to pump the required amount of 
sodium pentaborate at the higher operating pressure associated with 
power uprate. This test pressure increase is consistent with the 
ATWS [anticipated transient without scram] analysis provided in 
Section 9.3 [of NEDC-31994P]. The higher pressure setpoint for the 
SLCS relief valve does not exceed the design capability of the SLCS 
components. Surveillance testing at the increased pressure will 
maintain the system's design capability for operation at uprated 
conditions. These changes therefore do not increase the probability 
or consequences of a previously evaluated accident.
     TS Table 3.3.1-1 - Note (i), footnote (**), Action 6 
footnote (*), and Table 3.3.4.2-1 - footnote (**)
    The setpoints for the bypass of T/G [turbine generator] trip 
scram and RPT [recirculation pump trip] at 30% of rated power are 
changed to 125.8 psig and 136.4 psig to be consistent with uprated 
power.
    These changes reflect the redefinition of rated conditions. They 
are consistent with the approach discussed in Section F.4.2(c) of 
LTR1 (Reference 11-2) [General Electric Topical Report NEDC-31897P-
1, ``Generic Guidelines for General Electric Boiling Water Reactor 
Power Uprate,'' June 1991]. There is no significant impact on the 
transient safety analyses which establish core thermal operating 
limit since T/G trips at this partial power setpoint are not 
limiting. Therefore, there is no significant increase in the 
probability or consequences of an accident previously evaluated.
     TS Table 3.3.2-2 Item 1.C.3 - Increase in main 
steamline high flow isolation differential pressure setpoint and 
allowable value.
    The main steamline high flow trip TS changes reflect the 
redefinition of rated steam flow during uprated power operation and 
the application of GE [General Electric Company] setpoint 
methodology. The current analytical basis of 140% of rated steam 
flow is maintained for uprated operation to ensure that an adequate 
trip avoidance margin is maintained (e.g., for disturbances caused 
by full closure testing of MSIVs [main steam isolation valves] or 
turbine inlet valves). The revised setpoints ensure that there is no 
effect on the probability of inadvertent isolation; they have no 
effect on the probability of occurrence of a main steamline break. 
The same isolation initiation function for the main steam line break 
accident is maintained (Section 5.1.2.5) [of NEDC-31994P]. Therefore 
these setpoint changes do not significantly increase the consequence 
of the main steamline break accident.
     TS Table 3.3.2-2 Item 1d - Increase the main steamline 
tunnel temperature setpoints.
    These isolation setpoints are changed to reflect the slight 
increase (about 1 deg.F) in the operating temperature expected for 
uprated operation. Margins between trip setpoints and operating 
temperature are maintained. The increases will avoid unnecessary 
trips. The revised trip setpoints were derived using the GE setpoint 
methodology (documented in NEDC-31336). The setpoints still perform 
their isolation function equivalent to current operation.
    Therefore no significant increase in the probability or 
consequences of an accident previously evaluated results from these 
changes.
     TS Table 3.3.4.1-2 - Increases in the ATWS RPT reactor 
vessel high pressure trip and allowable setpoint.
    The ATWS RPT high pressure setpoints are increased to correspond 
to the increase in the steam dome operating pressure due to power 
uprate. This increase maintains the current margin between the 
operating condition and the trip setpoint to avoid unnecessary 
trips. The capability of the system to adequately perform its ATWS 
function with the new setpoints is shown in Section 9.3 [of NEDC-
31994P]. Therefore the change does not cause a significant increase 
in the probability or consequences of an accident previously 
evaluated.
     TS Figure 3.4.1.1-1 - The figure is revised to reflect 
new definition of rated thermal power in terms of megawatts thermal.
    This change is made to be consistent with the new definition of 
rated thermal power. The current restrictions on operation within 
the restricted power/flow zone are unchanged. The basis for this 
change is described in Section 3.2 of LTR2 (Reference 11-1) [NEDC-
31984P]. There is no significant change in the previously evaluated 
potential for initiation of core thermal hydraulic instability. 
Therefore, this TS change ensures that power uprate operation will 
not cause a significant increase in the probability of [or] 
consequences of an accident previously evaluated.
     TS 3.4.2 - Increase of spring setpoints for the two 
lowest set SRVs [safety relief valves].
    The two low set SRV setpoints are increased to accommodate the 
change in operating pressure after power uprate. This increase in 
the SRV setpoints ensures that approximately the same difference is 
maintained between the RPV [reactor pressure vessel] pressure and 
the lowest SRV setpoint such that there is no increase in the number 
of unnecessary SRV actuations. The increase in the spring setpoints 
by the same amount as the increase planned for normal operation also 
maintains acceptable simmer margin for the SRVs. The SRVs are 
capable of operating at uprated temperatures and pressures as 
evaluated generically in Section 4.6 of LTR2 [NEDC-31984P]. As 
described in Sections 3.2 and 9.3.1 [of NEDC-31994P] a higher RPV 
peak pressure results due to uprate conditions but it is maintained 
well within the ASME Code allowable peak pressure of 1375 psig.
    Therefore, no significant increase in the probability or 
consequences of an accident previously evaluated is caused by this 
change.
     TS 4.4.6.1.3-1 - Revision to the neutron fluence lead 
factor.
    The increase in the lead factor from 0.41 to 0.46 reflects 
updated calculations for the higher power level and projected 
fluence distributions. This calculation accounted for the locations 
of the NMP2 specimen capsules (at three locations on the vessel wall 
around the core beltline region), and the projected uprated 
equilibrium cycle spatial power distributions. Since the revised 
lead factor is consistent with the requirements for vessel 
surveillance, the change causes no significant increase in the 
probability or consequences or [of] an accident previously 
evaluated.
     TS 3.4.6.2 and TS 4.4.6.2 - Increase of reactor steam 
dome operating pressure limit.
    This change to the dome operating pressure limit is consistent 
with and meets the current design criteria used for evaluation of 
steady state operating conditions and for the most limiting 
transient and accident events, i.e., vessel overpressure protection 
and a loss-of-coolant accident (Sections 3.2 and 4.3) [of NEDC-
31994P]. Therefore, there is no significant increase in the 
probability or consequences of an accident previously evaluated.
     TS 4.7.4b - Increase in RCIC [reactor core isolation 
cooling] Surveillance Test Pressure.
    The increase in the RCIC surveillance test pressure requires 
system testing at the higher operating pressure with power uprate. 
The RCIC system has been evaluated and demonstrated to be capable of 
injecting its design flow rate at the higher reactor pressure 
associated with power uprate as discussed in Section 4.2 of LTR2 
(Reference 1) [NEDC-31984P]. This evaluation applies to NMP2 as 
described in Section 3.8 [of NEDC-31994P]. Therefore, this TS change 
ensures that power uprate operation will not cause a significant 
increase in the probability or consequences of an accident 
previously evaluated.
     TS Bases 3/4.2 (References), and TS 6.9.1.9.b(1) 
(Administrative Control) - Revised the reference for the LOCA [loss-
of-coolant accident] analysis methodology to the SAFER/GESTR-LOCA 
methodology report.
    These changes are made to incorporate the power uprate LOCA 
licensing basis. Reference 1 of TS Bases 3/4.2 (References) and the 
report noted in TS 6.9.1.9.b(1) are changed to reflect the improved 
SAFER/GESTR-LOCA methodology used for the NMP2 loss-of-coolant 
accident analysis for power uprate. This methodology has been 
previously approved by the NRC. These changes are made for 
documentation consistency and there is no significant increase in 
the probability or consequences of an accident previously evaluated.
     TS Bases Table B3.2.1-1 - Significant input parameters 
used in the LOCA analysis.
    The changes in the plant parameters used in the uprated LOCA 
analysis are provided from the power uprate analysis (Reference 4-
15, Section 4) [General Electric Report NEDC-31830P, ``NMP2 SAFER/
GESTR-LOCA Loss-of-Coolant Accident Analysis,'' Revision 1, November 
1990]. The power and steam flow values are consistent with 
Regulatory Guide 1.49. The new analysis parameters will also be 
included in USAR Table 6.3-1 as uprate is implemented. Detailed 
information about application of the SAFER/GESTR methodology is 
provided in Reference 1 of TS Bases 3/4.2 (Reference 4-14, Section 
4) [General Electric Report NEDE-23785-1-PA, ``The GESTR/LOCA and 
SAFER Models for the Evaluation of the Loss of Coolant Accident,'' 
Revision 1, October 1984]. The LOCA analysis (Section 4.3) [of NEDC-
31994P] shows that all required criteria are met for operation with 
the uprated parameters.
    Footnote (*) is revised to provide the correct reference for the 
LOCA analysis parameters for NMP2 power uprate. Since the LOCA 
analysis for uprated operation meets all required criteria, these TS 
changes do not cause an increase in the probability or consequences 
of an accident previously evaluated.
     TS Bases B3/4.5.1 and B3/4.5.2 - Increase in required 
capability of the HPCS [high pressure core spray] pump and the 
corresponding differential pressure.
    The increase in the differential pressure for HPCS pump flow 
accommodates the increase in SRV setpoint valves as discussed for 
changes to TS 3.4.2 earlier. This change maintains the currently 
designed functional capability of the HPCS system to provide coolant 
inventory during isolation conditions after a loss of feedwater flow 
transient (backup to RCIC) and during a main steam line break 
(outside containment) accident.
    The small change in the HPCS pump flow (517 versus 516) [gpm] is 
made to be consistent with the value used in the NMP2 SAFER/GESTR 
analysis for a loss of coolant accident. This small change corrects 
the Technical Specification bases for this parameter.
    Therefore, this TS change ensures that power uprate operation 
will not cause a significant increase in the probability or 
consequences of an accident previously evaluated.
     TS Bases B3/4.6.1.2, B3/4.6.1.5, and B3/4.6.2 - Maximum 
containment pressure for leakage testing.
    The bases for the value currently in the TS for the maximum 
containment pressure are reworded to clarify that the maximum 
containment pressure after power uprate will be maintained below the 
current value used for containment leak rate testing. Section 4.1 
[of NEDC-31994P] documents the containment analysis for power uprate 
and shows a peak DBA-LOCA calculated pressure of 36.8 psig (less 
than the current testing requirement of 39.75 psig). There is no 
impact on currently approved requirements and test procedures. 
Therefore, there is no significant increase in the probability or 
consequences of an accident previously evaluated.
    Will the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The Operating License changes in power level and allowable 
feedwater temperature, and the associated Technical Specification 
changes (all listed in Table 11-1) [of NEDC-31994P] will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Equipment that could be impacted by power uprate has been 
evaluated. No new operating mode, equipment lineup, accident 
scenario, or equipment failure mode has been identified. The full 
spectrum of accident considerations defined in Regulatory Guide 1.70 
have been reviewed and no new or different kind of accident has been 
identified. Power uprate uses already developed technology and 
applies it within the capabilities of already existing plant 
equipment in accordance with presently existing regulatory criteria 
to include NRC approved codes, standards, and methods. GE has 
designed BWRs of higher power levels than the uprated power of any 
of the currently operating BWR fleet and no new power dependent 
accidents have been identified.
    The Technical Specifications changes required to implement power 
uprate require minor changes to the configuration of the plant, and 
all the Technical Specification changes have been evaluated and are 
acceptable.
    Will the change involve a significant reduction in a margin of 
safety?
     OL2C(1), TS 1.34 - Increase in Rated Thermal Power to 
3467 MWt.
    Power uprate will not involve a significant reduction in a 
margin of safety, since the licensing evaluations were performed 
either at plant conditions higher than the proposed uprate 
conditions, or used approved methodologies which incorporate 
appropriate allowances for uncertainties. As discussed throughout 
this report (e.g., Section 11.1) [of NEDC-31994P] and in Section 5 
of Reference 11-2 [NEDC-31897P-1], the safety margins prescribed by 
the Code of Federal Regulations have been maintained by meeting the 
appropriate regulatory criteria. Similarly, the margins provided by 
the application of the American Society of Mechanical Engineers 
(ASME) design acceptance criteria where applicable have been 
maintained (e.g., see Section 3.2) [of NEDC-31994P]. Other margin-
assuring acceptance criteria have also been maintained.
    All limiting accident and transient analyses have been 
reperformed at uprated power operating conditions consistent with 
the requested Technical Specification changes. The NRC-approved 
SAFER/GESTR-LOCA methodology was used in the LOCA analysis. 
Additionally, Reference 11-2 [NEDC-31897P-1] addresses the BWR 
generic acceptability of analytical evaluations for the loss of 
feedwater transient, stability, core spray distribution, safety 
limit minimum critical power ratio, containment atmosphere 
combustibility, materials and coolant chemistry, and anticipated 
transients without scram (ATWS).
    As discussed in Section 5.2.3 of Reference 11-2 [NEDC-31897P-1], 
offsite doses for the DBA/LOCA will increase proportionally to 
reactor power and can be compared on a consistent basis.
    As evaluated in Sections 8.5 and 9.2 [of NEDC-31994P], the 
results remain a small fraction of the acceptance criteria of 
10CFR[Part]100.
    The radiological doses resulting from the DBA/LOCA and MSLB 
[main steam line break] accidents were initially analyzed at 3489 
MWt (105% of 3323 MWt). For the power uprate program, the increase 
in the analyzed power level is only 1.3% (from 3489 to 3536 MWt) 
which provides the uncertainty factor (2%) required by Regulatory 
Guide 1.49.
    It is concluded that there is no significant decrease in a 
margin of safety.
     OL2C(7) - Change in Allowable Feedwater Temperature
    The increase in the lower limit of the allowable operating range 
of the final feedwater temperature for uprate power was evaluated by 
reanalyzing the Feedwater Controller Failure - Maximum Demand 
transient at a feedwater temperature of 405 deg.F (Section 9.1.3) 
[of NEDC-31994P]. In addition, the reactor pressure vessel feedwater 
nozzle has been evaluated for the 20 deg.F range of feedwater 
temperature. The results of these evaluations demonstrate that 
current fuel thermal limits criteria and ASME Code criteria are met. 
Therefore, there is no significant decrease in a margin of safety.
     TS Table 2.2.1-1 - Reactor Protection System Instrument 
Setpoints
    The increases in the steam dome high pressure scram instrument 
setpoints for uprated power were evaluated by determining if the 
high pressure scram, which is used as a backup to other scram 
signals, provides adequate overpressure protection. The evaluation 
demonstrates that the backup protection function, with the revised 
setpoints, continues to provide adequate overpressure protection at 
uprated power conditions by meeting the applicable ASME Code 
criteria. It is concluded that there is no significant decrease in a 
margin of safety.
     TS Bases Table B2.1.2-2 - Add footnote to applicability 
of table to uprated operation
    The nominal values of parameters used in the statistical 
analyses of the fuel cladding integrity safety limit were re-
evaluated at power uprate conditions. This evaluation demonstrates 
that the average bundle power at uprated conditions for NMP2 is 
acceptable for application of the generic safety limit minimum 
critical power ratio statistical analysis. Therefore, it is 
concluded that there is no significant decrease in a margin of 
safety.
     TS 4.1.5.c and TS 4.1.5.d2 - Increase in SLCS 
Surveillance Test Pressure and SLCS pump discharge relief valve 
setpoint.
    The SLCS surveillance test pressure was increased to provide 
periodic demonstration of the ability of the SLCS to provide the 
required amount of sodium pentaborate solution at the higher 
pressure associated with an ATWS event postulated to occur at power 
uprate conditions. At this increased SLCS discharge pressure, the 
system provides an adequate shutdown backup capability by having the 
ability to bring the isolated reactor from full power to a cold, 
Xenon-free shutdown condition, assuming that the withdrawn control 
rods remain fixed in the uprated power pattern.
    For power uprate, the capability of the SLCS to respond with 
adequate margin to a postulated ATWS event was confirmed. The most 
limiting ATWS events evaluated for peak vessel pressure and peak 
suppression pool temperature were: (1) closure of all MSIVs and (2) 
inadvertent opening of a relief valve. The peak pressure for the 
MSIV closure event which included simulation of the higher relief 
setpoints and two relief valves out of service, demonstrates that 
the peak pressure, 1325 psig, remains below the ASME emergency 
overpressure protection criteria of approximately 1500 psig, which 
is applicable to an ATWS event. The reactor pressure is controlled 
by the relief valves (after the initial peak) within the pressure 
specified in this revised Technical Specification. SLCS injection 
takes place during this period with the relief valves controlling 
pressure.
    The peak suppression pool temperature for the inadvertent 
opening of a relief valve was demonstrated to remain below the ATWS 
peak pool temperature criteria of 190 deg. for a Mark II containment 
design, which is applicable to NMP2. Peak containment pressure was 
well below the 45 psig containment design pressure. For this event, 
SLCS injection will be at vessel pressures bounded by the revised 
Technical Specification. The higher pressure setpoint of the SLCS 
pump discharge relief valve provides adequate overpressure 
protection of the SLCS pressure boundary by meeting the applicable 
ASME Code criteria (equal to or less than the system piping design 
pressure).
    In summary, peak vessel pressure is below ASME code criteria, 
and suppression pool temperature is below the ATWS peak pool 
temperature criterion for Mark II containment design, peak 
containment pressure is well below the containment design pressure, 
the SLCS injection pressure during the bounding events is within the 
new Technical Specification testing requirement, and the SLCS 
pressure boundary is maintained in compliance with ASME Code 
criteria. Therefore, it is concluded that there is no significant 
decrease in a margin to safety.
     TS Table 3.3.1-1 - Note (i), footnote (**), Action 6, 
footnote (*), and Table 3.3.4.2-1 footnote (**)
    The increase in the setpoints for the bypass of T/G trip scram 
and RPT at 30% power are made to be consistent with uprated power. 
These increased setpoints do not significantly reduce a margin of 
safety since the T/G trips at this partial power setpoint continue 
to be non-limiting events.
     TS Table 3.3.2-2 Item 1.c.3 - Increase in main steam 
line high flow differential pressure setpoint and allowable valve.
    The increase in the main steam line high flow differential 
pressure setpoint and allowable value reflect the redefinition of 
rated conditions. The increased setpoint will maintain the same 
inadvertent trip avoidance margin, thereby avoiding any increase in 
the frequency of occurrence of isolation events. The closure of the 
MSIV remains assured during the limiting event (the steam line break 
accident). The break flow rate (controlled by the flow restrictor) 
will be about 190% (Section 3.5) [of NEDC-31994P], so the setpoint 
at less than 140% will sense the accident as effectively as for 
current operation. It is concluded that this change does not result 
in a significant decrease in a margin of safety.
    TS Table 3.3.2-2 Item 1.d - Increase in main steam line 
tunnel temperature setpoints.
    These isolation setpoints are changed to reflect the slight 
increase (about 1 deg.F) in the steam tunnel operating temperature 
expected for uprated operation. The increase in these setpoints 
ensures adequate inadvertent trip avoidance. The analytical upper 
limits for these setpoints are not changed so that their safety 
functions are not impacted by the Technical Specification changes. 
For example, the instruments will act at the same setpoints assumed 
in previous analysis for a main steamline break, ensuring that 
offsite radiological doses remain a small fraction of 10CFR[Part]100 
criteria and within GDC19 criteria for control room doses. 
Therefore, there is no significant decrease in a margin of safety.
     TS Table 3.3.4.1-2 - Increases in the ATWS RPT reactor 
vessel high pressure trip and allowable setpoints
    The purpose of the high pressure RPT is to reduce reactor power 
level during a postulated pressurization transient with scram 
assumed to fail (ATWS). The physical phenomenon involved is that the 
void reactivity feedback from a pressurization transient adds 
positive reactivity to the reactor system. However, the high 
pressure RPT system trips both recirculation pumps to the low speed 
condition, thereby increasing core void fraction and creating 
negative reactivity to reduce the power transient. This enables the 
safety/relief valves to maintain peak pressure within the ASME 
overpressure emergency limit for the bounding ATWS case (Section 
9.3.1) [of NEDC-31994P].
    For power uprate, the capability of the SLCS to respond to a 
postulated ATWS event with adequate margin was confirmed (Section 
9.3.1) [of NEDC-31994P]. By reducing reactor power until the SLCS 
can be injected to achieve full shutdown, the RPT also reduces 
suppression pool temperature for isolation cases (also shown to be 
acceptable for power uprate conditions in Section 9.3.1) [of NEDC-
31994P]. Therefore, it is concluded that there is no significant 
decrease in a margin of safety.
     TS Figure 3.4.1.1-1 - The figure is revised to reflect 
the new definition of rated thermal power in terms of megawatts 
thermal.
    This change is made to be consistent with the new definition of 
rated thermal power. As described in Section 3.2 of LTR2 (Reference 
11-1) [NEDC-31984], the change to the power flow restricted zone is 
made to maintain the same operating constraints and stability margin 
that were established for the current power level. This change 
avoids any increase in the possibility of occurrence or any increase 
in the potential effects of power oscillations. Therefore, there is 
no significant decrease in a margin of safety.
     TS 3.4.2 - Increase of spring setpoints for the two 
lowest set SRVs.
    The two low set SRV setpoints are increased to accommodate the 
change in operating pressure after power uprate. This change 
maintains a simmer margin of greater than 120 psig. Power uprate 
analysis shows that the revised SRVs still maintain the peak RPV 
pressure within the ASME Code Upset limit of 1375 psig for the 
limiting pressurization event (MSIV closure when credit is only 
taken for the backup high neutron flux scram) and provide adequate 
protection for postulated ATWS events. See Sections 3.2 and 9.3.1 
[of NEDC-31994P] for further discussion. Therefore, it is concluded 
that there is no significant decrease in a margin of safety.
     TS 4.4.6.1.3-1 - Revision of the neutron fluence lead 
factor.
    The increase in the lead factor includes consideration of the 
higher power level and projected spatial power distributions for an 
uprated equilibrium cycle. The evaluation at uprated conditions 
utilized the same calculational approach as performed for the 
current neutron fluence lead factor, but using more precise input 
parameters for uprated conditions. Therefore, it is concluded that 
there is no significant decrease in a margin of safety.
     TS 3.4.6.2 and TS 4.4.6.2 - Increase of reactor steam 
dome operating pressure limit.
    The change to the dome operating pressure limit is made to be 
consistent with the new operating pressure for uprated thermal 
power. This change is used as a direct initial condition analysis 
input or sensitivity study parameter in the evaluation of steady 
state operating conditions and for the most limiting transients and 
accident events, i.e., vessel overpressure protection and LOCA. With 
this revised limit, peak vessel pressure remains below ASME Code 
criteria, and LOCA fuel performance satisfies the requirements of 
10CFR50.46 and 10CFR[Part]50 Appendix K. Therefore, there is no 
significant decrease in a margin of safety.
     TS 4.7.4b - Increase in RCIC Surveillance Test 
Pressure.
    The RCIC surveillance test pressure was increased to provide 
periodic demonstration of the ability of RCIC system to perform 
consistent with the requirements of the analyses at the higher 
operating pressure associated with power uprate conditions. An 
evaluation of the RCIC system confirmed its ability to operate at 
slightly higher turbine speed and provide its design flow rate at 
power uprate conditions. RCIC system performance will be confirmed 
during the initial power ascension to uprated conditions (and 
periodically thereafter per the Technical Specification). Therefore, 
it is concluded that there is no significant decrease in a margin of 
safety.
     TS Bases 3/4.2 (References) and TS 6.9.1.9.b (1) 
(Administrative Control) - Revised the references for the LOCA 
analysis methodology to the SAFER/GESTR-LOCA methodology report.
    These changes are made to incorporate the power uprate LOCA 
licensing basis. Since SAFER/GESTR-LOCA methodology has been 
previously approved by the NRC and is acceptable for use for NMP2, 
it is concluded this change does not significantly decrease a margin 
of safety.
     TS Bases Table B3.2.1-1 - Significant input parameters 
used in the LOCA analyses.
    The changes in the plant parameters reflect the power uprate 
condition. These changes have been reflected as input parameters in 
the LOCA analyses consistent with Regulatory Guide 1.49. Since the 
LOCA analysis demonstrates that 10CFR50.46 and 10CFR[Part]50 
Appendix K criteria are met for operation with the uprated 
parameters, it is concluded that there is no significant decrease in 
a margin of safety.
     TS Bases B3/4.5.1 and B3/4.5.2 - Increase in required 
capability of the HPCS pump and the corresponding differential 
pressure.
    The increase in differential pressure accommodates the increase 
in SRV setpoint values previously discussed to TS 3.4.2. The 
increased differential pressure ensures there is sufficient HPCS 
flow, assuming the two lowest setpoint SRVs are out of service, and 
the commencement of flow as reflected in the analysis of isolation 
events. The increase in HPCS pump flow is reflected in the LOCA 
analyses (516 to 517 gpm). This small change corrects the Technical 
Specification bases for this parameter. The revised parameters for 
the HPCS pump differential pressure and flow are reflected as inputs 
to the LOCA analyses and analyses of isolation events. Since the 
LOCA analysis meets 10CFR50.46 criteria and 10CFR[Part]50 Appendix K 
criteria, and the isolation events meet all required criteria (e.g. 
top of fuel remains covered for the loss of feedwater transient) it 
is concluded that there is no significant decrease in a margin of 
safety.
     TS Bases B3/4.6.1.2, B3/4.6.1.5 and B3/4.6.2 - Maximum 
containment pressure for leakage testing.
    The bases for the value currently in the TS for the maximum 
containment pressure are reworded to clarify that the maximum 
containment pressure for power uprate has been calculated to remain 
below the current value used for containment leak rate testing. 
Therefore, it is concluded that there is no significant decrease in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Robert A. Capra

