[Federal Register Volume 61, Number 31 (Wednesday, February 14, 1996)]
[Notices]
[Pages 5809-5827]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-3124]



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NUCLEAR REGULATORY COMMISSION

Biweekly Notice Involving No Significant Hazards Considerations; 
Applications and Amendments to Facility Operating Licenses

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 January 22, 1996, through February 2, 1996. 
The last biweekly notice was published on January 31, 1996 (61 FR 
3497).

Notice of Consideration of Issuance of Amendments To Facility Operating 
Licenses, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for a Hearing

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By March 15, 1996, 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 

[[Page 5810]]
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, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendments request: January 16, 1996.
    Description of amendments request: The proposed amendments would 
revise the Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, 
Technical Specifications (TSs) to adopt Option B of 10 CFR Part 50, 
Appendix J, to require Type A containment leak rate tests to be 
performed on a performance-based testing schedule. Specifically, TSs 3/
4.6.1.2 and 4.6.1.6.3 will be revised to reference a new Containment 
Leakage Rate Testing Program, TS 6.0 will be revised to add the new 
Containment Leakage Rate Testing Program, identify the programmatic 
controls for the new program, and reference the source of the 
programmatic guidelines, Regulatory Guide 1.116, ``Performance-Based 
Containment Leak-Test Programs,'' dated September 1995. The TS Bases 
will be revised to reflect these changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Containment leakage rate testing is performed in accordance with 
10 CFR Part 50, Appendix J, ``Primary Reactor Containment Leakage 
Testing for Water-Cooled Power Reactors.'' The Appendix J 
containment leakage test requirements include performance of Type A 
tests, which measure the overall leakage rate of the containment, 
and Type B and C tests, which measure the leakage through 
containment penetrations and valves. The Commission has amended the 
regulations to provide a 

[[Page 5811]]
performance-based alternative, Option B, to the existing Appendix J. At 
this time, Baltimore Gas and Electric Company plans to adopt Option 
B for Type A testing only.
    Implementation of Option B involves no physical or operational 
changes to the plant structures, systems or components. Furthermore, 
leakage rate testing and containment surface visual inspections do 
not contribute to the initiation of any postulated accidents; 
therefore, this proposed change does not involve an increase in the 
probability of any previously evaluated accidents.
    Type A testing is necessary to demonstrate that leakage through 
the containment is within the limits assumed in the accident 
analyses. The only potential effect of the proposed change to the 
Type A test frequency is the possibility that containment leakage 
would go undetected between tests. As described in NUREG-1493, 
passive failures resulting in containment leakage in excess of that 
assumed in the accident analyses are extremely unlikely to develop 
between Type A tests. Additionally, the Calvert Cliffs Individual 
Plant Examination considered the phenomenological effects associated 
with severe accidents which could lead to containment failure. It 
was concluded that adopting a performance-based testing interval 
will not significantly affect the containment failure probabilities 
calculated for the Individual Plant Examination. Furthermore, the 
required frequency for containment surface examinations to identify 
containment degradation precursors will be relocated from the 
Technical Specifications to the Containment Leakage Rate Testing 
Program, but will remain at three examinations every ten years as 
recommended by Regulatory Guide 1.163, September 1995. Altogether, 
adoption of a performance-based testing frequency, as specified in 
10 CFR Part 50, Appendix J, Option B, will not significantly 
decrease the confidence in the leak-tightness of the containment. 
Therefore, this change will not result in a significant increase in 
the probability of undetected containment degradation or in the 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed Technical Specification change adopts a 
performance-based approach to containment leakage rate testing. This 
change does not add any new equipment, modify any interfaces with 
any existing equipment, or change the equipment's function, or the 
method of operating the equipment. The proposed change does not 
affect normal plant operations or configuration, nor does it affect 
leakage rate test methods. As the proposed change would not change 
the design, configuration or operation of the plant, it could not 
cause containment leakage rate testing to become an accident 
initiator.
    Therefore, the proposed change does not create the possibility 
of a new or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
     The purpose of the existing schedule for Type A tests is to 
ensure that the release of radioactive material will be restricted 
to those leak paths and leakage rates assumed in the accident 
analyses. The margin of safety associated with containment leakage 
rate is not reduced if containment leakage does not exceed the 
maximum allowable leakage rate defined in the Technical 
Specifications. The proposed Technical Specification change 
implements a performance-based Type A testing option, but does not 
affect the maximum allowable containment leakage rate. The proposed 
change does not affect a safety limit, a Limiting Condition for 
Operation, or the way in which the plant is operated.
     In NUREG-1493, the Commission included a sensitivity study to 
explore the risk affect of several alternate leakage rate testing 
schedules. This study concludes that decreasing the Type A testing 
frequency to one test per twenty years would ``lead to an 
imperceptible increase in risk.'' Additionally, it was determined 
that implementation of the performance-based testing option will not 
significantly affect the containment failure probability calculated 
in the Calvert Cliffs Individual Plant Examination. Based upon these 
studies, there is sufficient information to conclude that the risk 
increase, and that the probability of exceeding the maximum 
allowable containment leakage rate as a result of adopting Option B, 
is low.
     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 three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: November 27, 1995.
    Description of amendment request: The proposed change would revise 
technical specification (TS) section 3.2 to remove requirements for the 
chemical and volume control system (CVCS). The CVCS requirements would 
be relocated to a licensee-controlled document and controlled by the 10 
CFR 50.59 evaluation process.
    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 removes the Chemical and Volume Control 
System (CVCS) requirements from the Technical Specifications (TS) 
and relocates these requirement[s] to a licensee-controlled 
document. As such, the proposed change only affects plant 
documentation and does not change the operating requirements or the 
plant physical or operating configuration. The CVCS requirements 
will be controlled by the plant approved process for the licensee-
controlled document using the 10 CFR 50.59 evaluation process. The 
proposed change relocating the CVCS requirements from the TS to 
licensee control will not affect the probability of an accident 
previously evaluated because the operating restrictions will remain 
in effect and any change to the operating restrictions will be 
performed in accordance with 10 CFR 50.59.
    Examination of the H. B. Robinson Steam Electric Plant, Unit No. 
2 Updated Final Safety Analysis Report (UFSAR) Chapter 15, Accident 
Analysis, finds that no CVCS structure, system, or component 
functions or actuates to mitigate a design basis accident or 
transient. Valves at the CVCS to Reactor Coolant System (RCS) 
interface perform a containment isolation function. However, the TS 
Section 3.2 does not address the containment isolation aspect of the 
CVCS. As such, the proposed change to remove the CVCS requirements 
from the TS will not affect 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 change removes the CVCS requirements from the TS 
and relocates the requirements to a licensee-controlled document. As 
such, the proposed change only affects plant documentation and does 
not change the operating requirements or the plant physical or 
operating configuration. The CVCS requirements will be controlled by 
the plant approved process for the licensee-controlled document 
using the 10 CFR 50.59 evaluation process. The proposed change will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated because any future change to 
these operating restrictions will be performed in accordance with 10 
CFR 50.59.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change removes the CVCS requirements from the TS 
based on the criteria of 10 CFR 50.36(c)(2)(ii). The CVCS 
requirements will be relocated to a licensee-

