[Federal Register Volume 62, Number 98 (Wednesday, May 21, 1997)]
[Notices]
[Pages 27792-27810]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-10521]



NUCLEAR REGULATORY COMMISSION

Biweekly Notice


Applications And Amendments To Facility Operating Licenses 
Involving No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 28, 1997 through May 9, 1997. The last 
biweekly notice was published on May 7, 1997 (62 FR 24984).

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

[[Page 27793]]

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By June 20, 1997, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
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).

[[Page 27794]]

    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: December 27, 1996
    Description of amendments request: The proposed amendments would 
revise Technical Specification (TS) 3.6.1.3.b (peak containment 
internal pressure for the design basis loss of coolant accident (LOCA)) 
from 49.5 psig to 52 psig and the associated Bases Sections. The 
proposed amendments reflect values based on a revised LOCA analysis. 
The LOCA analysis was revised to reflect the maximum primary 
containment internal pressure specified in other TS. This maximum 
primary containment internal pressure was not used in the original LOCA 
analysis.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff's analysis is presented below.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed amendment increases the peak calculated 
containment internal pressure for the design basis LOCA from 49.5 
psig to 52 psig. The maximum pressure occurs following an accident. 
Since the pressure is a consequence of an accident, this change has 
no effect on the probability of accident initiation, and therefore, 
the probability of an accident previously evaluated has not been 
significantly increased.
    The consequences of an accident previously evaluated in the 
Updated Final Safety Analysis Report (UFSAR) will not be 
significantly increased. UFSAR Section 15.6.5.6, ``Analyses of 
Effects and Consequences - Large Break LOCA,'' states that ``It is 
assumed that the containment leaks at the maximum rates allowed by 
the Technical Specifications, i.e., 0.1 vol. %/d for the first 24 
hours and half of that rate thereafter.'' The dose calculation 
assumes that under accident conditions, the release of radionuclides 
to the containment is instantaneously homogenized within the 
containment free air volume. This results in a constant 
radioactivity per volume (curies/cc) regardless of containment 
internal pressure. Since radioactivity is assumed to be homogenized 
in the containment free air volume, the volume percent leaked per 
day is equivalent to the fraction of radioactivity which leaks from 
the containment per day. Therefore, the increase in the peak 
calculated containment internal pressure for the design basis LOCA 
from 49.5 psig to 52 psig does not effect dose consequences 
associated with the design basis LOCA. The proposed change to the 
peak calculated containment internal pressure for the design basis 
LOCA does not impact the radiological consequences of a LOCA as 
analyzed in Chapters 6 and 15 of the UFSAR.
    The proposed amendments do not, therefore involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The possibility of a new or different kind of accident has not been 
created. The increase in the peak calculated containment internal 
pressure for the design basis LOCA does not affect the design or 
operation of existing plant equipment, nor involve new plant equipment. 
Therefore, the proposed change does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The containment design pressure is 60 psig. The acceptance 
criteria in NRC Standard Review Plan, Section 6.2.1.1.A, ``PWR Dry 
Containments, including Subatmospheric Containments,'' requires in 
Item 11.1 that ``the containment design pressure should provide at 
least a 10% margin above the accepted peak calculated containment 
pressure following a loss of coolant accident.'' For PVNGS to 
maintain the required margin, this requires that the peak calculated 
containment internal pressure for the design basis LOCA would be no 
higher than 54 psig. Since the revised peak calculated containment 
internal pressure for the design basis LOCA remains below the 54 
psig limit, the proposed change does not involve a significant 
reduction in the margin of safety.
    Based on this review, it appears that the three standards of 10 
CFR50.92(c) are satisfied. Therefore, the NRC staff proposes to 
determine that the amendments request involve no significant hazards 
consideration. Local Public Document Room location: Phoenix Public 
Library, 1221 N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: 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: William H. Bateman

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 amendment request: April 9, 1997
    Description of amendment request: The proposed change will extend 
the existing Technical Specifications surveillance intervals from 7 
days to 14 days for the Channel Functional Tests for the refueling 
equipment interlocks and for the one-rod-out interlock. The change will 
permit, under most normal circumstances, a complete offloading, 
shuffling, or onloading of fuel, without the need to halt refueling 
activities solely for the performance of these surveillance tests.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change extends the Technical Specification 
Surveillance Requirement (SR) Frequency for the Channel Functional 
Tests (CFTs) for the refueling equipment interlocks and the one-rod-
out interlock. The refueling equipment interlocks and the one-rod-
out interlock are explicitly assumed in the analysis of the control 
rod removal error during refueling. Criticality, and therefore, 
subsequent prompt reactivity excursions are prevented during the 
insertion of fuel, provided all control rods are fully inserted 
during the fuel insertion. The refueling equipment interlocks 
accomplish this by preventing loading fuel into the core with any 
control rod withdrawn, or by preventing withdrawal of a control rod 
from the core during fuel loading. The one-rod-out interlock and 
adequate shutdown margin prevent criticality by preventing 
withdrawal of more than one control rod. With one control rod 
withdrawn, the core will remain subcritical, thereby preventing any 
prompt critical excursion. The proposed change does not change the 
function of any of these interlocks, only the frequency at which the 
interlocks undergo channel functional testing. A review of past test 
performances has demonstrated that extending the Frequency from 7 
days to 14 days will not result in any increase in test failures. 
Therefore, the proposed change will not change the ability of these 
interlocks to perform when required. Based on this, there can be no 
significant increase in the radiological consequences of any 
previously evaluated accident since all interlocks will continue to 
perform as presently analyzed. Therefore, the proposed change does 
not involve a significant increase in the

[[Page 27795]]