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of amendment request: January 10, 1994
    Description of amendment request: The proposed amendment would 
relocate the seismic monitoring instrumentation Limiting Condition for 
Operation, Surveillance Requirements and associated tables and Bases 
contained in TS sections 3.3.7.2 and 4.3.7.2 to the Updated Final 
Safety Analysis Report (UFSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed Technical Specifications (TS) changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The function of the seismic monitoring instrumentation system is 
to monitor the magnitude and effect of a seismic event only, and can 
not initiate or mitigate an accident previously evaluated. 
Furthermore, the proposed TS changes to relocate the seismic 
monitoring instrumentation requirements from TS to the UFSAR are in 
accordance with the criteria for determining those requirements that 
should remain in the TS as defined by the NRC in its final policy 
statement, ``Final Policy Statement on Technical Specifications 
Improvements for Nuclear Power Reactors,'' dated July 22, 1993. The 
seismic monitoring instrumentation LCO, SRs, and associated tables 
and Bases proposed for relocation from TS to the LGS UFSAR will 
continue to be implemented by administrative controls that will 
satisfy the applicable requirements of TS section 6 ``Administrative 
Controls.'' Those requirements include a review of changes to plant 
systems and equipment and to the applicable administrative controls 
in accordance with the provisions of 10CFR50.59.
    Criterion 2 of the July 22, 1993 NRC final policy statement 
states, ``A process variable, design feature, or operating 
restriction that is an initial condition of a Design Basis Accident 
or Transient Analysis that either assumes the failure of or presents 
a challenge to the integrity of a fission product barrier.'' The 
seismic monitoring instrumentation system is not a system that 
monitors a process variable that is an initial condition for 
accident or transient analyses. The seismic monitoring 
instrumentation is also not a design feature or an operating 
restriction that is an initial condition of a Design Basis Accident 
or transient analyses since it only provides information regarding 
the magnitude of and the plant equipment response to a Design Basis 
earthquake. Therefore, the current LGS seismic monitoring 
instrumentation TS requirements do not meet Criterion 2 of the July 
22, 1993 NRC final policy statement.
    Criterion 3 of the July 22, 1993 NRC final policy statement 
states, ``A structure, system, or component that is part of the 
primary success path and which functions or actuates to mitigate a 
Design Basis Accident or Transient that either assumes the failure 
of or presents a challenge to the integrity of a fission product 
barrier.'' The LGS seismic monitoring instrumentation system does 
not provide a function or actuate in order to mitigate the 
consequences of a Design Basis Accident or transient. Therefore, the 
current LGS seismic monitoring instrumentation TS requirements do 
not meet Criterion 3 of the July 22, 1993 NRC final policy 
statement.
    Criterion 4 of the July 22, 1993 NRC final policy statement 
states, ``A structure, system or component which operating 
experience or probabilistic safety assessment has shown to be 
significant to public health and safety.'' Operating experience has 
shown that the LGS seismic monitoring instrumentation system has no 
impact on public health and safety as defined by the NRC final 
policy statement. Furthermore, LGS specific probabilistic risk 
assessment (PRA) does not credit the seismic monitoring 
instrumentation system as a significant factor in the plant response 
to an accident. Therefore, the current LGS seismic monitoring 
instrumentation TS requirements do not meet Criterion 4 of the July 
22, 1993 NRC final policy statement for determining those 
requirements that should remain in TS. This conclusion is consistent 
with the function of the seismic monitoring instrumentation system 
stated above.
    These proposed TS changes will maintain the current operation, 
maintenance, testing, and system operability controls of the seismic 
monitoring instrumentation system. Furthermore, any further changes 
to the seismic monitoring instrumentation system will be evaluated 
for the effect of the those changes on system reliability as 
required by 10CFR50.59. The seismic monitoring instrumentation 
system performance will not decrease due to these proposed TS 
changes and the system will continue to be administratively 
controlled in accordance with TS Section 6, including the 
requirements of 10CFR50.59, thereby precluding a future decrease in 
its performance.
    In accordance with the current TS Section 3.3.7.2, with the 
seismic monitoring instrumentation inoperable, the plant would not 
be required to shut down and the provisions of TS Section 3.0.3 
(i.e., plant shutdown) would not be applicable. Therefore, the 
inoperability of this system and therefore the consequences of an 
accident while this system is inoperable, was previously evaluated 
as not significant enough to require a change to the plant operating 
condition.
    Since the seismic monitoring instrumentation system does not 
monitor a process variable that is an initial condition for an 
accident or transient analyses, or actuates any accident mitigation 
feature, and since the operation, maintenance, testing, and 
modification of the seismic monitoring instrumentation system will 
continue to be administratively controlled, including the 
requirements of 10CFR50.59; therefore, maintaining the reliability 
of the system, the proposed TS changes will not involve an increase 
in the probability or consequences of an accident previously 
evaluated.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The function of the seismic monitoring instrumentation system is 
to monitor the magnitude and effect of a seismic event only. The 
proposed TS changes to relocate the seismic monitoring instruments 
requirements from TS to the UFSAR are in accordance with the 
criteria for determining those requirements that should remain in 
the TS as defined by the NRC in its final policy statement, dated 
July 22, 1993. The seismic monitoring instrumentation system does 
not monitor a process variable that is an initial condition for an 
accident or transient analyses.
    The seismic monitoring instrumentation is also not a design 
feature or an operating restriction that is an initial condition of 
a Design Basis Accident or transient analyses since it only provides 
information regarding the magnitude of and the plant equipment 
response to a Design Basis earthquake.
    These proposed TS changes to relocate the TS requirements to the 
UFSAR will not alter the operation of the plant, or the manner in 
which the seismic monitoring instrumentation system will perform its 
function, and any future changes will continue to be 
administratively controlled in accordance with TS