[[Page 5812]]
controlled document. As such, the proposed change only affects plant 
documentation and does not change operating requirements or the 
plant physical or operating configuration. The CVCS requirements 
will be controlled by the plant approved process for the licensee-
controlled document using the 10 CFR 50.59 evaluation process. The 
proposed change will not result in any reduction in the margin of 
safety because any future change to the CVCS operating restrictions 
will be performed in accordance with 10 CFR 50.59. 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 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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of amendment request: December 10, 1995.
    Description of amendment request: The proposed change would revise 
technical specification (TS) section 3.5.1 and Tables 3.5-2, 3, and 4 
concerning the reactor trip system (RTS), engineered safety feature 
actuation system (ESFAS), and isolation function. TS would be revised 
to (1) specify actions to be taken when an instrument channel becomes 
inoperable, (2) add an ``Applicable Conditions'' column that defines 
the applicability and/or mode of operation of each functional unit, and 
(3) make editorial enhancements.
    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 to upgrade the RTS and ESFAS TS to more 
closely agree with Westinghouse Standard TS (i.e., NUREG-0452) will 
not result in any hardware changes. The RTS and ESFAS are not 
assumed to be initiators of analyzed events.
    The role of these systems is in mitigating and thereby limiting 
the consequences of accidents. The proposed changes will ensure the 
RTS and ESFAS remain capable of mitigating design basis events as 
described in the Updated Final Safety Analysis Report (UFSAR) and 
that the results of the analyses in the UFSAR remain bounding. 
Additionally, the proposed changes do not impose any new safety 
analyses limits or alter the plant's ability to detect and mitigate 
events. Therefore, this change does 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 change to upgrade the RTS and ESFAS TS to more 
closely agree with Westinghouse Standard TS (i.e., NUREG-0452) does 
not necessitate a physical alteration of the plant (i.e., no new or 
different type of equipment will be installed) or changes in 
parameters governing normal plant operation. Thus, 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 the margin of safety.
    The proposed change, which upgrades the RTS and ESFAS TS to be 
consistent with Westinghouse Standard TS (i.e., NUREG-0452) does not 
involve a significant reduction in a margin of safety. The proposed 
change has been developed to ensure the analyzed safety limits are 
not exceeded and ensures the RTS and ESFAS are available when 
necessary to mitigate the consequences of accidents. It also imposes 
additional requirements to ensure the RTS and ESFAS remain capable 
of mitigating the consequences of design basis accidents as 
described in the UFSAR accident analyses. In addition, this change 
provides a benefit of avoiding unnecessary plant transients when 
adequate compensatory measures are available to ensure the intended 
function of the instrumentation is satisfied.

    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: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut, and Northeast Nuclear Energy 
Company, et al., Docket Nos. 50-245, 50-336, and 50-423, Millstone 
Nuclear Power Station, Unit Nos. 1, 2, and 3, New London County, 
Connecticut

    Date of amendment request: November 22, 1995.
    Description of amendment request: The amendments would revise the 
Technical Specifications (TS) for Haddam Neck and Millstone Unit Nos. 
1, 2, and 3 to be consistent with the guidance of Generic Letter 93-07. 
The proposed changes will remove review of the emergency and security 
plans from the TS list of responsibilities of the Plant Operations 
Review Committee (PORC)/Site Operations Review Committee (SORC), and 
will also remove the requirement for PORC/SORC to review procedures and 
procedure changes necessary for the implementation of the emergency and 
security plans.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    * * * The proposed changes do not involve an SHC [significant 
hazards consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
     The proposed changes delete the technical specification 
requirement to review the emergency plans, security plans, and their 
implementing procedures by PORC/SORC. The requirement which mandates 
PORC/SORC review will be maintained in the respective emergency plan 
and security plan. These changes are purely administrative in 
nature. These changes do not affect the configuration, operation, or 
performance of any system, structure, or component. The proposed 
changes are therefore not relevant to the probability of initiation 
of any accident previously evaluated, and they are not related to 
the prevention or mitigation of any accident previously evaluated. 
Thus they do not increase the consequences of any design basis 
accident.
    Therefore, these proposed changes to the Technical 
Specifications do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes delete from the technical specifications 
the line item requiring the review of emergency plans, security 
plans, and their implementing 