probability or consequences of an accident previously evaluated.
    2. The proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change extends the SR Frequency for performing CFTs 
for refueling equipment and one-rod-out interlocks. This change does 
not result in a modification to the plant or to the manner in which 
the plant is operated. The testing will still demonstrated the 
operability of the interlocks. Thus, the interlocks will still 
function in the same manner. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. The proposed change will not involve a significant reduction 
in the margin of safety.
    The proposed change extends the SR Frequency for performing CFTs 
on the refueling equipment and one-rod-out interlocks from 7 days to 
14 days. Reviews of past test results indicate that extending the 
test interval to 14 days will not result in an increase in the 
number of CFT failures for these interlocks. This implies that 
extending the SR Frequency to 14 days will not result in an increase 
in the amount of time the instrument channels will be inoperable 
when required to be operable. Since the proposed change does not 
result in any reduction in the amount of time the instrument 
channels will be operable, the proposed change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile 
Island Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: April 21, 1997
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications that would (1) reduce the volume of 
borated water in the core flood tank (CFT) from 1040 cubic feet to 940 
cubic feet, (2) reduce the surveillance acceptance criteria for the 
emergency core cooling system (ECCS) high pressure injection (HPI) 
flowrate from 500 gallons per minute (GPM) to 431 GPM, and (3) revise a 
limiting condition for operation (LCO) which currently allows either 
local or remote manual operability of decay heat valves to delete the 
local manual valve operability option.
    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 (SHC), which is presented below:
    1. State the basis for the determination that the proposed 
activity will not represent a significant increase in the 
probability of occurrence or consequences of an accident.
    This TSCR [Technical Specification change request] revises the 
LCO for RB [reactor building] sump isolation valves, the LCO for the 
core flood tank level, and the surveillance requirement for HPI 
injection flow rate. The Core Flood and HPI systems are not actuated 
until an event occurs. The CFT level used in the new accident 
analysis is that level required to be maintained in the CFT 
throughout operation (i.e., pre-accident). The new CFT level does 
not prevent safe accident mitigation.
    Likewise, the reduced HPI flow cannot cause an event to occur, 
and while such flow results in less injection to the RCS [reactor 
coolant system] when actuated, this is acceptable as demonstrated in 
the LOCA [loss-of-coolant accident] analyses. Changes to the LCO for 
the RB sump isolation valves support the safety analysis 
assumptions. The action statements related to both the level 
requirement and flow rates remain unchanged by this request. The 
function, operation and surveillance intervals for the isolation 
valves (DH-V-6A/B), the CFT level and HPI injection system are not 
changed by this request. Therefore, this activity does not increase 
the probability of occurrence of an accident, previously evaluated 
in the SAR [safety analysis report].
    Reducing the CFT nominal volume and reducing the HPI flow 
acceptance criteria in the Technical Specifications will not 
increase the radiological consequences of any LOCA evaluated in the 
SAR. The results of analyses using the reduced CFT inventory and 
reduced HPI flow demonstrate that the consequences are within the 
limits of 10 CFR 50.46. No fuel failure in addition to that assumed 
in the evaluation of the dose consequences would occur. Therefore, 
the radiological consequences would not increase.
    The editorial changes described above have no impact upon the 
probability of occurrence or consequences of an accident.
    2. State the basis for the determination that the activity does 
not create the possibility of an accident of a new or different type 
than any previously analyzed in the SAR.
    This TSCR revises the LCO for RB sump isolation valves, the LCO 
for the core flood tank level, and the surveillance requirement for 
HPI injection flow rate. This change will not adversely affect the 
capability of the emergency core cooling systems in the event of a 
LOCA. The function, operation and surveillance intervals for both 
the borated water level in the core flood tank, and ECCS systems are 
not changed by this request and no physical changes or modifications 
are being made to Core Flood and HPI system boundaries. Therefore, 
because there are no configuration changes this activity does not 
create the possibility of an accident or malfunction of a different 
type than previously analyzed in the SAR.
    In addition, the editorial changes described above do not create 
the possibility of an accident of a new or different type than any 
previously analyzed in the SAR.
    3. State the basis for the determination that the margin of 
safety is not significantly reduced.
    This TSCR revises the LCO for RB sump isolation valves, the LCO 
for the core flood tank level, and the surveillance requirement for 
HPI injection flow rate. No system configuration changes (hardware 
modifications) will be made to implement the change request, upon 
approval of the license amendment. The action requirements for these 
technical specifications have not changed. Actions to be taken if 
operability requirements are not met include plant shutdown under 
certain conditions.
    Furthermore, impact upon the margin to safety is limited because 
the results of the LOCA analyses demonstrate that the 10 CFR 50.46 
acceptance criteria are met, specifically: the PCT [peak clad 
temperature] limit and the core-wide oxidation limit of 1 percent of 
the fuel cladding, as identified in the Technical Specification 
bases. Hence the margin of safety as defined in the bases of any 
technical specification is not significantly reduced or impacted by 
the implementation of this change request, or the editorial changes 
described above.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Patrick D. Milano, Acting

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: April 22, 1997
    Description of amendment request: The proposed amendments would 
revise Technical Specifications 5.3.1, Fuel Assemblies, and 6.9.1.6, 
Core Operating Limits Report, to allow use of

[[Page 27796]]

an alternate zirconium-based fuel cladding, ZIRLO, and limited 
substitution of fuel rods by ZIRLO filler rods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The methodologies used in the accident analyses remain 
unchanged. With the exception of a reduction in the heat flux hot 
channel factor (F<INF>Q</INF>), the operating limits will not be 
changed. The proposed changes will not result in any equipment 
exceeding its design limits under normal or accident conditions. The 
calculated doses presented in the UFSAR will remain bounding. Other 
than the changes to the fuel assemblies, there are no physical 
changes to the plant associated with this Technical Specification 
change. A reload safety analysis will continue to be performed for 
each cycle to demonstrate compliance with fuel safety design bases.
    VANTAGE+ fuel assemblies with ZIRLO clad fuel rods meet the same 
fuel assembly and fuel rod design bases as VANTAGE 5H fuel 
assemblies. Since the original design criteria are met, the ZIRLO 
clad fuel rods will not be an initiator for any new accident. The 
clad material is similar in chemical composition and has similar 
physical and mechanical properties to Zircaloy. Thus, cladding 
integrity is maintained and the structural integrity of the fuel 
assembly is not affected. ZIRLO cladding improves corrosion 
performance and dimensional stability. No concerns have been 
identified with respect to the mixed core of Zircaloy and ZIRLO clad 
assemblies. Also, no concerns have been identified with respect to 
the use of an individual assembly containing a combination of 
Zircaloy and ZIRLO clad fuel rods.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not result in any equipment exceeding 
its design limits under normal or accident conditions. All design 
and performance criteria continue to be met and no new failure 
mechanisms have been identified. The ZIRLO cladding material offers 
improved corrosion resistance and structural integrity.
    The proposed changes do not affect the operation of any system 
or component in the plant. The safety functions of the related 
structures, systems, or components are not changed, nor is the 
reliability of any structure, system, or component reduced. The 
changes do not affect the manner by which the facility is operated 
and do not change any facility design feature, structure, or system. 
No new or different type of equipment will be installed. Since there 
is no other change to the facility or operating procedures, and the 
safety functions and reliability of structures, systems, or 
components are not affected, the proposed changes do not create the 
possibility of a new accident or an accident different from those 
previously evaluated.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Use of ZIRLO fuel cladding material will not result in any 
equipment exceeding its design or licensing bases limits under 
normal or accident conditions. VANTAGE 5H reload design and safety 
analysis limits are unchanged. For each cycle reload core, the fuel 
assemblies will be evaluated using NRC-approved reload design 
methods, including consideration of the core physics analysis 
peaking factors and core average linear heat rate effects. ZIRLO 
fuel assemblies will be assessed for use under conditions consistent 
with normal core operating conditions allowed in the Technical 
Specifications. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment request: March 26, 1997
    Description of amendment request: The proposed amendment would 
modify the technical specifications (TSs) which describe the control 
room ventilation system autostart functions.
    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:
    Per 10 CFR 50.92, the proposed changes do not involve a 
significant hazards consideration if the proposed changes do not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated;
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated; or
    3. involve a significant reduction in a margin of safety.
    Criterion 1
    These changes are administrative in nature, intended to correct 
and clarify the TS description of control room ventilation system 
operation. Because no changes to plant operations or physical 
changes to the plant will occur due to these changes, they do not 
involve a significant increase in the probability or consequences of 
a previously evaluated accident.
    Criterion 2
    Because no changes to plant operations or the physical plant 
will occur due to these changes, the changes will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    Criterion 3
    These changes are administrative in nature, intended to correct 
and clarify the present TSs with regard to system operation 
descriptions. Thus, the changes involve no reduction in margins 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:  Maud Preston Palenske 
Memorial Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment request: March 26, 1997
    Description of amendment request: The proposed amendment would make 
three administrative changes to the technical specifications (TSs) 
dealing with a grammatical error, an inadvertently deleted frequency 
requirement, and a footnote which is no longer applicable.
    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:
    Per 10 CFR 50.92, the proposed changes do not involve a 
significant hazards consideration if the proposed changes do not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated;

[[Page 27797]]

    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.
    Criterion 1
    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed changes to the TSs do not affect the 
assumptions, parameters, or results of any UFSAR accident analysis. 
The firstproposed change, ``A'', is a grammatical correction; the 
second proposed change, ``B'', reformats the page, and returns a 
frequency requirement that, while inadvertently deleted from the 
TSs, was still met via procedure; the third proposed change deletes 
a footnote which is no longer applicable. As described in Section 
II.C. of licensee's application request dated March 26, 1997, a load 
drop analysis is not required for single-failure-proof load blocks.
    Criterion 2
    The proposed changes do not involve physical changes to the 
plant or changes in plant operating configuration. The changes 
described above are essentially administrative in nature, and thus 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Criterion 3
    The proposed changes are essentially administrative in nature. 
Per NUREG-0612, single-failure-proof cranes are exempt from the 
requirements of a load drop analysis; therefore, 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: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Gail H. Marcus