    Section 6, including the requirements of 10CFR50.59.
    These proposed TS changes will not impose new conditions nor 
result in new types of equipment which will result in different 
types of malfunctions of equipment important to safety than any type 
previously evaluated.
    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.
    These proposed TS changes to relocate the seismic monitoring 
instrumentation requirements from TS to the UFSAR are in accordance 
with the criteria for determining those requirements that should 
remain in the TS as defined by the NRC in final policy statement, 
dated July 22, 1993.
    Criterion 1 of the NRC final policy statement 
states,Installed instrumentation that is used to detect, 
and indicate in the control room, a significant abnormal degradation 
of the reactor coolant pressure boundary.'' The NRC final policy 
statement explains that ''...This criterion is intended to ensure 
that Technical Specifications control those instruments specifically 
installed to detect excessive reactor coolant leakage. This 
criterion should not, however, be interpreted to include 
instrumentation to detect precursors to reactor coolant pressure 
boundary leakage or instrumentation to identify the source of actual 
leakage (e.g., loose parts monitor, seismic instrumentation, valve 
position indicators).'' Based on the above NRC guidance, the LGS 
UFSAR, and TS Bases 3.3.7.2, the seismic monitoring instumentation 
does not detect, and indicate in the control room, a significant 
abnormal degradation of the reactor coolant pressure boundary. 
Therefore, the current LGS seismic monitoring instrumentation TS 
requirements do not meet Criterion 1. Furthermore, operating 
experience has shown that the LGS seismic instrumentation system has 
no impact on public health and safety as defined by the NRC final 
policy statement. In addition, the LGS specific PRA does not credit 
the seismic monitoring instrumentation system as a significant 
factor in the plant response to accidents.
    The seismic monitoring instrumentation LCO, SRs, and associated 
tables and Bases proposed for relocation to the LGS UFSAR will 
continue to be implemented by administrative controls that will 
satisfy the applicable requirements of TS section 6 ``Administrative 
Controls.'' Those requirements include a review of future changes to 
the system and applicable administrative controls in accordance with 
the provisions of 10CFR50.59.
    Accordingly, based on the above discussion of NRC specific 
guidance, operating experience, and continued imposition of 
administrative controls, the proposed TS changes do 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

South Carolina Electric & Gas Company, South Carolina Public 
ServiceAuthority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: December 13, 1993, as supplemented 
February 2, 1994
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to allow for the use and 
subsequent storage of fuel with an initial enrichment of 5.0 w/o 
[weight percent] Uranium 235. The TS currently allow the use of fuel 
with a maximum enrichment of 4.25 w/o Uranium 235. The proposed 
amendment would also revise the restrictions on fuel storage in regions 
1 and 2 of the spent fuel pool to ensure that the design basis for 
preventing criticality is maintained in the event of absorber panel 
shrinkage.
    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 change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    There is no increase in the probability of an accident because 
the physical characteristics of a fuel assembly are not changed when 
fuel enrichment is increased. Fuel assembly movement will continue 
to be controlled by approved fuel handling procedures.
    There is no increase in the consequences of an accident because 
fuel cycle designs will continue to be analyzed with NRC-approved 
codes and methods to ensure the design bases for VCSNS [Virgil C. 
Summer Nuclear Station] are satisfied. The double contingency 
principle of ANSI/ANS 8.1-1983 can be applied to any postulated 
accident in the spent fuel pool which could cause reactivity to 
increase beyond the analyzed conditions. As shown in Attachment IV, 
the level of boron in the VCSNS spent fuel pool is sufficient to 
maintain Keff [effective neutron multiplication factor] less than or 
equal to 0.95. There is no postulated accident which could cause 
reactivity to increase beyond the analyzed conditions in the new 
fuel rack.
    The radiological consequence analyses [...] performed to support 
the installation of replacement steam generators at VCSNS included 
the development of source terms which bound fuel enrichments up to 
5.0 w/o U235 [Uranium 235] and average discharge burnups up to 
65,730 MWD/MTU [megawatt days per metric ton uranium], which bounds 
the currently licensed burnup for fuel at VCSNS. These source terms 
were used to calculate offsite doses for accidents that are 
postulated to result in the release of fission products to the 
environment, including the fuel handling accident. In all cases, the 
dose results are within 10CFR100 limits.
    2. The change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed Technical Specification changes do not involve any 
physical changes to the plant or any changes to the method in which 
the plant is operated. They do not affect the performance or 
qualification of safety related equipment. Therefore the possibility 
of a different type of accident or malfunction than previously 
considered is not created.
    3. The change does not involve a significant reduction in a 
margin of safety.
    Criticality analyses [...] have been performed for the spent 
fuel pool to allow for storage of fuel assemblies with enrichments 
up to 5.0 without U-235. The proposed Technical specification 
changes include those necessary to maintain Keff less than or equal 
to 0.95, including conservative allowances for uncertainties and 
biases, when the pool is flooded with unborated water.
    The new fuel racks have been previously analyzed [...] for 
storage of fuel assemblies with enrichments up to 5.0 w/o U-235. For 
the flooded condition Keff does not exceed 0.95 including 
conservative allowances for uncertainties and biases. For the 
normally dry condition Keff does not exceed 0.98 for the low density 
optimum moderation condition. However, the proposed Technical 
Specification changes require fuel assemblies with enrichment above 
4.0 w/o U-235 to contain integral fuel burnable absorbers such that 
the maximum reference fuel [infinite neutron multiplication factor] 
is less than or equal to 1.460 in unborated water at 68 deg.F due to 
restrictions on spent fuel storage.
    Since the proposed changes ensure that the design basis for 
preventing criticality in the fuel storage areas is preserved and 
since fuel cycle designs will continue to be analyzed with NRC-
approved codes and methods to ensure the design bases for VCSNS are 
satisfied, there is no 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: Fairfield County Library, 
Garden and Washington Streets, Winnsboro, South Carolina 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: S. Singh Bajwa