[[Page 5813]]
procedures by PORC/SORC. Revisions to these plans will continue to be 
reviewed by PORC/SORC due to commitments to contain the requirement 
for PORC/SORC review in the emergency plan and security plan. These 
changes are purely administrative in nature.
    None of the proposed changes described above alter the 
configuration, normal operation, design bases, function, or 
performance of any components or systems. Thus, the proposed 
administrative changes do not create the possibility of a new or 
different kind of accident from any previously evaluated since these 
changes do not introduce any new or different equipment, operating 
mode, or design basis functions for the existing licensed 
structures, systems and components. Thus, the proposed changes do 
not create the possibility of a new or different kind of accident 
from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    None of the above proposed changes alter the configuration, 
normal operation, design bases, function, or performance of any 
components or systems. Therefore, the proposed changes do not affect 
the margin of safety inherent in the design, analysis, function, or 
operation of the relevant structures, systems or components.
    These proposed changes do not alter the fuel clad barrier, fuel 
integrity, reactor coolant system integrity or the containment 
boundary integrity; thus no margin of safety related to these 
barriers is involved.
    None of the proposed administrative changes described above 
alter the configuration, normal operation, design bases, function or 
performance of any components, systems, or barriers to a 
radiological release. Thus, the proposed administrative changes do 
not affect the margin of safety inherent in the design, analysis, 
function, or operation of the relevant structures, systems or 
components.
    Based on the above, these 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457 for the Haddam Neck Plant, and Learning 
Resources Center, Three Rivers Community-Technical College, 574 New 
London Turnpike, Norwich, CT 06360 for Millstone Units 1, 2, and 3.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: January 11, 1996.
    Description of amendment request: The Catawba Unit 1 and the 
Catawba Unit 2 containment process penetration M308 and associated 
containment isolation valves are currently not in service and serve no 
function other than providing containment integrity. The licensee plans 
to implement modifications for both units to remove containment 
isolation valves RN-429A and RN-432B of penetration M308, remove 
associated wiring and control room instrumentation, and cut and cap 
tubing providing containment valve injection water to these containment 
isolation valves during the forthcoming Unit 1 refueling outage, 
currently scheduled to begin by June 1996, and the Unit 2 refueling 
outage currently scheduled to begin in March 1997. The proposed 
Technical Specifications (TS) would be revised to delete these 
containment isolation valves and associated equipment to permit 
implementation of these modifications. The licensee's requested 
amendment removes process penetration M308 from TS Table 3.6-1 and 
removes containment isolation valves RN-429A and RN-432B from TS Table 
3.6-2a and Table 3.6-2b due to planned modifications which physically 
remove these valves from process penetration M308.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

Criterion 1

    The physical removal of containment isolation valves RN-432B and 
RN-429A, associated control room instrumentation, containment valve 
injection water connections to these valves and the subsequent 
sealing of process penetration M308 will decrease unnecessary 
challenges to containment isolation, containment valve injection 
water leak-rate testing and the condition of control room 
instrumentation, as opposed to the current configuration.
    Since the sealing of process penetration M308 will be performed 
per the requirements of the applicable ASME code piping safety class 
requirements, the confidence in the pressure boundary will be 
equivalent to the component as originally designed. Therefore, this 
Technical Specification amendment to remove process penetration M308 
from Technical Specification Table[] 3.6-1 and to remove containment 
isolation valves RN-429A and 432B from Technical Specification Table 
3.6-2a and Table 3.6-2b will not increase the probability or 
consequences of an accident that has been previously evaluated.

Criterion 2

    Since no new failure modes are created, on the basis that the 
penetration is equivalent in confidence to the original design, and 
the plant will operate the same way it does now, this Technical 
Specification amendment to remove process penetration M308 from 
Technical Specification Table[] 3.6-1 and to remove containment 
isolation valves RN-429A and 432B from Technical Specification Table 
3.6-2a and Table 3.6-2b does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3

    This proposed change to Technical Specifications will not cause 
a significant reduction in the margin of safety. Upon completion of 
the removal of containment isolation valves RN-432B and 429A and the 
subsequent sealing of process penetration M308, the penetration will 
be Type B leak rate tested as part of post-modification testing, and 
will be retested periodically and following each use of the 
penetration for temporary containment cooling purposes during 
refueling outages. Therefore, the fuel, cladding, reactor coolant 
pressure boundary, and containment are not negatively affected by 
the proposed Technical Specification amendment. No assumptions made 
in any accident analysis are compromised by this proposed Technical 
Specification amendment.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242.
    NRC Project Director: Herbert N. Berkow.

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-389, 
St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: January 4, 1996.
    Description of amendment request: The proposed revisions rectify a 
discrepancy in Specification 3.5.3 for each St. Lucie unit, and provide 
assurance that administrative controls for High Pressure Safety 
Injection pumps remain effective in the lower operational modes.