North Atlantic Energy Service Corporation, Docket No. 50-443, 
Seabrook Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: February 12, 1997
    Description of amendment request: The proposed amendment would 
change position titles in certain Seabrook Station, Unit No. 1 
(Seabrook) Appendix A Technical Specifications (TS) to reflect the 
present Seabrook organization, would clarify the approval authority for 
the Station Qualified Reviewer Program, and would correct a reference. 
Specifically, the proposed amendment would:
    1. Change TS 6.0, ``Administrative Controls'' to reflect accurately 
the current North Atlantic Management organization, their assigned 
duties as previously reported to the NRC, and their proper titles,
    2. Corrects an incorrect reference in TS 6.4.3.9.b., and
    3. Clarifies the term ``Manager'' in TS 6.4.2, ``Station Qualified 
Reviewer Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The changes do not involve a significant increase in the 
probabilityor consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)) because the proposed changes are merely 
administrative or editorial in nature. The proposed changes involve 
position title changes to reflect current organization, correct an 
incorrect reference, and provide clarification with regard to the 
organizational level for certain approvals. The changes do not 
affect the manner by which the facility is operated and do not 
change any facility design feature or equipment. Since there is no 
change to the facility or operating procedures, there is no effect 
upon the probability or consequences of any accident previously 
analyzed.
    B. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
(10 CFR 50.92(c)(2)) because they do not affect the manner by which 
the facility is operated or involve any changes to equipment or 
features which affect the operational characteristics of the 
facility. Therefore, no new accident initiator is introduced that 
could cause a new or different kind of accident from those 
previously evaluated. The proposed changes merely involve position 
title changes to reflect current organization, correct an incorrect 
reference, and provide clarification with regard to the 
organizational level for certain approvals.
    C. The changes do not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes 
do not affect the manner by which the facility is operated or 
involve equipment or features which affect the operational 
characteristics of the facility.Based on this review, it appears 
that the three standards of 10 CFR 50.92(c) are satisfied. 
Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esquire, Northeast 
Utilities Service Company, Post Office Box 270, Hartford CT 06141-0270.
    NRC Project Director: Patrick D. Milano

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London 
County, Connecticut

    Date of amendment request: April 15, 1997
    Description of amendment request: The proposed amendment would make 
changes to Technical Specification Sections 4.3.3.6 and 4.6.4.1, which 
require that the hydrogen monitors be periodically tested. 
Specifically, the changes to the surveillances would increase the 
testing of the monitor's hydrogen sensor, correct inconsistencies 
between surveillances, and make changes to the Bases of the 
surveillances.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO [Northeast Nuclear Energy Company] has reviewed the 
proposed changes in accordance with 10CFR 50.92 and has concluded 
that the change does not involve a significant hazards consideration 
(SHC). The bases for this conclusion is that the three criteria of 
10CFR 50.92(c) are not satisfied. The proposed changes do not 
involve [an] SHC because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test 
once per 92 days on a staggered test basis is consistent with the 
design and operation of the hydrogen monitor system. The hydrogen 
monitoring system is independent of the reactor coolant system 
boundary, has no effect on the probability of occurrence of a loss 
of coolant accident and performing surveillance testing does not 
significantly increase the probability of an accident previously 
evaluated.
    The proposed change to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test 
will not require the opening of a containment isolation valve and 
conducting surveillance testing does not significantly increase the 
consequence of an accident previously evaluated.
    The proposed change to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to change the channel check frequency from once 
per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an 
analog channel operational test to surveillance 4.3.3.6.2 and make 
editorial changes to the surveillances and bases sections are 
considered administrative changes. Administrative changes do not 
involve a significant increase in the

[[Page 27798]]

probability or consequence of an accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test do 
not add any new equipment to the plant and do not affect the way any 
system important to safety is operated either in normal or under 
accident conditions.
    The proposed changes to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to change the channel check frequency from once 
per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an 
analog channel operational test to surveillance 4.3.3.6.2 and make 
editorial changes to the surveillances and bases sections are 
considered administrative changes.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to perform a hydrogen sensor calibration test 
will provide assurance of expected instrument performance under 
accident conditions and performing surveillance testing do not 
involve a significant reduction in a margin of safety.
    The proposed changes to Technical Specification Surveillances 
4.3.3.6 and 4.6.4.1 to change the channel check frequency from once 
per 31 days to once per 12 hours on Table 4.3-7 Item 18, add an 
analog channel operational test to surveillance 4.3.3.6.2 and make 
editorial changes to the surveillances and bases sections are 
considered administrative changes. Administrative changes do not 
involve a significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed changes do not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270NRC Deputy Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), et al., Docket No. 50-
423, Millstone Nuclear Power Station, Unit No. 3, New London 
County, Connecticut

    Date of amendment request: April 17, 1997
    Description of amendment request: The proposed amendment would 
modify Technical Specification 3.7.14 by clarifying the actions to be 
taken when an area temperature exceeds its temperature limit.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO [Northeast Nuclear Energy Company] has reviewed the 
proposed change in accordance with 10CFR 50.92 and has concluded 
that the change does not involve a significant hazards consideration 
(SHC). The bases for this conclusion is that the three criteria of 
10CFR 50.92(c) are not satisfied. The proposed change does not 
involve [an] SHC because the change would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    The proposed change to Technical Specification 3.7.14 will 
establish allowable tolerances to ensure that the applicable 
systems, structures and components are operated within their 
existing design bases.
    Technical Specification 3.7.14 specifies the actions to be taken 
when an area temperature exceeds its temperature limit. The action 
taken is dependent on the amount and duration by which the area 
temperature exceeds its limit. Actions are currently specified for 
exceeding area temperature by less than 20  deg.F and greater than 
20 deg.F for periods less than 8 hours and for periods greater than 
8 hours. This change clarifies the actions to be taken when the 
temperature exceeds its limit by exactly 20  deg.F or exceed its 
limit for exactly 8 hours. It is concluded that this change is a 
clarification only in that it causes the more conservative actions 
to be taken at greater than or equal to 20  deg.F, or at greater 
than or equal to 8 hours.
    The proposed change, therefore, does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Establishment of tolerances and clarification of actions at a 
specific value does not [ ] change the operation of any system, 
structure or component during normal or accident conditions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The change is administrative in nature in that it resolves a 
discontinuity in the range of temperatures and in the duration 
period above the applicable limit for which action is required. 
Establishment of tolerances ensures parameters are set and 
maintained within allowable design constraints. Clarification of 
applicability for the required actions ensures that action is 
proscribed for all possible conditions thereby not permitting 
operation outside of allowable design.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed change does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: March 31, 1997
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) Sections 3/4.6.5.3.2, ``Filtration, 
Recirculation, and Ventilation System (FRVS),'' to (1) provide an 
appropriate Limiting Condition for Operation and ACTION Statement that 
reflects the design basis for the FRVS, and (2) clarify the manner in 
which FRVS testing is performed.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or

[[Page 27799]]