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama

    Date of amendments request: February 16, 1994
    Description of amendments request: The proposed amendment will 
change the Technical Specifications to modify the description of fuel 
and control rod assemblies in TS 5.3.1, Fuel Assemblies. The change to 
the fuel assembly description will permit the limited substitution of 
zirconium alloy, zircaloy-4, ZIRLOTM, or stainless steel filler 
rods for fuel rods in accordance with the NRC-approved applications of 
fuel rod configurations that have been analyzed with NRC-approved 
methods. This change will allow timely removal of fuel rods that are 
found to be a probable source of future leakage. The change will make 
provisions for the loading of lead test assemblies into the reactor 
without requiring a specific TS change. This amendment also allows the 
use of ZIRLOTM clad fuel as lead test assemblies. The specific 
descriptions of the fuel and control rod assemblies contained in the TS 
which are restrictive due to the unnecessary details are being deleted. 
The change will also make line item improvements in the Technical 
Specifications in accordance with Generic Letter 90-02, Supplement 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change to the Technical Specifications allowing 
reconstitution will not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because it will not result in a change to any of the process 
variables that might initiate an accident or affect the radiological 
release for an accident. The operating limits will not be changed 
and the analysis methods to demonstrate operation within the limits 
will remain in accordance with NRC-approved methodology. Other than 
the changes to the fuel assemblies, there are no physical changes to 
the plant associated with this Technical Specification change. The 
consequences of an accident previously evaluated will not be 
increased because the safety analysis to be performed for each cycle 
will continue to demonstrate compliance with all fuel safety design 
bases. The ability to remove potentially leaking fuel rods should 
result in a reduction in the radiological consequences of any 
transients or accidents.
    The probability or consequences of an accident previously 
evaluated are not significantly increased with the use of 
ZIRLOTM cladding. The VANTAGE 5 fuel assemblies containing 
ZIRLOTM clad fuel rods meet the same fuel assembly and fuel rod 
design bases as other VANTAGE 5 fuel assemblies. In addition, the 10 
CFR 50.46 criteria will be applied to the ZIRLOTM clad fuel 
rods. The use of these fuel assemblies will not result in a change 
to the proposed Farley VANTAGE 5 reload design and safety analysis 
limits. Since the original design criteria are being met, the 
ZIRLOTM clad fuel rods will not be an initiator for any new 
accident. The ZIRLOTM clad material is similar in chemical 
composition and has similar physical and mechanical properties as 
that of zircaloy-4. Thus, the cladding integrity is maintained and 
the structural integrity of the fuel assembly is not affected. The 
ZIRLOTM clad fuel rod improves corrosion performance and 
dimensional stability. No concerns have been identified with respect 
to the use of an assembly containing a combination of both zircaloy-
4 and selected ZIRLOTM clad fuel rods. Since the dose 
predications in the Farley safety analyses are not sensitive to the 
fuel rod cladding material used, the radiological consequences of 
accidents previously evaluated in the Farley safety analysis remain 
valid. Therefore the probability or consequences of an accident 
previously evaluated are not significantly increased.
    The proposed removal of detailed descriptions of fuel and 
control rod assemblies will not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because it will not result in a change to any of the process 
variables that might initiate an accident. The operating limits will 
not be changed and the analysis methods to demonstrate operation 
within the limits will remain in accordance with NRC-approved 
methodology. The consequences of an accident previously evaluated 
will not be increased because the safety analyses to be performed 
for each cycle will continue to demonstrate compliance with all fuel 
safety design bases.
    2. This change to the Technical Specifications allowing 
reconstitution will not create the possibility of a new or different 
kind of accident from any accident previously evaluated because it 
will only affect the assembly configuration and will be limited to 
NRC-approved applications of fuel rod configurations. The other 
aspects of plant design, operation, limitations and responses to 
events will remain unchanged.
    The possibility for a new or different kind of accident from any 
accident previously evaluated is not created by the use of 
ZIRLOTM cladding since the VANTAGE 5 fuel assemblies containing 
ZIRLOTM clad fuel rods will satisfy the same design bases as 
that used for other VANTAGE 5 fuel assemblies. All design and 
performance criteria will continue to be met and no new single 
failure mechanisms have been defined. In addition, the use of these 
fuel assemblies does not involve any alterations to plant equipment 
or procedures that would introduce any new or unique operational 
modes or accident precursors. Therefore, the possibility for a new 
or different kind of accident previously evaluated is not created.
    The removal of detailed descriptions of fuel and control rod 
assemblies will not create the possibility of a new or different 
kind of accident from any accident previously evaluated because they 
will be limited to NRC-approved applications of fuel rod 
configurations. The other aspects of plant design, operation, 
limitations and responses to events will remain unchanged.
    3. The use of zirconium alloy, zircaloy-4, ZIRLOTM, or 
stainless steel filler rods in fuel assemblies will not involve a 
significant reduction in a margin of safety because analyses using 
NRC-approved methods will be performed for each configuration to 
demonstrate continued operation within the limits that assure 
acceptable plant response to accidents and transients. These 
analyses will be performed using NRC-approved methods that have been 
approved for application to the fuel configuration.

    The margin of safety is not significantly reduced by the use of 
ZIRLOTM clad [sic] since the VANTAGE 5 fuel assemblies 
containing ZIRLOTM clad fuel rods do not change the proposed 
Farley VANTAGE 5 reload design and safety analysis limits. The use 
of these fuel assemblies will take into consideration the normal 
core operating conditions allowed for in the Technical 
Specifications. For each cycle reload core, the fuel assemblies will 
be evaluated using NRC Staff-approved reload design methods. This 
will include consideration of the core physics analysis peaking 
factors and core average linear heat rate effects. Therefore, the 
margin of safety as defined in the bases to the Farley Technical 
Specifications and VANTAGE 5 Licensing Amendment Request is not 
significantly reduced.
    The removal of detailed descriptions of fuel assemblies will not 
involve a significant reduction in a margin of safety because 
analyses using NRC-approved methods will be performed for each 
configuration to demonstrate continued operation within the limits 
that assure acceptable plant response to accidents and transients. 
These analyses will be performed using NRC-approved methods that 
have been approved for application to the fuel configuration.
    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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: James H. Miller, III, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: S. Singh Bajwa