[[Page 5814]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The amendment proposed for each St. Lucie Unit (1 and 2) 
rectifies an error in the Applicability statement for Technical 
Specification 3.5.3, which provides limiting conditions for 
operation (LCO) for the Emergency Core Cooling System (ECCS) 
subsystems during plant shutdown. The revision is administrative in 
nature and does not change the technical requirements within the LCO 
that are established to assure a minimum functional capability 
required of the ECCS systems to mitigate analyzed transients. 
Rather, the revision provides assurance that the effectiveness of 
certain administrative controls, established to restrict the number 
of operable HPSI [High Pressure Safety Injection] pumps during 
shutdown, will not be diminished by a misinterpretation of the modes 
and conditions for which the LCO must apply.
    This proposal does not create any accident initiators, nor does 
it change the availability or method of operation of equipment that 
is assumed to function in the success path(s) for mitigating 
accidents evaluated in the plant safety analyses. Therefore, 
operation of either facility in accordance with its proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed administrative change to the LCO 3.5.3 
Applicability statement for each St. Lucie unit will not change the 
physical plant or the modes of plant operation defined in the 
Facility License. The revision does not involve the addition or 
modification of equipment, nor does it alter the design or operation 
of plant systems. Therefore, operation of either facility in 
accordance with its proposed amendment would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed amendment involves an administrative change to LCO 
3.5.3 for each St. Lucie unit, which applies to the ECCS subsystems 
during the plant shutdown modes. The revision rectifies a 
discrepancy in the Applicability statement, and thereby provides 
assurance that the effectiveness of administrative controls 
established within the LCO to limit the number of operable High 
Pressure Safety Injection pumps during the shutdown modes will not 
be diminished. The changes do not alter the basis for any technical 
specification that is related to the establishment of, or the 
maintenance of, a nuclear safety margin. Therefore, operation of 
either facility in accordance with its proposed amendment would not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.
    Attorney for licensee: Harold F. Reis, Esquire, Newman and 
Holtzinger, 1615 L Street, NW., Washington, DC 20036.
    NRC Project Director: David B. Matthews.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: January 18, 1996.
    Description of amendment request: The proposed amendment would 
lower the Reactor Water Cleanup (RWCU) isolation setpoint from reactor 
low level to reactor low-low level.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed [technical specification] TS amendment will not 
significantly increase the probability or consequences of any 
previously evaluated accidents. The RWCU vessel level isolation 
occurs as a result of a [loss-of-coolant-accident] LOCA and 
therefore does not affect the probability of occurrence of a LOCA or 
any other previously evaluated accident.
    An IES calculation demonstrates that for all RWCU breaks or 
cracks considered, high ambient temperature, high differential 
temperature and/or high differential flow will provide the RWCU 
isolation signal prior to reaching reactor low level. Therefore, the 
level setpoint acts as a backup isolation signal for a break in RWCU 
piping outside primary containment.
    As discussed, this change will utilize four existing reactor 
level sensors. These reactor level sensors are safety related and 
located in the same physical area and in the same configuration as 
the four existing sensors. Therefore, the reliability of the RWCU 
vessel level isolation capability is not reduced.
    (2) The proposed changes will not create the possibility of a 
new or different kind of accident. The configuration of the RWCU 
isolation valves is unchanged. As before, the failure of any single 
active component in the new logic results in, at worst, failure of 
one containment isolation valve to close. Because the closure of one 
of the two valves is sufficient to achieve the containment 
isolation, the possibility of an accident of a different type is not 
increased.
    The modification to the RWCU vessel level isolation logic has 
been designed to the same standards as the original logic. This 
change will require the same surveillance requirements for the 
reactor low-low level trip point circuitry that are currently 
required for the reactor low level trip point circuitry. All other 
RWCU isolation functions remain unchanged. Consequently, no new 
accidents are postulated as a result of this proposed change.
    (3) The proposed change will not result in a significant 
reduction in any margin of safety. No margin of safety is affected 
by this change. The RWCU vessel level isolation occurs to establish 
primary containment and limit fluid loss. The proposed change will 
preserve these functions.
    It can be noted, however, that for a RWCU piping break outside 
primary containment, high ambient temperature, high differential 
temperature and/or high differential flow will provide the RWCU 
isolation signal. In the unlikely event that these temperature and 
flow sensing devices fail, isolation will be initiated upon reactor 
level reaching 119.5'' above [top of active fuel] TAF. Using 
blowdown rates and valve closure times, analysis shows reactor level 
will not drop below 105'' above TAF. The is well above the TAF. 
Additionally, lowering the RWCU isolation setpoint does not increase 
the consequences of a LOCA.

    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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of amendment request: January 30, 1996.
    Description of amendment request: The proposed amendment would 
revise certain control rod scram insertion time testing limits. The 
proposed change is compatible with the limits specified in the Improved 
Standard Technical Specifications (ITS), NUREG 1433, Revision 1, 
``Standard Technical Specifications, General Electric Plants, BWR/4.'' 