consequences of an accident previously evaluated.
    The proposed TS revisions involve: 1) no hardware changes; 2) no 
significant changes to the operation of any systems or components in 
normal or accident operating conditions; and 3) no changes to 
existing structures, systems or components. Therefore these changes 
will not increase the probability of an accident previously 
evaluated. Since the plant systems associated with these proposed 
changes will still be capable of: 1) meeting all applicable design 
basis requirements; and 2) retaining the capability to mitigate the 
consequences of accidents described in the HC [Hope Creek] UFSAR 
[Updated Final Safety Analysis Report], the proposed changes were 
determined to be justified. As a result, these changes will not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes contained in this submittal will not 
adversely impact the operation of any safety related component or 
equipment. Since the proposed changes involve: 1) no hardware 
changes; 2) no significant changes to the operation of any systems 
or components; and 3) no changes to existing structures, systems or 
components, there can be no impact on the potential occurrence of 
any accident. Furthermore, there is no change in plant testing 
proposed in this change request which could initiate an event. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes for the TS related to the Filtration 
Recirculation and Ventilation System (FRVS) Recirculation Subsystem 
provide consistency between the Hope Creek TS and post-accident 
descriptions of the FRVS Recirculation Subsystem operation already 
contained in the UFSAR and reflected in the Hope Creek SER [Safety 
Evaluation Report] (NUREG-1048). PSE&G [Public Service Electric & 
Gas] believes that the proposed allowed outage times and ACTION 
Statements for the FRVS Recirculation Subsystem: 1) will ensure that 
the required minimum number of FRVS recirculation units will be 
available to mitigate the consequences of accidents described in the 
UFSAR; and 2) provide appropriate direction and time requirements 
for placing the unit in a safe shutdown condition when the system is 
degraded. Therefore, the changes contained in this request do not 
result in a significant reduction in a margin of safety.
    The revisions to Surveillance Requirement 4.6.5.3.2.b provide an 
accurate and clearly defined basis for performing this surveillance 
test. The proposed changes implement PSE&Gs existing interpretation 
of the TS requirements and therefore do not alter the manner in 
which this surveillance test is currently being performed. PSE&G has 
concluded that this surveillance test method appropriately tests the 
FRVS Recirculation Subsystem. Since the FRVS recirculation units 
will continue to be tested with the heaters: 1) operable; and 2) set 
at the demand necessary to ``reduce the buildup of moisture,'' PSE&G 
believes that the proposed changes to clarify the TS are justified. 
Therefore, the changes contained in this request do not result in a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: March 31, 1997
    Description of amendment request: The proposed amendment would 
provide changes to Technical Specification (TS) 2.1.2, ``THERMAL POWER, 
High Pressure and High Flow,'' ACTION a.1.c for TS 3.4.1.1, 
``Recirculation Loops,'' and the Bases for TS 2.1, ``Safety Limits.'' 
These changes are being made to implement an appropriately conservative 
Safety Limit Minimum Critical Power Ratio (SLMCPR) for all Hope Creek 
core and fuel designs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The derivation of the revised SLMCPRs for Hope Creek for 
incorporation into the Technical Specifications, and its use to 
determine cyclespecific thermal limits, have been performed using 
NRC approved methods. Additionally, interim implementing procedures 
which incorporate cyclespecific parameters have been used which 
result in a more restrictive value for SLMCPR. These calculations do 
not change the method of operating the plant and have no effect on 
the probability of an accident initiating event or transient.
    There are no significant increases in the consequences of an 
accident previously evaluated. The basis of the MCPR [Minimum 
Critical Power Ratio] Safety Limit is to ensure that no mechanistic 
fuel damage is calculated to occur if the limit is not violated. The 
new SLMCPRs preserve the existing margin to transition boiling and 
the probability of fuel damage is not increased. Therefore, the 
proposed change does not involve an increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes contained in this submittal result from an 
analysis of the Cycle 7 core reload using the same fuel types as 
previous cycles. These changes do not involve any new method for 
operating the facility and do not involve any facility 
modifications. No new initiating events or transients result from 
these changes. Therefore, the proposed Technical Specification 
changes do not create the possibility of a new or different kind of 
accident, from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety as defined in the Technical Specification 
bases will remain the same. The new SLMCPRs are calculated using NRC 
approved methods which are in accordance with the current fuel 
design and licensing criteria. Additionally, interim implementing 
procedures, which incorporate cyclespecific parameters, have been 
used. The MCPR Safety Limit remains high enough to ensure that 
greater than 99.9% of all fuel rods in the core will avoid 
transition boiling if the limit is not violated, thereby preserving 
the fuel cladding integrity. Therefore, the proposed Technical 
Specification 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey

    Date of amendment request: April 11, 1997
    Description of amendment request: The proposed amendments would 
change Technical Specification 3.6.2.3, ``Containment Cooling System'' 
and the

[[Page 27800]]

associated bases. The changes would increase the cooling water flow 
rate for the 31-day and 18-month surveillances and specify that during 
the 31-day surveillance the fans are started and operated in low speed. 
The changes are being proposed to ensure that the cooling water flow 
rate and the fan speed being verified are representative of the 
Containment Fan Cooling Unit post-accident mode of 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:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes ensure that the fan speed and cooling water 
flow rate being verified is representative of the fan speed and 
cooling water flow rate required for the post-accident mode of 
operation. The proposed changes affect an accident mitigation system 
and are being made to assure that the system is being tested in its 
accident mitigation mode. There are no new accident initiators 
created by the proposed changes. Therefore, the proposed changes do 
not involve a significant increase in the probability of an accident 
previously evaluated.
    The proposed changes provide assurance that the CFCUs will be 
capable of maintaining peak containment pressure and temperature 
within design limits by verifying the proper post-accident cooling 
water flow to the CFCUs. No physical changes to the plant result 
from the proposed changes to the surveillance requirements. 
Therefore, the proposed changes do not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes for demonstrating operability of the CFCUs 
in the low speed mode, with the required post-accident cooling water 
flow rate, are consistent with the existing safety function of the 
CFCUs following a Design Basis Accident (DBA). The proposed changes 
to the surveillance requirements do not involve any physical changes 
to plant components, systems or structures, or the operation of the 
CFCUs in the post-accident mode. Therefore, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes to the surveillance requirements provide 
assurance that the CFCUs will perform their intended design function 
of maintaining peak containment pressure and temperature consistent 
with the current design basis following a DBA by verifying the 
proper post-accident cooling water flow to the CFCUs. Since the high 
speed and low speed control circuits are independent and there are 
separate breakers used to energize the CFCU motors in high and low 
speed, the CFCUs would be capable of starting in the low speed mode 
following a DBA although the high speed breaker and control circuit 
may not be available.
    Verification of the post-accident flow rate during the 31 day 
surveillance also ensures that the required supporting system, 
Service Water, is available for normal operation. To ensure that the 
containment air temperature is maintained below the initial 
temperature condition assumed in the accident analysis during normal 
operation, Technical Specification 3/4.6.1.5 requires verification 
of the average containment temperature once every 24 hours in Modes 
1 through 4.
    The proposed changes to the CFCU surveillance requirements do 
not affect the ability of the CFCUs to perform their normal and 
post-accident functions. These proposed changes ensure the 
verification of the proper post-accident service water flow rate to 
the CFCUs. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

South Carolina Electric & Gas Company (SCE&G), South Carolina 
Public Service Authority, Docket No. 50-395, Virgil C. Summer 
Nuclear Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: March 26, 1997
    Description of amendment request: The proposed amendment would 
revise the Virgil C. Summer Nuclear Station Technical Specifications to 
change the definition of ``Core Alteration.'' The proposed definition 
will not consider movement of components other than fuel, sources, or 
reactivity control components. These proposed changes are technically 
consistent with the requirements of NUREG-1431, Revision 1, 
``Westinghouse Standard Technical Specifications,'' issued on April 7, 
1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes revise the definition of Core Alteration to be 
the movement of fuel, sources, or reactivity control components; and to 
delete ``or manipulation'' and ``conservative'' from the text. These 
changes do not affect the probability of an accident previously 
evaluated. The movement of components other than fuel, sources, and 
reactivity control components, within the reactor vessel is enveloped 
by the analyzed event. Deleting the words ``or manipulation'' and 
``conservative'' from the definition of Core Alteration are 
administrative changes and also do not impact initiators of analyzed 
events. The only component assumed to be an initiator of an analyzed 
event is dropping an irradiated fuel assembly, however, fuel is still 
part of the definition. Furthermore, a fuel handling accident is 
minimized by administrative controls and physical limitations imposed 
on fuel handling operations. The movement of components other than 
fuel, sources, and reactivity control components within the reactor 
vessel will be controlled under plant administrative controls. This 
change has no effect on the boron dilution event because when boron 
concentration is below limits, Core Alterations are restricted to 
maintain the maximum Shutdown Margin. Movement of other components will 
have a negligible impact on core reactivity.
    The changes to the definition of Core Alteration do not increase 
the consequences of an accident previously evaluated. The accident 
analysis assumes an irradiated fuel assembly is dropped with the 
consequences well within the 10 CFR 100 limits. The dropping of 
other components was not addressed in the plant safety analyses, 
however, the analysis of the dropped fuel assembly encompasses other 
components. The consequences of a boron dilution event are not 
addressed because Core Alterations are not allowed when the boron 
concentration is below limits. These changes do not affect the 
mitigation capabilities of any component or system nor do they 
affect the assumptions relative to the mitigation of accidents or 
transients. Therefore, the change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?