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 7, 1994 TS 93-19
    Description of amendment request: The proposed change would revise 
Technical Specification 5.3.1 to allow the substitution of filler rods 
for fuel rods in fuel assemblies. This would permit the timely removal 
of fuel rods that are found to be leaking or are determined to be the 
probable source of future leaks.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification change 
and has determined that it does not represent a significant hazards 
consideration based on criteria established in 10 CFR 50.92(c). 
Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The substitution of filler rods will be justified using NRC-
approved methodology. This methodology will demonstrate that the 
existing design limits and safety analyses criteria are met. 
Therefore, the proposed change does not increase the consequences of 
an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change involves the substitution of filler rods for 
fuel rods. This substitution requires the utilization of NRC- 
approved methodology. This methodology will ensure that the specific 
analyses will not cause any new or different kind of accident from 
that previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The substitution of filler rods for fuel rods would result in 
less active fuel in the core. Therefore, the amounts of radiological 
effluents that may be released offsite would not increase. The NRC-
approved methodology by which any reanalyses would be performed 
already accounts for the affects on grid strength or the mass, 
stiffness, and fundamental frequency of the fuel assembly during 
seismic and loss-of-cooling accident conditions. Thus, the margin of 
safety is not reduced when substituting filler rods for fuel rods.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
NuclearPlant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 7, 1994 TS 93-11
    Description of amendment request: The proposed change would revise 
Surveillance Requirement (SR) 4.7.9.i, ``Snubber Service Life 
Program,'' to replace the present wording that describes the service 
life hydraulic snubber monitoring and evaluation program with that from 
the Westinghouse Electric Corporation Standard Technical 
Specifications, Revision 4a. This would eliminate the need to perform 
an engineering evaluation for drag-force increases of 50 percent or 
greater of the previously measured value and substitute a requirement 
to establish a monitoring program. This program would require that a 
maximum service life for the snubber components be determined and the 
monitoring program be established to ensure that the maximum service 
life is not exceeded based on test results and failure history. A 
proposed change to SR 4.7.9.c would remove the wording that is 
inconsistent with Generic Letter 90-09 by removing the term ``if 
applicable'' for performance of an as-found functional test and the 
requirement related to tests of hydraulic snubbers that have uncovered 
fluid ports.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    TVA proposes to delete the current TS requirements to perform an 
evaluation of snubber test data when there is a greater than 50 
percent increase in drag force for mechanical snubbers. The 50 
percent evaluation requirement is considered unnecessary where 
snubbers have small drag forces during their previous test. During 
subsequent testing, small increases in drag forces (when compared 
with the rated load of the snubber) may exceed 50 percent of the 
previous test value. The relative change in drag force is small when 
compared with the overall rating of the snubber; however, under the 
current TS, an engineering evaluation for impending failure will 
still be required. Eliminating the current evaluation requirement 
from SQN's TSs will reduce the burden associated with performing 
unnecessary evaluations. The proposed change is consistent with the 
standard TS (Revision 4a), and a Snubber Service Life Program 
continues to exist at SQN. Therefore, there is no increase in the 
probability or consequences of an accident previously evaluated.
    In addition, the testing language associated with Visual 
Inspection Performance and Evaluation has been deleted to provide 
consistency with Generic Letter 90-09. It should be noted that SQN 
tests its hydraulic snubbers in either direction as necessary. This 
is more conservative than the present TS requirement; therefore, 
this change 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 
from any previously analyzed.
    The two proposed changes involve deleting a requirement to 
perform unnecessary analyses and a potentially nonconservative 
testing requirement. These changes do not alter any plant operation, 
maintenance requirements, or system design or function. Therefore, a 
new or different kind of accident is not created by this proposed 
change.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to the visual inspection and the service 
life section will not modify the plant or revise its mode of 
operation or the present safety analysis. The trending criteria to 
be utilized provide adequate assurance that snubber impending 
failure will be predicted in a timely manner. The deleted sections 
will not change the requirement to test and trend data for snubbers 
to predict failure; therefore, there is not a reduction in any 
margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 8, 1994 TS 93-14
    Description of amendment request: The proposed change would revise 
the setpoints in Technical Specification Table 3.3-4, ``Engineered 
Safety Feature Actuation System Instrumentation,'' for the pressure 
switches used to control switchover of the motor-driven Auxiliary 
Feedwater pump suction from the normal condensate storage tank supply 
to the essential raw cooling water supply. The setpoints would be 
changed from the present trip setpoint of 2 psig and allowable value of 
1 psig, to new values of 3.21 psig and 2.44 psig, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The auxiliary feedwater (AFW) system is designed to mitigate the 
effects of the design basis accidents and anticipated operational 
transients listed below:
    A. Loss of normal feedwater
    B. Loss of offsite power to station auxiliaries
    C. Accidental depressurization in the main steam system
    D. Rupture of a main steam line
    E. Major rupture of a main feedwater pipe
    F. Steam generator tube rupture
    G. Small break loss of coolant accident
    The AFW system only provides mitigation of the events listed 
above and cannot initiate design-basis accident. Therefore, the 
proposed change in the low-pressure setpoint of the motor-driven AFW 
pump supply line will not result in an increase in the probability 
of a previously analyzed accident. In addition, the proposed change 
does not affect the overall water supply to the AFW system. Instead, 
the proposed change results in a transfer from the condensate 
storage tanks (CST) to the essential raw cooling water system at a 
slightly higher CST water level, thus enhancing the continuous 
supply of water. Therefore, this change will not result in an 
increase in the consequences of a previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As discussed previously, the AFW system provides only mitigation 
functions. In addition, the proposed change does not affect the 
overall function and operation of the AFW system or its associated 
water supplies. Instead, this change will provide additional 
assurance of the proper operation of the AFW system. Therefore, the 
proposed revision of the low-pressure setpoint of the motor-driven 
AFW pump supply line will not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The TS bases for the AFW system require that AFW be available to 
ensure that the reactor coolant system (RCS) can be cooled down to 
less than 350 degrees Fahrenheit from normal operating conditions in 
the event of a total loss of offsite power. In addition, the TS 
bases for the CST require that a minimum water volume be available 
to maintain the RCS at hot standby condition for two hours with 
steam discharge to the atmosphere concurrent with a total loss of 
offsite power.
    The proposed TS revision does not affect the overall operation 
of either the AFW system or the CST. The proposed setpoint revision 
does slightly reduce the usable volume of water in the CST. However, 
sufficient margin remains to ensure compliance with the bases of the 
SQN TSs.
    Therefore, the proposed changes to the SQN TSs do not involve a 
reduction in the margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 8, 1994 TS 93-14
    Description of amendment request: The proposed change would revise 
the setpoints in Technical Specification Table 3.3-4, ``Engineered 
Safety Feature Actuation System Instrumentation,'' for the pressure 
switches used to control switchover of the motor-driven Auxiliary 
Feedwater pump suction from the normal condensate storage tank supply 
to the essential raw cooling water supply. The setpoints would be 
changed from the present trip setpoint of 2 psig and allowable value of 
1 psig, to new values of 3.21 psig and 2.44 psig, respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The auxiliary feedwater (AFW) system is designed to mitigate the 
effects of the design basis accidents and anticipated operational 
transients listed below:
    A. Loss of normal feedwater
    B. Loss of offsite power to station auxiliaries
    C. Accidental depressurization in the main steam system
    D. Rupture of a main steam line
    E. Major rupture of a main feedwater pipe
    F. Steam generator tube rupture
    G. Small break loss of coolant accident
    The AFW system only provides mitigation of the events listed 
above and cannot initiate design-basis accident. Therefore, the 
proposed change in the low pressure setpoint of the motor-driven AFW 
pump supply line will not result in an increase in the probability 
of a previously analyzed accident. In addition, the proposed change 
does not affect the overall water supply to the AFW system. Instead, 
the proposed change results in a transfer from the condensate 
storage tanks (CST) to the essential raw cooling water system at a 
slightly higher CST water level, thus enhancing the continuous 
supply of water. Therefore, this change will not result in an 
increase in the consequences of a previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As discussed previously, the AFW system provides only mitigation 
functions. In addition, the proposed change does not affect the 
overall function and operation of the AFW system or its associated 
water supplies. Instead, this change will provide additional 
assurance of the proper operation of the AFW system. Therefore, the 
proposed revision of the low-pressure setpoint of the motor-driven 
AFW pump supply line will not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The TS bases for the AFW system require that AFW be available to 
ensure that the reactor coolant system (RCS) can be cooled down to 
less than 350 degrees Fahrenheit from normal operating conditions in 
the event of a total loss of offsite power. In addition, the TS 
bases for the CST require that a minimum water volume be available 
to maintain the RCS at hot standby condition for two hours with 
steam discharge to the atmosphere concurrent with a total loss of 
offsite power.
    The proposed TS revision does not affect the overall operation 
of either the AFW system or the CST. The proposed setpoint revision 
does slightly reduce the usable volume of water in the CST. However, 
sufficient margin remains to ensure compliance with the bases of the 
SQN TSs.
    Therefore, the proposed changes to the SQN TSs do not involve a 
reduction in the margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of amendment request: February 9, 1994 TS 93-21
    Description of amendment request: The proposed change would revise 
Technical Specification Table 3.3-11, ``Fire Detection Instruments,'' 
by adding one detector to Fire Zones 184, 185, 186, and 187 for each 
Unit. These fire zones are located in the 6.9 kv shutdown board room 
corridors in the auxiliary building.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). The operation of Sequoyah Nuclear Plant (SQN) in 
accordance with the proposed amendment will not:
    1.Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the fire detection instrumentation adds 
two additional cross-zone detectors in each of the Units 1 and 2 
6900-volt shutdown board room corridors on Elevation 734 of the 
auxiliary building. The additional fire detection instrumentation 
provides additional assurance that the fire
    1detection instrumentation will operate as required in the event 
of a fire. However, neither the fire detection instrumentation nor 
the equipment associated with this instrumentation is considered to 
be the source of an accident. In addition, this equipment is not 
taken credit for in the safety analysis. Therefore, there is no 
increase in the probability or consequences of a previously 
evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The fire detection and/or suppression functions affected by this 
change enhance fire mitigation functions only and do not result in a 
change in plant functions. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any preciously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The equipment functions affected by the proposed changes are not 
assumed for any accident in the SQN safety analysis and are not an 
input to the TS margin of safety. Therefore, the proposed change 
will not result in a reduction in the margin of safety.
    The NRC 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: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of amendment request: January 31, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.1.1.2 to permit the reduction of 
boron concentration of water within the reactor coolant system (RCS), 
subject to certain restrictions, when the reactor is in Mode 5 and RCS 
flow is less than 2800 gpm. The proposed amendment is related to 
Amendment No. 176, which was issued by the NRC on December 8, 1992, and 
incorporated a similar revision for Mode 6 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:The Nuclear Regulatory 
Commission has provided standards in 10 CFR 50.92(c) for determining 
whether a significant hazard exists due to a proposed amendment to an 
Operating License for a facility. A proposed amendment involves no 
significant hazards if operation of the facility in accordance with the 
proposed changes would: (1) Not involve a significant increase in the 
probability or consequences of an accident previously evaluated; (2) 
Not create the possibility of a new or different kind of accident from 
any accident previously evaluated; or (3) Not involve a significant 
reduction in a margin of safety. Toledo Edison has reviewed the 
proposed change and determined that a significant hazards consideration 
does not exist because operation of the Davis-Besse Nuclear Power 
Station, Unit Number 1, in accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
conditions or assumptions are significantly affected by the proposed 
changes. The proposed change to Technical Specification (TS) 3/
4.1.1.2 would revise an exception to make it applicable in Mode 5 as 
well as Mode 6. The revised exception would allow water of a lower 
boron concentration than the Reactor Coolant System (RCS) to be 
added to the RCS, with the flowrate of reactor coolant through the 
RCS less than 2800 gpm, provided that the water to be added meets 
the requirements of TS 3.1.1.1 (Mode 5) or TS 3.9.1 (Mode 6). TS 
3.1.1.1 requires that in Mode 5, the boron concentration of the RCS 
be maintained such that the Shutdown Margin shall be less than or 
equal to one percent delta k/k. TS 3.9.1 requires that in Mode 6, 
the boron concentration of all filled portions of the RCS and the 
refueling canal shall be maintained uniform and sufficient to ensure 
that the more restrictive of two reactivity conditions is met. If 
the RCS meets these reactivity condition requirements, and water is 
added to the RCS that also meets the reactivity condition 
requirements of TS 3.1.1.1 or TS 3.9.1, then the RCS is assured to 
remain in compliance with the reactivity condition requirements. The 
possibility that the added water may be of lower boron concentration 
than the RCS, therefore, does not significantly increase the 
probability of an accident previously evaluated.
    The proposed change to TS 3/4.9.8.1 makes TS 3/4.9.8.1 and TS 3/
4.9.8.2 consistent with the current TS 3/4.1.1.2, and is considered 
to be administrative in nature.
    The proposed changes to TS Bases 3/4.1.1.2 and TS Bases 3/4.9.8 
are considered to be administrative in nature.
    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. As discussed in 
item 1a. above, the proposed revision of the exception to TS 3/
4.1.1.2 will not cause a condition that would result in the RCS not 
meeting the requirements of TS 3.1.1.1 or TS 3.9.1, as applicable. 
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. As also discussed in item 1a. above, the 
proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases 
3.4.9.8 are considered to be administrative in nature.
    2a. Not create the possibility of a new kind of accident from 
any accident previously evaluated because no new accident initiators 
or assumption are introduced by the proposed changes. The proposed 
changes do not alter any accident scenarios. As discussed in item 
1a. above, the proposed revision of the exception to TS 3/4.1.1.2 
will not cause a condition that would result in the RCS not meeting 
the requirements of TS 3.1.1.1 or TS 3.9.1. The proposed changes to 
TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases 3/4.9.8 are 
considered to be administrative in nature. None of the proposed 
changes creates the possibility of a new kind of accident.
    2b. Not create the possibility of a different kind of accident 
from any accident previously evaluated because no different accident 
initiators or assumptions are introduced by the proposed changes. 
The proposed changes do not alter any accident scenarios. As 
discussed in item 1a. above, the proposed revision of the exception 
to TS 3/4.1.1.2 will not cause a condition that would result in the 
RCS not meeting the requirements of TS 3.1.1.1 or TS 3.9.1. The 
proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1, and TS Bases 
3/4.9.8 are considered to be administrative in nature. None of the 
proposed changes creates the possibility of a different kind of 
accident from any accident previously evaluated.
    3. Not involve a significant reduction in the margin of safety 
because the proposed change to TS 3/4.1.1.2, as described above, 
will not cause a condition that would result in the RCS not meeting 
the requirements of TS 3.1.1.1 or TS 3.9.1. The margin of safety 
will be maintained by adhering to the limits specified in these TSs. 
The proposed changes to TS Bases 3/4.1.1.2, TS 3/4.9.8.1 and TS 
Bases 3/4.9.8 are considered to be administrative in nature.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: John N. Hannon

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: July 14, 1993

    Description of amendment request: The proposed amendment would 
modify Sections 3.6 and 4.6 of the Technical Specifications to add 
Reactor Coolant System leakage detection requirements to address 
Generic Letter 88-01, ``NRC Position on Intergranular Stress Corrosion 
Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping.''
    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 will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed amendment would add a more conservative 
requirement into the plant Technical Specifications, in addition to 
those that presently exist. Hence, approval of this change will have 
no affect on any previously evaluated accident scenario.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No physical changes are being made to the plant and now 
new operating techniques or procedures are being proposed. The 
proposed amendment would add an additional Limiting Condition for 
Operation and an increased Surveillance
    Requirement to plant Technical Specifications. Hence, approval 
of this change will not create the possibility of a new or different 
kind of accident.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety. The proposed change adds more 
restrictive requirements into the Technical Specifications. Hence, 
approval of this change would not reduce the margin of 
safety.GI21The Commission has also provided guidance concerning the 
application of these standards by providing certain examples (March 
6, 1986, 51FR7751). An example of an amendment that is considered 
not likely to involve a significant hazards consideration is Example 
(ii) which is an additional limitation, restriction or control not 
presently included in the Technical Specifications. This proposed 
amendment provides for an additional Limiting Condition of Operation 
and an increased Surveillance Requirement in the plant Technical 
Specifications. Therefore, based on the above, it is determined this 
change does not constitute a significant hazards consideration as 
defined in 10CFR 50.92(c).
    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: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624
    NRC Project Director: Walter R. Butler

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Units No. 1 and No. 2, Surry County, Virginia

    Date of amendment request: December 27, 1993
    Description of amendment request: The proposed change would revise 
the Technical Specifications (TS) for the Surry Power Station, Units 
No. 1 and No. 2 (SPS-1&2). The proposed changes revise the review 
responsibilities of the Station Nuclear Safety and Operating Committee 
(SNSOC) and the Management Safety Review Committee (MSRC).
    The SPS-1&2 TS address the organization and responsibilities of 
both the onsite and offsite review groups: SNSOC and MSRC, 
respectively. The responsibilities of the SNSOC include the review of 
new procedures and changes to procedures that affect nuclear safety. 
The MSRC review responsibilities include the review of safety 
evaluations and SNSOC meeting minutes and reports. The extent of these 
review activities would be revised by the proposed changes to ensure 
the two review groups are focusing on nuclear safety issues and not 
spending an unnecessary amount of time on activities of minimal safety 
significance. Specifically, the proposed changes would revise the 
review responsibilities of SNSOC regarding procedure changes. Rather 
than reviewing all procedure changes, SNSOC would only review procedure 
changes that require a safety evaluation. The proposed changes also 
would revise the review responsibilities of the MSRC. Rather than 
reviewing all of the safety evaluations and SNSOC meeting minutes and 
reports as presently required by the TS, the MSRC would only review a 
representative sample of these documents.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    [T]he elimination of the SNSOC review of procedure changes that 
do not require a safety evaluation, revising the wording for 
approval of procedure changes, and the modification of the MSRC's 
duties regarding their review of safety evaluations and SNSOC 
meeting minutes and reports will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated. As administrative 
changes, the proposed Technical Specifications changes have no 
direct or indirect effect on accident precursors. No plant 
modifications are being implemented and operation of the plant is 
unchanged. SNSOC review of new procedures and procedure changes that 
require a safety evaluation ensures that activities that could 
affect nuclear safety are being properly reviewed. The MSRC's 
overview of representative samples of safety evaluations and SNSOC 
meeting minutes and reports based on performance ensures these 
programs are being properly implemented and nuclear safety is not 
being compromised; or
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated since physical modifications 
are not involved and systems and components will be operated as 
before the change. The proposed changes are wholly administrative in 
nature and have no impact on plant operations or accident 
considerations. These changes modify the scope of SNSOC review of 
procedure changes and MSRC's review functions concerning safety 
evaluations and SNSOC meeting minutes and reports. Procedure changes 
will continue to receive management review in accordance with 
administrative procedures, however, only changes that require a 
safety evaluation will require SNSOC approval. MSRC review of 
representative samples of safety evaluations and SNSOC meeting 
minutes and reports based on performance will continue to provide 
adequate assurance that nuclear safety is being properly considered; 
or
    3. Involve a significant reduction in a margin of safety as 
defined in the basis of any Technical Specification since the 
responsibilities of the SNSOC and MSRC are not addressed by the 
existing Technical Specification Bases, nor are review requirements 
for procedures. The proposed changes are administrative in nature 
and have no impact on, nor were they considered in, existing UFSAR 
accident analyses. Safety significant procedure changes, i.e., 
changes that require a safety evaluation to be prepared, will 
continue to be reviewed by SNSOC, as will new procedures. Procedure 
changes still require cognizant management approval and preparation 
of an activity screening to determine whether or not the change 
impacts nuclear safety. This ensures activities important to nuclear 
safety are being appropriately reviewed. The effectiveness of the 
safety evaluation program, and the thoroughness of SNSOC meetings 
and reports will be assured through the MSRC's plant overview 
function which is based on observed performance.
    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