[[Page 5815]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) The proposed amendment does not involve a change in the 
probability or consequences of an accident previously evaluated. The 
amount of reactivity inserted at rod position 46 (approximately 5% 
of rod insertion) is small and the time required to insert this 
amount of reactivity is not explicitly considered in the plant 
transient analysis. A generic BWR/2-5 study (Reference 3 [EAS-56-
0889, ``BWR/2-5 Scram Time Technical Specification'', dated August 
1989]) performed on behalf of the [boiling water reactor] BWR 
Owner's Group to support the ITS demonstrated that relaxing the 5% 
rod insertion time requirement had a negligible impact on plant 
transient performance provided the insertion time requirements to 
the other rod positions are met. We have confirmed that this study 
is applicable to the [Duane Arnold Energy Center] DAEC. Increasing 
the allowable average scram insertion time to rod position 46 for 
all Operable control rods in addition to increasing the allowable 
average scram insertion time to rod position 46 for the three 
fastest control rods in any 2X2 array would still demonstrate that 
the [control rod drive] CRD system will perform its intended 
function. Scram time is a measure of CRD performance for 
operability. As such, it is not the initiator of any plant event. 
Therefore, the proposed change will not result in an increase in the 
probability of an accident occurring.
    (2) The amount of reactivity inserted at rod position 46 
(approximately 5% of rod insertion) is small and the time required 
to insert this amount of reactivity is not explicitly considered in 
the transient analysis. A generic BWR/2-5 study showed that relaxing 
the 5% rod insertion time requirement had a negligible impact on 
plant transient performance. Increasing the allowable average scram 
insertion time to rod position 46 for all Operable control rods, 
while increasing the allowable average scram insertion time to rod 
position 46 for the three fastest control rods in any 2X2 array, 
would still demonstrate that the CRD system will perform its 
intended function. Therefore, increasing the limits proposed does 
not create the possibility of a new or different kind of accident 
from any previously evaluated. Scram time is a measure of CRD 
performance for operability. As such, it is not the initiator of any 
plant event.
    (3) The safety limit most affected by an increase in scram times 
is the Minimum Critical Power Ratio (MCPR). The DAEC [technical 
specification] TS safety limit for MCPR is 1.07. To ensure that the 
MCPR safety limit is not exceeded during design basis transients and 
accidents, an operating limit is conservatively placed on the MCPR 
during normal plant operation (OLMCPR). The amount of reactivity 
inserted at rod position 46 (approximately 5% of rod insertion) is 
small. The analysis used to establish the OLMCPR does not consider 
the scram insertion time at position 46 but does consider the scram 
insertion time to rod position 38 for the most limiting transient 
(turbine load rejection without bypass). The required scram time to 
position 38 remains unchanged by this proposed amendment. A generic 
BWR/2-5 study showed that relaxing the 5% rod insertion time 
requirement had a negligible impact on plant transient performance. 
This change will not result in any changes to the calculated OLMCPR, 
which assures that the safety limit MCPR will not be exceeded. 
Therefore, this change will not reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Project Director: Gail H. Marcus.

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

    Date of amendment request: December 18, 1995.
    Description of amendment request: The Allowable Value for the 
Reactor Coolant Flow Instrumentation contained in Table 2.2-1 is 
proposed to be changed to reflect the design changes implemented during 
the last refueling outage. The Reactor Coolant System (RCS) Steam 
Generator Differential Pressure Instrumentation Loops have been 
modified to reflect a re-calibration of the differential pressure 
transmitter from ``-8 to 64 psid'' to ``0 to 35 psid,'' and an 
elimination of the Foxboro signal characterizer modules from the 
instrument loop string.
    Additionally, an editorial change is proposed for the text 
associated with the allowable value. The current wording ``reactor 
coolant'' is being changed to ``reactor coolant flow.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    Pursuant to 10 CFR 50.92, NNECO has reviewed the proposed 
changes. NNECO concludes that these changes do not involve a 
significant hazards consideration (SHC) since the proposed changes 
satisfy the criteria in 10 CFR 50.92(c). That is, the proposed 
changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to the Allowable Value of the Reactor 
Coolant Flow Instrumentation is based on design changes that reduce 
the uncertainties in the overall instrument loop, as well as 
improved calculation methodology for instrument uncertainty and 
setpoint. The new hardware configuration results in calculated 
uncertainties which are bounded by the Safety Analysis assumptions. 
There is no adverse impact on any design basis analysis due to this 
change, and, therefore does not affect the probability or 
consequence of any previously evaluated accident.
    Additionally, the proposed change to add the word ``flow'' is an 
editorial correction and therefore does not affect the probability 
or consequence of any previously evaluated accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The new Allowable Value has been calculated using an improved 
methodology. The new hardware configuration results in calculated 
uncertainties which are bounded by the Safety Analysis assumptions. 
The function of the Allowable Value is not changed. Therefore no new 
accident scenarios are created.
    Additionally, the proposed change to add the word ``flow'' is an 
editorial correction and therefore no new accident scenarios are 
created.
    3. Involve a significant reduction in a margin of safety.
    The change to the Allowable Value for the Reactor Coolant Flow 
Instrumentation reflects the design changes implemented during the 
last refueling outage. The design improvement of the loop 
performance ensures that the assumptions of the Safety Analysis are 
met. Since the proposed changes do not affect the consequences of 
any accident previously analyzed, there is no reduction in a margin 
of safety.
    Additionally, the proposed change to add the word ``flow'' is an 
editorial correction and has no effect on the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee.
    
[[Page 5816]]