    The proposed changes revise the definition of Core Alteration to be 
the movement of fuel, sources, or reactivity control components; and to 
delete ``or manipulation'' and ``conservative'' from the text. The 
change does not involve a

[[Page 27801]]

significant change in the design or operation of the plant. The changes 
do not involve a physical alteration of the plant (no new or different 
type of equipment will be installed), or new or unusual operator 
actions. The changes will not impose any new or different requirements 
or eliminate any existing requirements. The definition of Core 
Alteration is being clarified and made consistent with NUREG-1431, Rev. 
1. Therefore, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in margin of 
safety?
    The proposed changes revise the definition of Core Alteration to 
be the movement of fuel, sources, or reactivity control components; 
and to delete ``or manipulation'' and ``conservative'' from the 
text. The safety analysis assumes an irradiated fuel assembly is 
dropped. Controls for handling components other than fuel, sources, 
or reactivity control components within the reactor vessel are in 
plant administrative controls. The effect of a boron dilution event 
on Shutdown Margin is limited due to the requirement to suspend Core 
Alterations. The movement of other components have a negligible 
impact on core reactivity. No change is being proposed, in the 
applicability of the definition, to the movement of components which 
factor in the design basis analyses (fuel handling accident). 
Deleting the terms ``or manipulation'' and ``conservative'' from the 
definition of Core Alteration results in a clarification to the 
definition that does not technically alter the meaning. Therefore, 
the change does not involve a significant reduction in a margin of 
safety
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: F. Mark Reinhart, Acting

South Carolina Electric & Gas Company (SCE&G), South Carolina 
Public Service Authority, Docket No. 50-395, Virgil C. Summer 
Nuclear Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: March 26, 1997
    Description of amendment request: The proposed amendment would 
revise the Virgil C. Summer Nuclear Station Technical Specifications 
(TS), Surveillance Requirement (SR) 4.5.2.a, to add (1) the charging/
high head safety injection (HHSI) pump cross connect valves, and (2) 
the charging pump mini-flow header isolation valve, to the SR valve 
list. The proposed change is an administrative change to meet the 
recommendations of NRC Branch Technical Position (BTP) EICSB 18, which 
establishes the acceptability of disconnecting power to electrical 
components of fluid systems as one means of designing against a single 
failure that might cause an undesirable component action. TS SR 4.5.2.a 
includes a list of the required positions of manually-controlled, 
electrically-operated valves, and identify those valves to which the 
requirements for removal of electrical power is applied in order to 
satisfy the single failure criterion.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change adds the charging/HHSI pump cross connect 
valves and the charging pump mini-flow header isolation valve to the 
ECCS [Emergency Core Cooling System] Subsystems - T<INF>avg</INF> 
(greater than or equal to) 350 deg.F Technical Specification 
Surveillance Requirement. This Surveillance Requirement will require 
the valves to be verified open with power to the valve operators 
removed once per 12 hours. ... The charging/HHSI pump cross connect 
valves and the charging mini-flow header isolation valve are not 
initiators of any analyzed event. ... The charging pump/HHSI pump 
cross connect valves are being modified to meet the recommendations 
of the BTP (including this Technical Specification change). The 
charging pump mini-flow header isolation valve meets the 
requirements of the BTP except it is not located in the Technical 
Specifications. ... Requiring the valves to be verified open with 
power removed from the valve operator once per 12 hours does not 
affect the assumptions relative to the mitigation of accidents or 
transients. This requirement ensures that the valves are in a 
position with power removed so that a failure will not occur that 
will affect the mitigation of an accident. These valves are required 
to be open during a LOCA [loss-of-coolant accident]. This change 
will ensure that the valves are open with power removed. Therefore, 
the change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Does this change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    ...This change does not involve a significant change in the 
design or operation of the plant. This change is a result of BTP 
EICSB 18. The charging/HHSI pump cross connect valves are being 
modified to have power lockout capability, redundant indication on 
the main control board, and be included in the Technical 
Specifications. This will ensure that a single failure (hot short in 
the controls of either valve) will not cause spurious actuation of 
the valves during the injection or recirculation phase of the ECCS. 
The charging pump mini-flow header isolation valve meets the 
requirements of the BTP except it is not located in the Technical 
Specifications. The charging/HHSI pump cross connect valves and 
charging pump mini-flow header isolation valve are required to 
remain open during a LOCA. This modification will ensure that the 
valves will remain open during an accident which requires ECCS 
operation. The proposed change will not introduce any new accident 
initiators. Therefore, the change does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in margin of 
safety?
    ...The ECCS is required to operate upon receipt of a safety 
injection signal. The charging/HHSI pump cross connect valves and 
the charging pump mini-flow header isolation valve are required to 
remain open during ECCS operation. However, a single failure may 
cause a spurious actuation (closure) of the valves which could 
hinder HHSI flow. The modification to the charging/HHSI cross 
connect valves (the addition of a power lockout feature and 
redundant position indication) and the added TS Surveillance 
Requirement will eliminate this failure scenario and ensure the 
valves remain in their safety function position (open). The charging 
pump mini-flow header isolation valves already contain a power 
lockout feature and redundant position indication. These valves are 
being added to the Technical Specifications to meet the requirements 
of BTP EICSB 18. Therefore, the change does not involve a 
significant reduction in a margin of safety[.]
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: Mark Reinhart, Acting

[[Page 27802]]

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

    Date of application for amendments: March 13, 1997 (TS 97-01)
    Brief description of amendments: The amendments change the 
Technical Specifications by raising the allowable U-235 enrichment, as 
specified in Section 5.6.1.2, of fuel stored in the new fuel pit 
storage racks from 4.5 to 5.0 weight percent.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Operation of 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 allowed enrichment of new fuel stored 
in the new fuel storage racks does not change the criticality 
potential with the proposed fuel arrangement requirements for the 
storage racks. The potential k<INF>eff</INF> values are maintained 
the same as the current TS [Technical Specification] requirements. 
In addition, the storage racks are not modified, other than the 
locations that cannot be filled with fuel assemblies, and the 
processes for loading and unloading fuel in these racks and the 
controls for these racks remain the same. Since the k<INF>eff</INF> 
limits and operating processes are unchanged by the proposed 
revision, there is no increase in the probability of an accident 
previously evaluated. Likewise, there is no impact to the 
consequences of an accident or increase in offsite dose limits as a 
result of the proposed TS change because the criticality 
requirements are unchanged and plant equipment will be utilized and 
operated without change considering the fuel storage location limits 
imposed by this request.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    As stated above, the plant equipment and operating processes 
will not be altered by the proposed TS change with the exception of 
allowed fuel storage locations in the new fuel storage racks. The 
limitations on acceptable fuel storage locations in the racks ensure 
that the k<INF>eff</INF> limits are maintained at the same limits as 
currently required. TVA has not postulated a criticality event at 
SQN for the spent or new fuel storage locations because the design 
of the associated storage racks, potential moderation, and TS 
allowable fuel enrichments do not support the potential for this 
condition. Considering the physical barriers that will be installed 
and verified to be in place prior to initial loading of fuel in the 
new fuel storage racks, the new fuel storage rack physical 
limitations will continue to ensure that criticality events are not 
credible for the proposed change. Therefore, this change does not 
create the potential for a new accident from any previously 
analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed TS change maintains the existing requirements for 
criticality by utilizing limited storage locations in the new fuel 
pit storage racks. There is no change to operating practices 
associated with the use and control of these racks except for the 
storage limitations. For these reasons, there will be no reduction 
in the margin of safety as a result of implementing the proposed TS 
change.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of amendment request: April 4, 1997 (TSCR 197)
    Description of amendment request: The proposed amendments revise TS 
15.6, ``Administrative Controls,'' and 15.7, ``Radiological Effluent 
Technical Specifications,'' to change the corporate officer responsible 
for nuclear operations from ``Vice President-Nuclear Power,'' to 
``Chief Nuclear Officer,'' and to require that the position be an 
officer of the company.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes are administrative only. There are no 
physical changes to the facility or its operation. All Limiting 
Conditions of Operation, Limiting Safety System Settings, and Safety 
Limits specified in the Technical Specification remain unchanged. 
Additionally, there are no changes in the Quality Assurance Program, 
Emergency Plan, Security Plan, and Operator Training and 
Requalification Program. Therefore, an increase in the probability 
or consequences of an accident previously evaluated cannot occur.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes are administrative only. No changes to the 
facility structures, systems and components or their operation will 
result. The design and design basis of the facility remain 
unchanged. The plant safety analyses remain current and accurate. No 
new or different failure mechanisms are introduced. Therefore, the 
possibility of a new or different kind of accident from any accident 
previously evaluated is not introduced.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.
    The proposed amendments are administrative only. All safety 
margins established through the design and facility license 
including the Technical Specifications remain unchanged. In 
addition, the proposed amendments ensure continued emphasis and 
assignment of responsibility for overall nuclear safety. Therefore, 
all margins of safety are maintained.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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