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of amendment request: December 6, 1993
    Description of amendment request: The proposed amendment would 
modify the testing requirements for the Main Steam Relief Valves 
(MSRVs) in the Technical Specifications (TS). Specifically, the 
proposed amendment would allow deferral of MSRV Position Indication 
Channel Calibration, including the Channel Functional Test that is the 
focus of the request, for 24 hours after the plant reaches conditions 
that would allow the Channel Functional Test to be conducted under 
operating conditions (above 10% rated reactor power). In addition, the 
proposed change would extend the current deferral for two related TS, 
one for MSRVs and the other for Automatic Depressurization System (ADS) 
valves, from 12 hours to the proposed 24 hours after reaching 
conditions that would allow the Channel Functional Test to be conducted 
under operating conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The staff's evaluation of the licensee's analysis of the 
change that would defer Safety/Relief Valve Position Indicators Channel 
Calibration, specifically the Channel Functional Test, is presented 
below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No credit is taken for the MSRV position indication in the 
initiation or mitigation of any analyzed accident. The inoperability of 
valve position indication does not affect the manual or automatic 
actuation of the MSRVs. The analysis for inadvertent opening of an MSRV 
(FSAR Section 15.1.4) assumes that the alarm function of the MSRV 
discharge line temperature sensors and Reactor Pressure Vessel (RPV) 
level control systems provide the signals for manual and automatic 
system actuation, respectively. Therefore, because previously analyzed 
accidents are mitigated without the use of the MSRV position 
indication, this change that would allow temporary plant operation in 
modes 1 and 2 with inoperable MSRV position indication to allow testing 
does not increase the probability or consequences of a previously 
analyzed accident.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No new mode of operation of any equipment results from the delay of 
the Channel Functional surveillance. The valve position indication 
provides information for operator response to previously evaluated 
accidents, and does not provide any automatic system actuations that 
could initiate an accident or abnormal operating occurrence sequence. 
This change would not, therefore, create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The margin of safety involved in this amendment is the time it 
takes to identify inadvertent MSRV operation to initiate either 
automatic or manual plant response. Since no credit is taken for MSRV 
position indication in the WNP-2 safety analyses for initiation of 
automatic or operator manual response, the lack of MSRV position 
indication for a 24 hour time period will not affect the analyzed time 
to identify inadvertent MSRV operation. The proposed change does not, 
therefore, affect the margin of safety.
    The staff's evaluation of the licensee's analysis of the change 
that would allow an increase in the current deferral of Safety/Relief 
Valve and Automatic Depressurization System (ADS) valve testing of 12 
hours to the proposed 24 hours is presented below:
    1. Does the amendment involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change represents an additional 12 hours in which the 
operability of the MSRV position indication would not be verified, and 
could be inoperable. No credit is taken for the MSRV position 
indication in the initiation or mitigation of any analyzed accident. 
The inoperability of valve position indication does not affect the 
manual or automatic actuation of the MSRVs as discussed in the 
following: The analysis for inadvertent opening of an MSRV (FSAR 
Section 15.1.4) assumes the alarm function of the MSRV discharge line 
temperature sensors and RPV level control systems for manual and 
automatic system actuation, respectively. Thus, the inoperability of 
the MRSV position indication for an additional 12 hours would not 
affect the probability or consequences of an accident previously 
evaluated.
    The proposed change represents an additional 12 hours in which the 
operability of the ADS valves, which are also the MSRVs, would not be 
verified, and could be inoperable. The cause of any potential 
inoperability would also be undetermined, and any unidentified failure 
mode affects the uncertainties assumed in the probability of 
inadvertent MSRV opening, which is an analyzed accident (FSAR Section 
15.1.4). In the worst case, the probability of inadvertent MSRV opening 
would be increased due to an unidentified problem with the MSRVs. The 
additional 12 hours, considering plant operation for approximately 5800 
hours of power operation each year (assuming 60 day annual refueling 
outage and 80% capacity factor for the remaining time), represents only 
an estimated 0.2% increase in time at power without verifying 
operability of the ADS valves, which contributes a very small increase 
in the probability of an inadvertent MSRV opening due to an 
undiscovered fault condition. Performing the surveillance testing using 
other possible methods, either removal of the MSRV position indication 
from the valve and testing separately, or testing in modes 3, 4, or 5 
such that adequate steam back pressure does not exist, introduce other 
uncertainties and potential for valve damage that would, using 
engineering judgement, create a greater increase in probability of an 
inadvertent MSRV opening than the additional 12 hours would contribute. 
The proposed change would not, therefore, significantly increase the 
probability of an accident previously evaluated.
    Regarding the potential effect on the consequences of an 
inadvertent MSRV opening, the accident analysis in FSAR Section 15.1.4 
analyzes consequences assuming an MSRV sticks open. Thus, increasing 
the time that a potential malfunction of a valve may go undetected does 
not affect the consequences of an MSRV that is already considered open 
as assumed in the accident analysis. Consequently, the increase in time 
allowed for conducting the surveillance testing, and therefore the time 
that the operability of the ADS-related MSRV is not verified does not 
affect the consequences of an accident previously evaluated.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No new or modified mode of operation of any structure, system, or 
component results from delaying verification of the status of 
operability of the MSRVs for an additional 12 hours. The proposed 
change does not, therefore, create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the amendment involve a significant reduction in a margin 
of safety?
    The margin of safety involved in this proposed amendment is the 
time, established by the current TS, that operation in modes 1 and 2 is 
allowed with the operability of the MSRVs and associated ADS valves 
undetermined. The current TS allow 12 hours to conduct testing to 
verify operability of these valves. The proposed TS would extend that 
time to 24 hours. The additional 12 hours, considering plant operation 
for approximately 5800 hours of power operation each year (assuming 60-
day annual refueling outage and 80% capacity factor for the remaining 
time), represents only an estimated 0.2% increase in time at power 
without verifying operability of the ADS valves, which the staff 
considers, by engineering judgement, as a very small reduction in the 
margin to safety affected by the proposed change. This change would, 
therefore, not involve a significant reduction in the margin of safety 
provided by the current TS.
    Based on this review, it appears that the three standards of 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

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 amendment request: January 26, 1994
    Description of amendment request: The proposed amendments would 
change Technical Specification Section 15.3.0, ``General 
Considerations.'' This section specifies the actions which should be 
taken for conditions not directly addressed in the action statements of 
the Technical Specifications. The changes would provide more 
operational flexibility. Changes to the applicable bases and editorial 
changes are also proposed.
    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 this facility under the proposed Technical 
Specifications change will not create a significant increase in the 
probability or consequences of an accident previously evaluated. 
This proposed change will incorporate requirements contained in 
NUREG 1431, Revision 0, ``Westinghouse Owner's Group Improved 
Technical Specifications,'' into the Point Beach Technical 
Specifications Section 15.3.0, ``General Considerations.'' The 
proposed revisions will not remove any existing requirements. 
Several new requirements will be added. However, the proposed 
revisions will allow a longer period of time for the affected 
unit(s) to be placed in hot shutdown should one of the applicable 
Limiting Conditions for Operation not be met. This longer time limit 
is identical to the time limit specified in NUREG 1431, Revision 0, 
and permits the shutdown to proceed in a controlled and orderly 
manner that is well within the capabilities of the unit(s), assuming 
that only the minimum required equipment is operable. This reduces 
thermal stresses on components of the Reactor Coolant System and the 
potential for a plant transient that could challenge plant safety 
systems. The amount of time to reach cold shutdown would decrease 
from 48 hours to 37 hours. The slightly longer time to reach hot 
shutdown is more than offset by the shorter time to cold shutdown, 
thereby reducing the total amount of time a unit may be operated in 
a condition in which a system or component required to mitigate the 
consequences of an accident is unavailable, or that is otherwise 
prohibited by the specifications.
    Should a shutdown of both units be required, 15.3.0.A and 
15.3.0.B allow an orderly and sequential shutdown of each unit to 
take place. This ensures that the plant operators are not overloaded 
during the shutdown process and allows the units shutdowns to 
proceed in a controlled and orderly manner. The revised 
specifications will clarify the requirements, enhancing the 
effectiveness of the Point Beach Technical Specifications. There is 
no physical change to the facility, its systems, or its operation. 
Thus, an increased probability or consequences of an accident 
previously evaluated cannot occur.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This proposed change will incorporate requirements contained in 
NUREG 1431, Revision 0, ``Westinghouse Owner's Group Improved 
Standard Technical Specifications,'' into the Point Beach Technical 
Specifications Section 15.3.0, ``General Considerations.'' The 
proposed revisions will not remove any existing requirements. 
Several new requirements will be added.
    However, the proposed revisions will allow a longer period of 
time for the affected unit(s) to be placed in hot shutdown should 
one of the applicable Limiting Conditions for Operation not be met. 
This longer time limit is identical to the time limit specified in 
NUREG 1431, Revision 0, and permits the shutdown to proceed in a 
controlled and orderly manner that is well within the capabilities 
of the unit(s), assuming that only the minimum required equipment is 
operable. This reduces thermal stresses on components of the Reactor 
Coolant System and the potential for a plant transient that could 
challenge plant safety systems. The amount of time to reach cold 
shutdown would decrease from 48 hours to 37 hours. The slightly 
longer time to reach hot shutdown is more than offset by the shorter 
time to cold shutdown, thereby reducing the total amount of time a 
unit may be operated in a condition in which a system or component 
required to mitigate the consequences of an accident is unavailable, 
or that is otherwise prohibited by the specifications.
    Should a shutdown of both units be required, 15.3.0.A and 
15.3.0.B allow an orderly and sequential shutdown of each unit to 
take place. This ensures that the plant operators are not overloaded 
during the shutdown process and allows the units shutdown to proceed 
in a controlled and orderly manner. The revised specifications will 
clarify the existing specifications and will add some additional 
requirements, enhancing the effectiveness of the Point Beach 
Technical Specifications. There is no physical change to the 
facility, its systems, or its operation. Thus, a new or different 
kind of accident cannot occur.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety. This proposed change will incorporate requirements 
contained in NUREG 1431, Revision 0, ``Westinghouse Owner's Group 
Improved Standard Technical Specifications,'' into the Point Beach 
Technical Specifications Section 15.3.0, ``General Considerations.'' 
The proposed revisions will not remove any existing requirements. 
Several new requirements will be added.
    However, the proposed revisions will allow a longer period of 
time for the affected unit(s) to be placed in hot shutdown should 
one of the applicable Limiting Conditions for Operation not be met. 
This longer time limit is identical to the time limit specified in 
NUREG 1431, Revision 0, and permits the shutdown to proceed in a 
controlled and orderly manner that is well within the capabilities 
of the unit(s), assuming that only the minimum required equipment is 
operable. This reduces thermal stresses on components of the Reactor 
Coolant System and the potential for a plant transient that could 
challenge plant safety systems. The amount of time to reach cold 
shutdown would decrease from 48 hours to 37 hours. The slightly 
longer time to reach hot shutdown is more than offset by the shorter 
time to cold shutdown, thereby reducing the total amount of time a 
unit may be operated in a condition in which a system or component 
required to mitigate the consequences of an accident is unavailable, 
or that is otherwise prohibited by the specifications.
    Should a shutdown of both units be required, 15.3.0.A and 
15.3.0.B allow an orderly and sequential shutdown of each unit to 
take place. This ensures that the plant operators are not overloaded 
during the shutdown process and allows the units shutdowns to 
proceed in a controlled and orderly manner. The revised 
specifications will clarify the existing specifications and will add 
some additional requirements, enhancing the effectiveness of the 
Point Beach Technical Specifications. There is no physical change to 
the facility, its systems, or its operation. Thus, a significant 
reduction in a margin of safety cannot 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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the action 
involved exigent circumstances. They are repeated here because the 
biweekly notice lists all amendments issued or proposed to be issued 
involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: November 4, 1993
    Description of amendments request: The proposed amendments would 
allow the removal of an orifice plate in the containment vent/purge 
line to allow greater flow through the line. The restoration of full 
flow capability will result in less time required to vent the 
containment. A reanalysis of the maximum hypothetical accident, as 
currently described in the Updated Final Safety Analysis Report, was 
performed to support the requested amendments. The results of the 
reanalysis indicate that the consequences of the accident previously 
analyzed would be increased. Although the consequences result in an 
increase in the fission product release, the total doses are well 
within the limits of 10 CFR Part 100, ``Factors to be considered when 
evaluating sites.''Date of Publication of Individual Notice in Federal 
Register: February 25, 1994 (59 FR 9254)
    Expiration date of individual notice: March 28, 1994
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra