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

    Date of amendment request: January 5, 1996.
    Description of amendment request: Northeast Nuclear Energy Company 
(NNECO) is proposing to implement the guidance of Generic Letter 93-08 
and relocate Tables 3.3-2, ``Reactor Protective Instrumentation 
Response Times'' and 3.3-5, ``Engineered Safety Features Response 
Times'' from the technical specifications to the Millstone Unit No. 2 
Technical Requirements Manual (TRM). In accordance with Generic Letter 
93-08, the Limiting Conditions for Operations for Technical 
Specifications 3.3.1.1, 3.3.2.1, and 3.7.1.6 are also proposed to be 
revised to eliminate their references to the aforementioned tables. 
NNECO has also proposed to revise Bases 3/4.3.1 and 3/4.3.2 to 
reference that the instrument response times are located in the TRM and 
that these tables in the TRM are now controlled under 10CFR50.59. NNECO 
also proposes to remove a cycle-specific note from Tables 3.3-3 and 
3.3-4.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    In accordance with 10CFR50.92, NNECO has reviewed the attached 
proposed changes and has concluded that they do not involve a 
significant hazards consideration. The basis of this conclusion is 
that the three criteria of 10CFR50.92(c) are not compromised. The 
proposed changes do not involve a significant hazards consideration 
because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed license amendment will remove the reactor 
protective system and engineered safety feature actuation response 
times from the technical specifications. This proposed change will 
not affect the operation of the reactor protective system and the 
engineered safety feature actuation system. Operability and 
surveillance requirements are still maintained in the technical 
specifications and the response times will be included and 
maintained in the Technical Requirements Manual (TRM). Once 
relocated to the TRM, any future proposed changes will require a 
safety evaluation and Plant Operations Review Committee review.
    The proposed license amendment will also delete the cycle-
specific note contained in Tables 3.3-2 and 3.3-4. This is 
administrative in nature and do not result in changes to plant 
configuration, operation, accident mitigation, or analysis 
assumptions. The notes was in effect only during Cycle 12.
    Since the systems will not be affected by the proposed changes, 
there is no impact on the performance of these systems or on the 
probability or consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    There are no new failure modes associated with the proposed 
changes. Since the plant will continue to operate as designed, the 
proposed changes will not modify plant responses to the point where 
it can be considered a new or different kind of accident.
    Involve a significant reduction in a margin of safety.
    The proposed changes do not have any adverse impact on the 
protective boundaries nor do they affect the consequences of any 
accident previously analyzed. The portion of the change associated 
with Generic Letter 93-08 will not affect the technical 
specification operability and surveillance requirements which will 
still ensure that the systems are tested and are within limits. 
Changing the limits requires a safety evaluation and Plant 
Operations Review Committee review. This will ensure that the 
licensing basis is maintained.
    The proposed changes to delete the cycle-specific notes are 
administrative in nature and do not result in changes to plant 
configuration, operation, accident mitigation, or analysis 
assumptions. The notes were in effect only during Cycle 12.
    Therefore, the proposed changes will not result in a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: January 26, 1996.
    Description of amendment request: The licensee proposes to modify 
the Technical Specifications for Millstone Unit No. 2 as follows:
    1. Limiting Condition for Operation 3.6.1.2.a-c: Replace the less 
than or equal to sign with a ``<'' sign for consistency with Appendix J 
wording on leakage limits.
    2. Surveillance Requirements:
    a. Type ``A'' tests: Surveillance Requirements 4.6.1.2.a-c are 
revised to replace specific guidance with a reference to the 
Containment Leakage Testing Program.
    b. Type ``B & C'' tests: Surveillance Requirement 4.6.1.2.d-e are 
revised to replace specific guidance with a reference to the 
Containment Leakage Testing Program.
    c. Air lock tests: Surveillance Requirements 4.6.1.3.a-c are 
revised to replace specific guidance with a reference to the 
Containment Leakage Testing Program.
    d. Containment Linear Plate Visual Inspection: Surveillance 
Requirement 4.6.1.6.3 is revised to replace specific guidance with a 
reference to the Containment Leakage Testing Program.
    e. Other Surveillance Requirements: 4.6.1.1.d and 4.6.1.2.g-h are 
replaced by the reference to the Containment Leakage Testing Program.
    3. Bases section 3/4.6.1.2 Containment Leakage is revised to 
reflect the above changes including a reference to the Containment 
Leakage Testing Program. In addition, the specific value of Pa is being 
deleted. Since Pa is a calculated value it is possible for the value of 
Pa to change should the loss of coolant accident be reanalyzed.
    4. Administrative Controls: Section 6.19 is added to establish a 
Containment Leakage Testing Program, as specified in Regulatory Guide 
1.163, dated September 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:

    Pursuant to 10CFR50.92, NNECO has reviewed the proposed use of 
10CFR50, Appendix J, Option B Containment Leak Rate Testing criteria 
for Millstone Unit No. 2. NNECO concludes that these changes do not 
involve a significant hazards consideration since the proposed 
change satisfies the criteria in 10CFR50.92(c). That is, the 
proposed changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The changes involved in this license amendment request revise 
the testing criteria for the containment penetrations. The revised 
criteria will be based on the guidance in Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.'' This guidance 
allows for the use of relaxed testing frequencies for containment 
penetrations that have performed satisfactorily on a historical 
basis. The Containment Leak Rate Testing 