Wisconsin Electric Power Company, Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Power Plant, Unit Nos. 1 and 2, Town of Two 
Creeks, Manitowoc County, Wisconsin

    Date of amendment request: April 14, 1997 (TSCR 198)
    Description of amendment request: The proposed amendments revise TS 
15.3.1, ``Reactor Coolant System,'' to require both reactor coolant 
pumps to be operable when the reactor is critical and to require that 
the reactor be placed in hot shutdown within 6 hours if one or both 
reactor coolant pumps cease operating. This revision eliminates the 
current provision which allows single pump operation up to 3.5 percent 
power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 27803]]

licensee has provided its analysis of the issue of no significant 
hazards consideration which is presented below:
    1. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not result in a significant increase in 
the probability or consequences of an accident previously evaluated.
    The amendments proposed eliminate an inconsistency in the 
Technical Specifications in a conservative manner. The proposed 
changes ensure that required protection functions remain operable in 
all required modes of operation. Since the protection functions 
remain operable in accordance with existing Technical Specification 
requirements and serve to mitigate analyzed events no increase in 
the consequences of a previously analyzed accident results. The 
protective functions are not accident initiators and are maintained 
and tested in accordance with existing Technical Specification 
requirements, therefore the probability of a previously analyzed 
accident cannot increase. Therefore, operation of the Point Beach 
Nuclear Plant in accordance with the proposed changes does not 
result in an increase in probability or consequences of a previously 
analyzed accident.
    2. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed amendments restore consistency within the Technical 
Specifications thus ensuring the protections functions remain 
operable as required and the units are operated within the bounds of 
the existing safety analyses. Therefore, operation of the Point 
Beach Nuclear Plant in accordance with the proposed amendments does 
not result in a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the Point Beach Nuclear Plant in accordance with 
the proposed amendments does not involve a significant reduction in 
a margin of safety.
    Margins of safety are defined by the bounds of the design and in 
the safety analyses performed for the Point Beach Nuclear Plant. The 
proposed amendments eliminate an inconsistency within the Technical 
Specifications and ensure the plant will respond as analyzed in the 
Safety Analyses. There is no physical change in the facility or 
operation. Therefore, operation of the Point Beach Nuclear Plant in 
accordance with the proposed amendments does not involve a reduction 
in 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: 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

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

    Date of amendment request: March 21, 1997, as supplemented by 
letter dated April 15, 1997.
    Description of amendment request: This amendment request proposes 
to revise the technical specifications associated with the inspection 
of the reactor coolant flywheel to provide an exception to the 
recommendations of Regulatory Guide 1.14, Revision 1, ``Reactor Coolant 
Pump Flywheel Integrity.'' The proposed exception would allow either an 
ultrasonic volumetric examination or surface examination to be 
performed at approximately 10-year intervals. In addition, a correction 
of the issuance date of a referenced regulatory guide is included.
    This amendment would also allow delaying the complete flywheel 
examination for the ``D'' reactor coolant pump until the Fall 1997 
outage.
    This supersedes the staff's proposed no significant hazards 
consideration determination evaluation for the requested changes that 
was published on January 2, 1997 (62 FR 133).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The safety function of the RCP [reactor coolant pump] flywheels 
is to provide a coastdown period during which the RCPs would 
continue to provide reactor coolant flow to the reactor after loss 
of power to the RCPs. The maximum loading on the RCP flywheel 
results from overspeed following a LOCA [loss-of-coolant accident]. 
The maximum obtainable speed in the event of a LOCA was predicted to 
be less than 1500 rpm. Therefore, a peak LOCA speed of 1500 rpm is 
used in the evaluation of RCP flywheel integrity in WCAP-14535. This 
integrity evaluation shows a very high flaw tolerance for the 
flywheels. The proposed change does not affect that evaluation. 
Reduced coastdown times due to a single failed flywheel is bounded 
by the locked rotor analysis, therefore, it would not place the 
plant in an unanalyzed condition. Therefore, these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any previously evaluated since 
the proposed amendments will not change the physical plant or the 
modes of plant operation defined in the facility operating license. 
No new failure mode is introduced due to the proposed change, since 
the proposed change does not involve the addition or modification of 
equipment, nor do they alter the design or operation of affected 
plant systems, structures, or components.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are basically unchanged by the 
proposed amendment. The results of the flywheel inspections 
performed have identified no indications affecting flywheel 
integrity. As identified in WCAP-14535, detailed stress analysis as 
well as risk analysis have been completed with the results 
indicating that there would be no change in the probability of 
failure for RCP flywheels if all inspections were eliminated.
    Therefore these changes do not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

[[Page 27804]]

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.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, located in Grundy County, 
Illinois

    Date of amendment request: January 24, 1997.
    Description of amendment request: The application proposed to 
change the Technical Specifications to reflect the installation of new 
reactor water level instrumentation for the Emergency Core Cooling 
System actuation.
    Date of publication of individual notice in Federal Register: April 
18, 1997 (62 FR 19143).Expiration date of individual notice: May 19, 
1997
    Local Public Document Room location: The Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, located in Grundy County, 
Illinois

    Date of amendment request: March 5, 1997.
    Description of amendment request: The application proposed to 
remove the Main Steam Line Radiation Monitor High scram and the Main 
Steam Line Tunnel Radiation High input to the Main Steam Line Isolation 
function requirement from the Technical Specifications (TS). The 
proposed changes are a result of a Boiling Water Reactor Owners Group 
initiative to minimize inadvertent scrams and Main Steam Isolation 
Valve closure due to erroneous radiation monitor actuation.
    Date of publication of individual notice in Federal Register: April 
18, 1997 (62 FR 19141).Expiration date of individual notice: May 19, 
1997
    Local Public Document Room location: The Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, located in Rock Island 
County, Illinois

    Date of amendment request: April 21, 1997
    Description of amendment request: The amendments would reflect a 
change in the Quad Cities, Unit 2, Minimum Critical Power Ratio (MCPR) 
Safety Limit and add the Siemens Power Corporation (SPC) methodology 
for application of the Advanced Nuclear Fuel for Boiling Water Reactors 
(ANFB) Critical Power Correlation to coresident General Electric fuel 
for Quad Cities, Unit 2, Cycle 15, to Technical Specification Section 
6.9.A.6.b.
    Date of publication of individual notice in Federal Register: April 
30, 1997 (62 FR 23499)
    Expiration date of individual notice: May 30, 1997
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: March 31, 1997
    Brief description of amendment: The proposed amendment would revise 
the Ginna Station Improved Technical Specifications to reflect a 
planned modification to the spent fuel pool storage racks.Date of 
publication of individual notice in Federal Register: April 30, 1997 
(62 FR 23502)
    Expiration date of individual notice: May 30, 1997
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendments: March 5, 1997, as supplemented 
May 9, 1997. The May 9, 1997, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    Brief description of amendments: The amendments incorporate a new 
Technical Specification for instrumentation associated with automatic 
isolation of a pathway for release of non-condensible gases from the 
main condenser.
    Date of issuance: May 9, 1997
    Effective date: May 9, 1997
    Amendment Nos.: 185 and 216
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17224) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 9, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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: March 14, 1997
    Brief description of amendment: The amendment extends the allowed 
outage time for its refueling water storage tank