Duke Power Company, Docket No. 50-414, Catawba Nuclear Station, 
Unit No. 2, York County, South Carolina

    Date of amendment request: January 10, 1994
    Description of amendment request: The proposed amendment would 
change the method of measuring the reactor coolant system flow rate 
(Technical Specification 2.0 and 3/4.2) during the 18-month 
surveillance for Catawba, Unit 2. Date of publication of individual 
notice in Federal Register: March 1, 1994 (59 FR 9785)
    Expiration date of individual notice: March 31, 1994
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
Creek Generating Station, Coffey County, Kansas

    Date of application for amendment: February 7, 1994
    Brief description of amendment request: The proposed amendment 
would allow an increase in reactor coolant temperature in order to 
support operation at the rated thermal power of 3565 megawatts thermal 
(MWt). The proposed amendment would change reactor protection system 
setpoints by increasing the nominal reactor coolant average temperature 
from 581.2 deg.F to 586.5 deg.F, changing the axial flux difference 
penalties and setpoint uncertainty allowances. The proposed amendment 
also increases the maximum indicated reactor coolant system average 
temperature from 585.0 deg.F to 590.5 deg.F.
    Date of individual notice in Federal Register: February 15, 1994 
(59 FR 7269)
    Expiration date of individual notice: March 17, 1994
    Local Public Document Room location: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

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.

Baltimore Gas and Electric Company, Docket No. 50-318, Calvert 
Cliffs Nuclear Power Plant, Unit No. 2, Calvert County, Maryland

    Date of application for amendment: November 1, 1993, as 
supplemented on February 1, 1994
    Brief description of amendment: The amendment revises the heatup 
and cooldown curves which will allow operation beyond the current 12 
effective full-power years (EFPY) to approximately 13.8 EFPY. The 
increase in this EFPY will allow Unit 2 to operate until its next 
refueling outage (RFO-10) in accordance with the requirements of 10 CFR 
Part 50, Appendix G.
    Date of issuance: March 1, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 162
    Facility Operating License No. DPR-69: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62150)The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: August 27, 1993, as supplemented 
November 10, 1993 and February 1, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications and allows elimination of the existing reactor coolant's 
resistance temperature detector (RTD) bypass manifold system and the 
substitution of a new design with RTDs mounted in thermowells that 
extend directly into the flow stream of the reactor coolant system.
    Date of issuance: February 18, 1994
    Effective date: February 18, 1994
    Amendment No. 43
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48379)The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 18, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: July 16, 1993, as supplemented 
February 17, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) 3/4.2.1, Axial Flux Difference, 3/4.2.2, Heat Flux 
Hot Channel Factor (FQ), deletes surveillance requirement 4.2.2.2 
and 4.2.2.3, and changes 6.9.1.6, Core Operating Limit Report, and 
associated Bases related to the transition from nuclear fuel supplied 
by Westinghouse to nuclear fuel supplied by Siemens Power Corporation 
(SPC) beginning with Cycle No. 6.
    Date of issuance: March 1, 1994
    Effective date: March 1, 1994
    Amendment No. 44
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41500)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: May 15, 1993, as supplemented 
February 17 and February 25, 1994
    Brief description of amendment: The amendment modifies the SHNPP 
Operating License to provide for a one-time exemption from compliance 
with License Condition 2.C.(8) which requires periodic emergency diesel 
generator engine teardowns for component inspections.
    Date of issuance: March 3, 1994
    Effective date: March 3, 1994
    Amendment No. 45
    Facility Operating License No. NPF-63. Amendment revises the 
Operating License Condition 2.C.(8) as defined in Attachment 1 to 
Operating License NPF-63.
    Date of initial notice in Federal Register: June 9, 1993 (58 FR 
32378) The February 17, 1994, letter provided supplemental information 
and the February 25, 1994, letter modified the May 15, 1993, letter and 
requested a one-time exemption. Neither supplemental letter changed the 
initial proposed determination of no significant hazards consideration 
as published in the Federal Register.The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
March 3, 1994.No significant hazards consideration comments received: 
No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Carolina Power & Light Company, et al., Docket No. 50-400, Shearon 
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, 
North Carolina

    Date of application for amendment: July 16, 1993, as supplemented 
February 17, 1994
    Brief description of amendment: The amendment revises the SHNPP 
Technical Specification to incorporate changes to reactor core safety 
limits, reactor trip system instrumentation setpoints, power 
distribution limits, and shutdown boron concentration control in 
support of the transition from nuclear fuel supplied by Westinghouse 
Electric Corporation to nuclear fuel supplied by Siemens Power 
Corporation and a reactor core safety average temperature reduction 
effort.
    Date of issuance: March 3, 1994
    Effective date: March 3, 1994
    Amendment No. 46
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50966) The February 17, 1994, letter provided clarifying information 
that did not change the initial determination of no significant hazards 
consideration as published in the FEDERAL REGISTER.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 3, 1994. No significant hazards consideration comments 
received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: August 13, 1993, as 
supplemented by letters dated September 15, 1993, September 16, 1993, 
December 17, 1993, January 19, 1994, February 11, 1994, and February 
24, 1994.
    Brief description of amendments: These amendments revise Technical 
Specification 3/4.4.5, ``Steam Generators,'' to allow sleeving of 
defective steam generator tubes as an alternative to tube plugging. Two 
different methods of sleeving are approved for Byron and Braidwood 
stations: Westinghouse laser-welded tube sleeving and Babcock and 
Wilcox (B&W) kinetic-welded tube sleeving.
    Date of issuance: March 4, 1994
    Effective date: March 4, 1994
    Amendment Nos.:  47, 47, 59, and 59
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57846) The additional information contained in the supplemental letters 
dated September 15, 1993, September 16, 1993, December 17, 1993, 
January 19, 1994, February 11, 1994, and February 24, 1994, served to 
clarify the amendments, were within the scope of the initial notice, 
and did not affect the Commission's proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 4, 1994.No 
significant hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: September 2, 1993, as 
supplemented by letters dated January 7, 1994, and February 10, 1994.
    Brief description of amendments: The amendments revise the Byron 
and Braidwood Technical Specifications (TS) to allow replacement of the 
125 volt DC Gould batteries with new 125 volt DC AT&T batteries and 
rephrase the specification for their design duty cycle. In addition, 
the amendments revise the crosstie loading limitations and crosstie 
breaker limitations. The associated TS Bases are also revised to 
include the purpose for the crosstie limitations and a discussion of 
the design duty cycle requirements.
    Date of issuance: March 4, 1994
    Effective date: March 4, 1994
    Amendment Nos.:  59, 59, 47, and 47
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4936) The February 10, 1994, supplemental submittal did not change the 
original no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 4, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: December 15, 1989, as 
supplemented September 16, 1992.
    Brief description of amendments: The amendments revise Technical 
Specifications (TSs) 3.0.4, 4.0.3, and 4.0.4 to incorporate guidance 
provided by the NRC in Generic Letter 87-09, ``Sections 3.0 and 4.0 of 
the Standard Technical Specifications (STS) on the Applicability of 
Limiting Conditions for Operation and Surveillance Requirements.''
    Date of issuance: February 24, 1994
    Effective date: February 24, 1994

    Amendment Nos.:  94 and 78
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 12, 1992 (57 
FR 53785)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 24, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: January 28, 1994
    Brief description of amendments: The amendments minimize 
unnecessary testing for certain instruments in the Reactor Protection 
System and the End-of Cycle Recirculation Pump Trip system for LaSalle 
County Station, Units 1 and 2 technical specifications.
    Date of issuance: February 25, 1994
    Effective date: February 25, 1994

    Amendment Nos.:  95 and 79
    Facility Operating License Nos. NPF-11 and NPF-18. The amendments 
revised the Technical Specifications.Public comments requested as to 
proposed no significant hazards consideration: No (59 FR 6062 dated 
February 9, 1994). That notice provided an opportunity to submit 
comments on the Commission's proposed no significant hazards 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by March 11, 
1994, but indicated that if the Commission makes a final no significant 
hazards consideration determination any such hearing would take place 
after issuance of the amendment. The Commission's related evaluation of 
the amendment, finding of exigent circumstances, and final 
determination of no significant hazards consideration is contained in a 
Safety Evaluation dated February 25, 1994.
    Attorney to licensee: Michael I. Miller, Esquire; Sidley & Austin, 
One First National Plaza, Chicago, Illinois 60690.
    Local Public Document Room location: Public Library of Illinois 
Valley Community College, Rural Route No. 1, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments: November 4, 1993
    Brief description of amendments: The amendment revises the 
Technical Specifications surveillance frequency and acceptance criteria 
requirements for the steam generator safety valves.
    Date of issuance: March 2, 1994
    Effective date: March 2, 1994

    Amendment Nos.:  154 and 142
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64605)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 2, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan

    Date of application for amendment: May 24, 1993
    Brief description of amendment: The amendment revises Technical 
Specification 4.6.2.1 to allow a one-time relief from the requirement 
to perform accelerated Type A integrated leak rate tests (ILRT) after 
two consecutive tests fail to meet the acceptance criteria. 
Concurrently, the Commission granted a one-time exemption from the 
requirement in 10 CFR Part 50, Appendix J, III.A.6.(b) to perform the 
Type A containment ILRTs on an accelerated frequency following failure 
of two previous Type A tests.
    Date of issuance: February 22, 1994
    Effective date: February 22, 1994, with full implementation within 
45 days.

    Amendment No.:  97
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43925)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 22, 1994 No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, 
Appling County, Georgia

    Date of application for amendments: October 19, 1993
    Brief description of amendments: The amendments revise both the 
surveillance requirements of the Unit 1 Technical Specification (TS) 
3.9.D and Unit 2 TS 3/4.8.2, ``Electrical Power Monitoring for Reactor 
Protection System,'' to add time delays to the reactor protection 
system electrical power monitoring trips.
    Date of issuance: February 24, 1994
    Effective date: To be implemented within 60 days from the date of 
issuance.

    Amendment Nos.:  191 and 130
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67846)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 24, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-321 and 50-366, Edwin I. Hatch Nuclear Plant, Units 1 and 2, 
Appling County, Georgia

    Date of application for amendments: November 9, 1993
    Brief description of amendments: The amendments revise the 
Surveillance Requirements of Hatch Unit 1 Technical Specification (TS) 
Section 4.9, ``Auxiliary Electrical Systems,'' and Hatch Unit 2 TS 
Section 4.8, ``AC Sources - Operating,'' to change the requirements for 
diesel generator testing under hot initial conditions.
    Date of issuance: February 24, 1994
    Effective date: To be implemented within 60 days from the date of 
issuance.