[[Page 5817]]
Program considers the type of service, the design of the penetration, 
and the safety impact of the penetration in determining the testing 
interval of each penetration. The NRC Staff has reviewed the 
potential impact of performance-based testing frequencies for 
containment penetrations during the development of the Option B 
regulation. The NRC Staff review is documented in NUREG-1493 
``Performance-Based Containment Leakage Test Program.'' The review 
concluded that reducing the frequency of Type A tests (Integrated 
Leak Rate Tests) from three per ten years to one per ten years leads 
to an imperceptible increase in risk. For Type B and C testing 
(Local Leak Rate Tests), the change in testing frequency should not 
have significant impact since this leakage contributes less than 0.1 
percent of the overall risk based on the existing regulations. The 
use of Option B will allow the extension of testing intervals with a 
minimal impact on the radiological release rates since most 
penetration leakage is continually well below the specified limits. 
In the accident risk evaluation, the NRC Staff noted that the 
accident risk is relatively insensitive to the containment leakage 
rate because the accident risk is dominated by accident sequences 
that result in failure of or bypass of the containment. The use of a 
performance-based testing program will continue to provide assurance 
that the accident analysis assumptions remain bounding. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    Changes to the Administrative section describe the containment 
testing program only and cannot increase the probability or 
consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed license amendment does not change the operation or 
equipment of the plant. The change in the test frequency is 
dependent on the establishment of a Containment Leak Test Program. 
This test program will ensure the performance history of each 
penetration is satisfactory prior to the changing of any test 
frequency. Since the performance history of the penetration will be 
known, there is no possibility of the implementation of the program 
creating a new or different kind of accident than previously 
analyzed. Since there is no change to the equipment or the operation 
of the plant, there is no possibility of creating a new or different 
kind of accident than previously analyzed. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    Changes to the Administrative section describe the containment 
testing program only and cannot create a different accident from any 
previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    During the development of 10CFR50, Appendix J, Option B, the NRC 
Staff determined the reduction in safety associated with the 
implementation of the performance-based testing program. The results 
of this review are documented in NUREG-1493. The review concluded 
that reducing the frequency of Type A tests (Integrated Leak Rate 
Tests) from three per ten years to one per ten years leads to an 
imperceptible increase in risk. For Type B and C testing (Local Leak 
Rate Tests), the increase in testing frequency should not have 
significant impact since this leakage contributes less than 0.1 
percent of the overall risk-based on the existing regulations. The 
use of Option B will allow the extension of testing intervals with a 
minimal impact on the radiological release rates since most 
penetration leakage is continually well below the specified limits. 
In the accident risk evaluation, the NRC Staff noted that the 
accident risk is relatively insensitive to the containment leakage 
rate because the accident risk is dominated by accident sequences 
that result in failure of or bypass of the containment. The use of a 
performance based testing program will continue to provide assurance 
that the accident analysis assumptions remain bounding. Therefore, 
this change does not involve a significant reduction in the margin 
of safety.
    Changes to the Administrative section describe the containment 
testing program only and cannot reduce the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project 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: January 11, 1996.
    Description of amendment request: The proposed amendments would 
revise Section 6.0 (Administrative Controls) of the Salem and Hope 
Creek Technical Specifications to: (1) relocate the requirements of 
Section 6.5 (Station Operations Review Committee, Nuclear Safety Review 
and Audit, and Technical Review and Control) to the Quality Assurance 
Program, (2) replace specific management titles with generic management 
functional positions, (3) change Operating Engineer to Assistant 
Operations Manager, (4) require a Senior Reactor Operator license be 
held by either the Operations Manager or one of the Assistant 
Operations Managers, and 5) correct some typographical errors in 
Section 6.0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A portion of the proposed changes involves the relocation of the 
requirements for the Station Operations Review Committee, Nuclear 
Safety Review and Audit, and Technical Review and Control. These 
requirements are contained in Administrative Controls Section 6.5 of 
the Salem and Hope Creek Technical Specifications. The requirements 
to be relocated do not meet the criteria set forth in the 
Commission's Final Policy Statement for inclusion in Technical 
Specifications and therefore, may be relocated to an appropriate 
licensee controlled document (i.e., the Quality Assurance Program). 
Another element of the proposed change involves a modification which 
consists of stating that either the Operations Manager or Assistant 
Operations Manager shall hold a Senior Reactor Operator (SRO) 
license and replacing the title of Operating Engineer with Assistant 
Operations Manager.
    The requirements being changed are not required by 10 CFR 50.36 
and are not required to obviate the possibility of an abnormal 
situation or event giving rise to an immediate threat to the public 
health and safety. The changes are consistent with NUREG-1431 and 
NUREG-1433, Revision 1, and have been previously evaluated by the 
NRC. The remaining portions of the proposed changes consist of 
management title changes, including changing Operating Engineer to 
Assistant Operations Manager, and correction of typographical 
errors.
    All of the proposed changes are administrative in nature and do 
not affect assumptions contained in the plant safety analysis, the 
physical design and/or operation of the plant, nor do they affect 
Technical Specifications that preserve safety analysis assumptions. 
Implementation of these changes is expected to enable PSE&G [Public 
Service Electric & Gas] and the NRC to focus on requirements 
important to safety. Therefore, the proposed changes will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes are purely administrative and do not 
involve changes to operating procedures or physical modifications to 
the plants. Therefore, the proposed changes will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated. 

[[Page 5818]]