[[Page 27805]]

while performing surveillance testing of its reactor coolant system 
pressure isolation valves (Surveillance 4.4.6.2.2).
    Date of issuance: May 6, 1997
    Effective date: May 6, 1997
    Amendment No. 71
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14459) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 6, 1997.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: April 18, 1997, as supplemented 
April 29, 1997.
    Brief description of amendment: The amendment approves the 
modification to the protection circuitry for emergency diesel 
generators. The associated Safety Evaluation delineates the staff's 
review and findings that the modification and related Final Safety 
Analysis Report (FSAR) changes are acceptable.
    Date of issuance: May 8, 1997
    Effective date: May 8, 1997
    Amendment No. 72
    Facility Operating License No. NPF-63. The amendment approves 
modification to the protection circuitry for emergency diesel 
generators and related FSAR changes.
    Date of initial notice and proposed no significant hazards 
consideration in Federal Register: (62 FR 19818 dated April 23, 1997). 
The notice provided an opportunity to submit comments on the 
Commission's proposed no significant hazards consideration 
determination. No comments have been received. The notice also provided 
for an opportunity to request a hearing by May 23, 1997, 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
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    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: November 4, 1996, as 
supplemented on December 4, 1996, and March 20, 1997.
    Brief description of amendments: The amendments revise the 
technical specifications (TS) to permit the removal of containment 
tendon sheathing filler grease in up to 35 tendons for Byron, Unit 1, 
and Braidwood, Unit 1, in advance of the steam generator replacement 
outages. The grease will be removed approximately 6 months prior to the 
respective steam generator replacement outages. In addition, in 
Amendment No. 80 issued on April 16, 1997, the title in Braidwood's TS 
6.9.1.7 was unintentionally left uncorrected. The corrected page is 
included in this amendment.
    Date of issuance: May 6, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 89, 89 and 81, 81
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 15, 1997 (62 FR 
2186). The March 20, 1997, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 6, 1997No significant hazards 
consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: February 17, 1997, as 
supplemented February 27, March 12, March 26, April 2, and April 10, 
1997
    Brief description of amendments: The amendments evaluate the 
Unreviewed Safety Question (USQ) associated with the use of containment 
pressure to compensate for the deficiency in Net Positive Suction Head 
(NPSH) for the Emergency Core Cooling System (ECCS) pumps following a 
Design Basis Accident (DBA). In the resolution of the USQ, the licensee 
changed the Updated Final Safety Analysis Report (UFSAR) in the 
following areas:
     1. containment analysis,
     2. decay heat model,
     3. increase in the suppression pool temperature and the effect on 
other associated systems following a DBA, and
    4. ECCS heat exchanger duty and containment cooling service water 
(CCSW) system flow.In addition, the proposed amendments would change 
the Technical Specification (TS) allowable water temperature limits for 
the suppression chamber and the ultimate heat sink from less than or 
equal to 75 degrees Fahrenheit to less than or equal to 95 degrees 
Fahrenheit. The original licensing basis water temperature for both the 
suppression chamber and ultimate heat sink was 95 degrees Fahrenheit. 
Both values were changed in the TS in Amendment Nos. 152 and 147 for 
Dresden, Units 2 and 3, respectively, issued on January 28, 1997. The 
amendments to lower the ultimate heat sink and suppression pool 
temperature limits in the TS was in response to the resolution of a USQ 
associated with the operation of Dresden, Units 2 and 3, following the 
discovery of a calculational error concerning the head loss across the 
ECCS suction strainers. The proposed amendments will return both units 
to normal operating conditions allowing for continued power operations 
when the ultimate heat sink temperature goes above 75 degrees 
Fahrenheit during warm weather.
    Date of issuance: April 30, 1997
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 157; 152.
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the licenses, TS and USFAR.
    Date of initial notice in Federal Register: February 27, 1997 (62 
FR 8998). The February 27, March 12, March 26, April 2 and April 10, 
1997, submittals provided additional clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated

[[Page 27806]]

April 30, 1997No significant hazards consideration comments received: 
No
    Local Public Document Room location:  Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Commonwealth Edison Company, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of application for amendments: February 17, 1997
    Brief description of amendments: The amendments would change the 
Technical Specifications by increasing the load test values of the 
emergency diesel generators in Surveillance Requirement 4.9.A.8.h from 
between 2625 kW and 2750 kW to 2730 kW and 2860 kW.
    Date of issuance: May 1, 1997
    Effective date: Immediately, to be implemented within 60 days.
    Amendment Nos.: 176 and 172
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14460). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 1, 1997.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: August 22, 1996, as supplemented 
March 28, 1997.
    Brief description of amendment: The amendment revises Technical 
Specification Sections 3.3 and 4.5 to allow the deletion of the 
requirement to utilize sodium hydroxide (NaOH) as an additive in the 
post-accident containment spray system.
    Date of issuance: April 23, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 191
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4345) The March 28, 1997, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the scope of the 
amendment request as originally noticed. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
April 23, 1997No significant hazards consideration comments received: 
No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Duke Power Company, et al., Docket Nos. 50-413 and 50-414, Catawba 
Nuclear Station, Units 1 and 2, York County, South Carolina

    Date of application for amendments: March 7, 1997, as supplemented 
by letters dated April 2, 10, 16, 22, and 28, 1997
    Brief description of amendments: The amendment revise Section 3/
4.7.1.6 of the Technical Specifications to require four instead of 
three steam generator pressure operated relief valves operable.
    Date of issuance: April 29, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days. Implementation of the amendments include the incorporation in 
the Updated Final Safety Analysis Report (UFSAR) of the changes to the 
description of the facility as set forth in the licensee's application 
dated March 7, 1997, as supplemented by letters dated April 2, 10, 16, 
22, and 28, 1997, as evaluated in the staff's Safety Evaluation dated 
April 29, 1997.
    Amendment Nos.:  159 and 151
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications and License Conditions.
    Date of initial notice in Federal Register: March 13, 1997 (62 FR 
11931) The April 2, 10, 16, 22, and 28, 1997, letters provided 
additional and clarifying information that did not change the scope of 
the March 7, 1997, application and the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 29, 1997.No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: September 30, 1994, as 
supplemented by letters dated September 18, 1995, and March 15, April 
29, May 16, September 23, and October 28, 1996, and January 16, April 
22, and May 2, 1997
    Brief description of amendments: The amendments revise the 
Technical Specifications related to the replacement of the Westinghouse 
Model ``D'' type preheat steam generators with feedring steam 
generators designed by Babcock and Wilcox International.
    Date of issuance: May 5, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days for Unit 1; and effective upon replacement of the steam 
generators for Unit 2.
    Amendment Nos.: 175 and 157
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56366) The March 15, April 29, May 16, September 23, and October 28, 
1996, and January 16, April 22, and May 2, 1997, letters provided 
clarifying information that did not change the scope of the September 
30, 1994, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated May 5, 1997.No significant hazards 
consideration comments received: No
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001

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

    Date of amendment request: November 15, 1996
    Brief description of amendment: The amemdment revises the technical 
specifications to allow the performance of the 24-hour emergency diesel 
generator maintenance run while the unit is in either Mode 1 or Mode 2.
    Date of issuance: May 5, 1997
    Effective date: May 5, 1997
    Amendment No.: 94
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
127) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 5, 1997.No significant hazards 
consideration comments received. No.
    Local Public Document Room location:  Government Documents