    Amendment Nos.:  192 and 131
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67846)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 24, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Energy Operations, Inc., Docket No. 50-458, River Bend Station, 
Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: December 8, 1993, as supplemented by 
letter dated February 3, 1994.
    Brief description of amendment: The amendment revises the River 
Bend Station, Unit 1 technical specifications to permit extending 
certain surveillance intervals so that testing can be performed during 
the refueling outage scheduled to start April 16, 1994, rather than 
requiring an earlier shutdown solely to perform the testing.
    Date of issuance: February 18, 1994
    Effective date: February 18, 1994

    Amendment No.:  Amendment No. 72
    Facility Operating License No. NPF-47: The amendment revised the 
technical specifications.
    Date of initial notice in Federal Register: January 18, 1994 (59 FR 
2630) The February 3, 1994, letter provided clarifying information and 
did not change the initial no significant hazards consideration 
determination.The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 18, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

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: February 1, 1990, as supplemented by 
letters dated November 27, 1990, June 5, 1991, November 3, 1992, 
November 11, 1992, August 16, 1993, October 22, 1993, November 5, 1993 
(two letters), and November 29, 1993.
    Brief description of amendments: The amendments consist of changes 
to the technical specifications (TS) to change the allowed outage times 
(AOTs) and/or surveillance test intervals (STIs) as a result of the 
South Texas probabilistic safety assessment (PSA) considered in 
conjunction with other information. Ten of the TS changes were based on 
changes to the core damage frequency as calculated using the PSA. Five 
additional TS changes to the AOTs and STIs were not specifically 
modeled in the PSA, but had little or no impact on risk.
    Date of issuance: February 17, 1994
    Effective date: February 17, 1994, to be implemented within 45 days 
of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 59; Unit 2 - Amendment No. 
47
    Facility Operating License Nos. NPF-76 and NPF-80: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 21, 1990 (55 FR 
10535) The supplemental letters provided additional clarifying 
information, were within the scope of the original application and did 
not change the original proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 17, 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

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile 
Point Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: December 14, 1993
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 3/4.8.2, ``DC Sources,'' to delete two notes which 
indicated that two 125-volt full capacity battery chargers were 
required when the Uninterruptible Power Supply was powered by its 
backup DC power supply. These notes applied to the Divisions I and II 
DC sources during operating and shutdown conditions. The amendment also 
revises TS 3/4.8.2 to increase the minimum allowable electrolyte 
temperature for the 125-volt batteries from 60  deg.F to 65  deg.F. In 
addition, the amendment makes administrative changes to TS 4/3.8.4, 
``Electrical Equipment Protective Devices,'' and the TS Bases.
    Date of issuance: March 2, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 55
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2868) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 2, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
Connecticut

    Date of application for amendment: May 14, 1993
    Brief description of amendment: The amendment modifies the 
Technical Specifications relating to the spent fuel pool (SFP) by 
removal of the cell blockers in Region C, thus increasing by 234 fuel 
assemblies the storage capacity of the SFP. To accommodate the 
reactivity requirements, the required burnup of fuel in Region C has 
been increased and neutron absorbing (poison) rodlets (pins) are 
required to be introduced in fuel assemblies not meeting the maximum 
burnup requirements for fuel assemblies without rodlets.
    Date of issuance: March 1, 1994
    Effective date: As of the date of issuance to be implemented 
within30 days.
    Amendment No.: 172
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 10, 1993 (58 FR 
42581) The submittals of November 30, 1993, December 1, 1993, and 
January 27, 1994, provided clarifying information that did not change 
the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-313, 
DiabloCanyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County, California

    Date of application for amendments: December 22, 1992 (Reference 
LAR 92-09)
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2. Specifically, TS Section 2.2, ``Reactor Trip System 
InstrumentationSetpoints,'' would be revised to change the 
Overtemperature Delta-R reactor trip setpoint as a result of a non-
conservativsm in the Westinghouse methodology used to calculate the 
f(delta I) penalty function.
    Date of issuance: February 24, 1994
    Effective date: During 6th refueling outage for Units 1 and 2, 
March andSeptember 1994, respectively
    Amendment Nos.: 88 and 87
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 3, 1993 (58 FR 
7003)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 24, 1974.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, 
Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis 
Obispo County,California

    Date of application for amendments: January 10, 1994, as 
supplemented February 3, 1994 (Reference LAR 94-01)
    Brief description of amendments: The proposed amendments revise the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant Unit Nos. 1 and 2 to change TS 3/4.3.2, ``Engineered Safety 
Features Actuation System Instrumentation,'' and TS 3/4.6.2.3, 
``Containment Cooling System.'' Specifically, TS 3/4.3.2, Table 3.3-3, 
``Engineered Safety Features Actuation System Instrumentation,'' and 
Table 4.3-2, ``Engineered Safety Features Actuation System 
Instrumentation Surveillance Requirements,'' are revised to include 
Mode 4 applicability requirements for the high-high containment 
pressure signal. In addition, TS 3/4.6.2.3 is revised to clarify 
acceptable containment fan cooling unit (CFCU) configurations that 
satisfy the safety analysis requirements and to clarify the minimum 
required component cooling water flow supplied to the CFCU cooling 
coils.
    Date of issuance: March 2, 1994
    Effective date: March 2, 1993
    Amendment Nos.: 89 and 88
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1994 (59 FR 
4121)The February 3, 1994 Federal Register submittal provided 
clarifying information and did not affect the no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 2, 1994.No 
significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: May 21, 1993
    Brief description of amendment: The amendment revises the TS 
Surveillance Requirements and associated Bases for the emergency diesel 
generator (EDG) fuel oil transfer system and the EDG air starting 
compressors to clarify that testing of these systems/components can be 
conducted either concurrently or independently of the monthly EDG 
tests. The changes would also clarify the Bases for EDG fuel quality 
testing and make an editorial change.
    Date of issuance: February 23, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 205
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36443)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 23, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: December 22, 1993
    Brief description of amendment: The amendment adds Limiting 
Conditions for Operation and Surveillance Requirements to Tables 
3.12.1, ``Water Spray/Sprinkler Protected Areas,'' and 4.12.1, ``Water 
Spray/Sprinkler System Tests,'' and clarifies the associated Bases to 
reflect the installation of a new full-area fire suppression system in 
the east and west cable tunnels. This new full-area fire suppression 
system was installed because the previous sprinkler system did not 
provide coverage to some cable trays and the sprinkler head orientation 
did not provide full coverage of the cable trays where it was 
installed. The amendment also corrects other portions of Tables 3.12.1 
and 4.12.1 to ensure consistency with changes made to reflect the east 
and west cable tunnel modification.
    Date of issuance: February 28, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.

    Amendment No.:  206
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1994 (59 FR 
2634) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 28, 1994. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Public Service Electric & Gas Company, Docket No. 50-354, Hope 
Creek Generating Station, Salem County, New Jersey

    Date of application for amendment: August 30, 1993
    Brief description of amendment: The amendment made the following 
changes:1. Revised Technical Specification (TS) TS Table 3.8.4.1-1 to 
delete Breaker No. 52-263022 which was disconnected and converted to 
spare status by a plant design change, and 2. Revised TS 3.11.2.7 to 
change radioactivity rate units and associated action statement 
reference from HOT STANDBY to read HOT SHUTDOWN, and changed the 
reference name of the Offgas Radioactivity Monitor.
    Date of issuance: February 28, 1994
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment No.: 66
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50974)The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 28, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

    Date of application for amendment: January 25, 1993
    Brief description of amendment: The amendment revised the pressure-
temperature limit curves for heatup, cooldown, hydrostatic tests and 
criticality from 10 effective full power years to 15 effective full 
power years.
    Date of issuance: February 22, 1994
    Effective date: February 22, 1994
    Amendment No.: 129
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2871)The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 22, 1994.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

Sacramento Municipal Utility District, Docket No. 50-312, Rancho 
Seco Nuclear Generating Station, Sacramento County, California

    Date of application for amendment: January 19, 1993, and 
supplemented May 14, and December 22, 1993. The supplemental 
information submitted May 14, and December 22, 1993, did not affect the 
proposed no significant hazards consideration determination.
    Brief description of amendment: This amendment modifies the nuclear 
organization of the Sacramento Municipal Utility District (SMUD) that 
will oversee the operation of Rancho Seco at least through the 
Custodial SAFSTOR phase of the decommissioning of Rancho Seco.
    Date of issuance: February 23, 1994
    Effective date: February 23, 1994
    3Amendment No.: 121Facility Operating License (Possession Only) No. 
DPR-54: This amendment modifies the nuclear organization of SMUD that 
will oversee the operation at Rancho Seco at least through the 
Custodial SAFSTOR phase of decommissioning.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34092). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 23, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Central Library, Government 
Documents 828 I Street, Sacramento, California 95814.

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: September 19, 1990 as 
supplemented on February 26, 1993
    Brief description of amendment: The amendment revised Technical 
Specification (TS) 3.6.1.3 to specify actions in the event a 
containment air lock interlock mechanism becomes inoperable and to 
clarify the limitations on the use of an inoperable air lock. The 
associated bases were also revised per telecon of February 2, 1994.
    Date of issuance: February 23, 1994
    Effective date: February 23, 1994
    Amendment No.: 56
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 15, 1991 (56 FR 
22479) The application for amendment was renoticed on April 14, 1993 
(58 FR 19473). The telecon of February 2, 1994, just changed the Bases 
section of the TS and did not affect the initial proposed finding of no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated February 23, 1994.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa 
County, Virginia

    Date of application for amendments: July 16, 1993, as supplemented 
November 15, 1993
    Brief description of amendments: These amendments implement the 
revised 10 CFR Part 20, Standards for Protection Against Radiation
    Date of issuance: February 17, 1994
    Effective date: February 17, 1994

    Amendment Nos.:  178 and 159
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43937)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 17, 1994 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Virginia Electric and Power Company, et al., Docket Nos. 50-338 and 
50-339, North Anna Power Station, Units No. 1 and No. 2, Louisa 
County, Virginia

    Date of application for amendments: March 18, 1993 and December 9, 
1993
    Brief description of amendments: The amendments revise the NA-1&2 
TS which will allow plant personnel to effect repairs to the Rod 
Control System in an orderly manner while continuing to ensure that the 
control and shutdown banks are capable of performing their designed 
safety function.
    Date of issuance: March 1, 1994
    Effective date: March 1, 1994
    Amendment Nos.:  79 and 160
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 14, 1993 (58 FR 
19492)The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 1, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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 application for amendments: July 2, 1993, as supplemented
    December 10, 1993. The December 10, 1993 letter provided clarifying 
information within the scope of the original amendment application and 
did not change the staff's no significant hazards consideration 
determination.
    Brief description of amendments: These amendments update the 
augmented inspection program for sensitized stainless steel to the 
newer Code requirements.
    Date of issuance: February 18, 1994
    Effective date: February 18, 1994

    Amendment Nos.:  187 and 187
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46241) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 18, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf 
CreekGenerating Station, Coffey County, Kansas

    Date of amendment request: February 7, 1994
    Brief description of amendment: The changes allow an increase in 
reactor coolant temperature in order to support operation at the rated 
thermal power of 3565 megawatts thermal (MWt). The amendment changes 
reactor protection system overtemperature and overpower delta-
temperature setpoints by increasing the nominal reactor coolant 
temperature from 581.2 deg.F to 586.5 deg.F, changing the axial flux 
difference penalties, and changing the setpoint uncertainty allowances. 
The amendment also increases the maximum indicated reactor coolant 
system average temperature of Technical Specification 3/4.2.5, DNB 
Parameters, from 585. deg.F to 590.5 deg.F.
    Date of issuance:March 3, 1994
    Effective date: March 3, 1994, to be implemented within 30 days of 
issuance.
    Amendment No.: Amendment No. 72
    Facility Operating License No. NPF-42. Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: Yes (59 FR 7269 dated February 15, 
1994). That notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by March 17, 1994, but 
indicated that if the Commission makes a final no significant hazards 
consideration determination any such hearing would take place after 
issuance of the amendment.The Commission's related evaluation of the 
amendment, finding of exigent circumstances, and final determination of 
no significant hazards consideration is contained in a Safety 
Evaluation dated March 3, 1994.Local Public Document Room Locations: 
Emporia State University, William Allen White Library, 1200 Commercial 
Street, Emporia, Kansas 66801 and Washburn University School of Law 
Library, Topeka, Kansas 66621
    Dated at Rockville, Maryland, this 7th day March 1994.
    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director of Reactor Projects - I/II
[Doc. 94-5971 Filed 03-15-94; 8:45 am]
BILLING CODE 7590-01-F