[[Page 27807]]

Department, Louisiana State University, Baton Rouge, LA 70803

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: April 11, 1997
    Brief description of amendment: The amendment would permit steam 
generator tubes with intergranular corrosion indications that may 
exceed through-wall limits to remain in service until the next 
refueling outage.
    Date of issuance: May 7, 1997
    Effective date: May 7, 1997
    Amendment No.: 189
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration (NSHC): Yes (62 FR 19628 dated April 
22, 1997). The notice provided an opportunity to submit comments on the 
Commission's proposed NSHC determination. No comments have been 
received. The notice also provided for an opportunity to request a 
hearing by May 22, 1997, but indicated that if the Commission makes a 
final NSHC 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 
NSHC are contained in a Safety Evaluation dated May 7, 1997.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    Local Public Document Room location:  Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of application for amendment: December 19, 1996
    Brief description of amendment: The proposed changes revise 
Technical Specification Table 4.3-1 to change the power calibration 
requirements for the linear power level, the Core Protection Calculator 
(CPC) delta T power and the CPC nuclear power signals between 15 and 80 
percent power to allow more conservative settings.
    Date of issuance May 5, 1997
    Effective date: May 5, 1997, to be implemented within 30 days.
    Amendment No.: 183
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4348) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 5, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: November 12, 1996, as 
supplemented November 27, 1996 (TSCR 224)
    Brief description of amendment: The amendment updates the technical 
specifications to reflect the implementation of the revised 10 CFR Part 
20, ``Standards for Protection Against Radiation.''

    Date of issuance : May 8, 1997
    Effective date:  May 8, 1997, with full implementation within 30 
days.
    Amendment No.:  191
    Facility Operating License No. DPR-16. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66708). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated May 8, 1997.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of application for amendment: February 7, 1997
    Brief description of amendment: The amendment revises TS 3.12, 
``Station Service Power,'' to require both 115 kV power circuits to be 
operable when the reactor is critical and to limit or restrict the time 
during which Maine Yankee may continue to operate if one or both of the 
115 kV power circuits become inoperable.
    Date of issuance May 2, 1997
    Effective date: May 2, 1997, to be implemented within 30 days.
    Amendment No.: 157
    Facility Operating License No. DPR-36: Amendment revised the 
Technical Specifications and/or License.
    Date of initial notice in Federal Register: February 26, 1997 (FR 
8799) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 2, 1997No significant hazards 
consideration comments received: No
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

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 24, 1997, as supplemented by 
letters dated March 13, April 11, 23, and 29, 1997
    Brief description of amendments: The amendments change the 
Technical Specification surveillance requirements for the Control Room 
Emergency Filtration System, the Penetration Room Filtration System, 
and the Containment Purge Exhaust Filter System.
    Date of issuance May 1, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 127 and 121
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: March 6, 1997 (62 FR 
10294) The March 13, April 11, 23, and 29, 1997, letters provided 
clarifying information that did not change the scope of the February 
24, 1997, application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated May 1, 1997.No 
significant hazards consideration comments received: No
    Local Public Document Room location:  Houston-Love Memorial 
Library, 212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 
36302

Tennessee Valley Authority, Docket Nos. 50-260, and 50-296, Browns 
Ferry Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: June 21, 1996, supplemented 
February 7, 1997 (TS 377)
    Brief description of amendments: The amendments provide a new 
minimum critical power ratio safety limit to replace a nonconservative 
value. Technical Specification Bases are also updated to clarify usage 
of the residual heat removal system supplemental spent fuel pool 
cooling mode.
    Date of issuance : May 7, 1997
    Effective date:  As of the date of issuance to be implemented 
within 30 days from the date of issuance.
    Amendment Nos.:  247 and 207

[[Page 27808]]

    Facility Operating License Nos. DPR-52 and DPR-68: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: 
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 7, 1997.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Athens Public library, 405 E. 
South Street, Athens, Alabama 35611

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: September 4, 1996, as 
supplemented February 3, 1997. The February 3, 1997 submittal provided 
clarifying information only, and did not change the proposed no 
significant hazards consideration determination.
    Brief description of amendments: The amendments revise the license 
and technical specifications (TS) to permit the insertion of four 
demonstration fuel assemblies into the reactor core of either North 
Anna 1 or North Anna 2, as described in the licensee's submittal. The 
four lead test assemblies, fabricated by Framatome Cogema Fuels, will 
incorporate several advanced design features, including: a debris 
filter bottom nozzle, mid-span mixing grids, a floating top end grid, a 
quick disconnect top nozzle, and use of advanced zirconium alloys for 
fuel assembly structural tubing and for fuel rod cladding.
    Date of issuance May 9, 1997
    Effective date: May 9, 1997
    Amendment Nos.: 204 and 185
    Facility Operating License Nos. NPF-4 and NPF-7. These amendments 
revised the License and Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64396) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 9, 1997.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.

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
Nuclear Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: July 18, 1996, as supplemented 
on January 29, 1997.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification 3.8, ``Refueling,'' and its 
associated Basis, by allowing the containment personnel air lock doors 
to remain open during refueling operations.
    Date of issuance May 7, 1997
    Effective date: May 7, 1997
    Amendment No.: 132
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1996 (61 FR 
42285). The January 29, 1997, submittal provided supplemental 
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 
May 7, 1997.No significant hazards consideration comments received: No.
    Local Public Document Room location:  University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.

Notice Of Issuance Of Amendments To Facility Operating LicensesAnd 
Final Determination Of No Significant Hazards ConsiderationAnd 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to

[[Page 27809]]

Facility Operating License, and (3) the Commission's related letter, 
Safety Evaluation and/or Environmental Assessment, as indicated. All of 
these items are available for public inspection at the Commission's 
Public Document Room, the Gelman Building, 2120 L Street, NW., 
Washington, DC, and at the local public document room for the 
particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By June 20, 1997, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket No. 50-265, Quad Cities Nuclear 
Power Station, Unit 2, Rock Island County, Illinois

    Date of application for amendment: April 29, 1997.
    Brief description of amendment: The proposed amendment modifies 
Section 5.3.A, ``Design Features'' of the Technical Specifications (TS) 
to reflect the ATRIUM-9B fuel design and would include various Siemens 
Power Corporation (SPC) topical reports in TS Section 6.9.A.6, ``Core 
Operating Limits Report,'' to reflect mechanical design criteria for 
this fuel and topical reports required for operation. This change would 
allow this fuel to be loaded into the core only under Operational Modes 
3 (Hot Shutdown), 4 (Cold Shutdown), and 5 (Refueling) and does not 
permit startup or power operation using the ATRIUM-9B fuel.
    Date of issuance May 2, 1997
    Effective date: May 2, 1997
    Amendment No.: 173
    Facility Operating License No. DPR-30: The amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated May 2, 1997.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603

[[Page 27810]]

    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    NRC Project Director: Robert A. Capra

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: May 2, 1997, as superseded May 5, 1997.
    Brief description of amendment: The proposed amendment relocates 
and revises the requirements for the control of the setpoint for the 
Standby Liquid Control system relief valves. The requirements would be 
relocated from Section 4.4.A.2.a and Bases Section 3.4.A of the Cooper 
Technical Specifications to the Updated Safety Analysis Report and the 
Inservice Testing Augmented Testing Program.
    Date of issuance May 9, 1997
    Effective date: May 9, 1997
    Amendment No.: 176
    Facility Operating License No. DPR-46: The amendment revised the 
Technical Specifications.Public comments requested as to proposed no 
significant hazards consideration: No.The Commission's related 
evaluation of the amendment, finding of emergency circumstances and 
final determination of no significant hazards consideration are 
contained in a Safety Evaluation dated May 9, 1997.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499
    NRC Project Director: William D. Beckner
    Dated at Rockville, Maryland, this 14th day of May, 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects III/IV, Office of Reactor 
Regulation
[Doc. 97-13190 Filed 5-20-97; 8:45 am]
BILLING CODE 7590-01-F