    3. Will not involve a significant reduction in a margin of 
safety.
    The changes discussed herein will not involve a significant 
reduction in a margin of safety since the proposed changes do not 
eliminate any existing Technical Specification requirements. All 
requirements removed from Technical Specifications are relocated to 
another licensee controlled program (i.e., the Quality Assurance 
Program). The Quality Assurance Program is controlled by existing 
regulations which provide a more appropriate vehicle for addressing 
changes and compliance. There are no administrative control 
requirements removed from the Technical Specifications which are not 
addressed by other regulations and regulatory requirements (i.e., 
10CFR50 Appendix B, 10CFR50.59, 10CFR50.54(a), and NUREG-0737).
    Prior to this proposed change it was a Technical Specification 
requirement that the Operating Engineer hold an SRO license.
    Specification 5.2.2.f of NUREG-1431 and NUREG-1433, Revision 1, 
states that an SRO license shall be held by either the Operations 
Manager or Assistant Operations Manager. The Operating Engineer and 
Assistant Operations Manager are equivalent positions at Salem and 
Hope Creek. Chapter 13 of the respective plant's Updated Final 
Safety Analysis Report, states that the Operations Manager is 
assisted by the Assistant Operations Manager (formerly the Operating 
Engineer) and other supervisory personnel. The Assistant Operations 
Manager reports directly to the Operations Manager and will assume 
the authority and responsibility of the department in the absence of 
the Operations Manager. The title change from Operating Engineer to 
Assistant Operations Manager reflects the organizational changes 
underway at Salem and Hope Creek. The duties and responsibilities 
associated with the two positions are identical. The option that 
either the Operations Manager or Assistant Operations Manager hold 
an SRO license is consistent with prior approved amendments for 
Salem and Hope Creek. These amendments [were] approved based on the 
fact that the organizational structure contained a direct report to 
the Operations Manager [who] is required to hold an SRO license. 
With the proposed change either the Operations Manager or a direct 
report (i.e., Assistant Operations Manager), is required to hold an 
SRO license. The change is also consistent with the 1993 version of 
ANSI/ANS 3.1, ``American National Standard for Selection, 
Qualification and Training of Personnel for Nuclear Power Plants,'' 
and NUREG-1431 and 1433, Revision 1. This change will not involve a 
significant reduction in a margin of safety since it is still 
required that either the Operations Manager or Assistant Operations 
Manager holds an SRO license.
    The other management title changes also will not involve a 
significant reduction in a margin of safety since all organizational 
responsibilities are and will continue to be implemented in 
accordance with applicable requirements.
    The proposed changes are administrative in nature and do not 
relate to or modify a margin of safety defined and maintained by the 
Technical Specifications. Therefore, the proposed changes will 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: 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 Nos. 50-272 and 50-311, 
Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem County, New 
Jersey

    Date of amendment request: January 4, 1996.
    Description of amendment request: The proposed amendments would 
change Technical Specification 3/4.8.2.5, ``28-Volt D.C. Distribution-
Operating.'' The amendments would make Unit 1 requirements similar to 
Unit 2 by defining the specific battery chargers that are required for 
each train and by restricting the use of the backup battery charger for 
a 7-day period. The amendments would also require the 28-Volt DC bus be 
energized for that bus to be OPERABLE.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
     The proposed changes do not alter plant configuration or 
operation. The proposed changes do not invalidate any of the 
parameters assumed in the UFSAR [Updated Final Safety Analysis 
Report] accident analyses. The proposed changes provide additional 
guidance to be used to ensure the operability of the safety related 
batteries, and requires the DC buses to be operable and energized 
consistent with the Limiting Condition for Operation (LCO). 
Operability of these buses provide control room instrumentation 
power in support of mitigating Design Basis Accidents.
    The changes to the Unit 1 Technical Specification (TS) 3.8.2.5 
LCO and Action Statements restrict the use of the backup battery 
chargers, thereby limiting the amount of time that the chargers are 
allowed to be powered from another AC Vital bus. This change brings 
the Unit 1 TS into agreement with Unit 2, and results in a more 
conservative Unit 1 TS since both alternate battery chargers are fed 
from the same 230 V vital AC bus.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed changes do not introduce any design or physical 
configuration changes to the facility, or change the function of the 
28-Volt DC Distribution System. Therefore, the proposed amendment 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes provide additional guidance to be used to 
ensure the operability of the safety related batteries. The changes 
to the Unit 1 Technical Specification (TS) 3.8.2.5 LCO and Action 
Statements restrict the use of the backup battery chargers, thereby 
limiting the amount of time that the chargers are allowed to be 
powered from another AC Vital bus. This change brings the Unit 1 TS 
into agreement with Unit 2, and results in a more conservative Unit 
1 TS by precluding the possibility of both the 1A and 1B battery/
buses from being supplied from a single bus. Therefore, the proposed 
amendment will 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, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

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: January 11, 1996.
    Description of amendment request: The proposed amendments would 
revise Section 6.0 (Administrative Controls) of the Salem and Hope 
Creek Technical Specifications to: (1) relocate the requirements of 
Section 6.5 (Station Operations Review Committee, Nuclear Safety Review 
and Audit, and Technical Review and Control) to the Quality Assurance 
Program, (2) replace specific management titles with generic management 
functional positions, (3) 

[[Page 5819]]
change Operating Engineer to Assistant Operations Manager, (4) require 
a Senior Reactor Operator license be held by either the Operations 
Manager or one of the Assistant Operations Managers, and (5) correct 
some typographical errors in Section 6.0.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    A portion of the proposed changes involves the relocation of 
requirements for the Station Operations Review Committee, Nuclear 
Safety Review and Audit, and Technical Review and Control. These 
requirements are contained in Administrative Controls Section 6.5 of 
the Salem and Hope Creek Technical Specifications. The requirements 
to be relocated do not meet the criteria set forth in the 
Commission's Final Policy Statement for inclusion in Technical 
Specifications and therefore, may be relocated to an appropriate 
licensee controlled document (i.e., the Quality Assurance Program). 
Another element of the proposed change involves a modification which 
consists of stating that either the Operations Manager or Assistant 
Operations Manager shall hold a Senior Reactor Operator (SRO) 
license and replacing the title of Operating Engineer with Assistant 
Operations Manager.
    The requirements being changed are not required by 10 CFR 50.36 
and are not required to obviate the possibility of an abnormal 
situation or event giving rise to an immediate threat to the public 
health and safety. The changes are consistent with NUREG-1431 and 
NUREG-1433, Revision 1, and have been previously evaluated by the 
NRC. The remaining portions of the proposed changes consist of 
management title changes, including changing Operating Engineer to 
Assistant Operations Manager, and correction of typographical 
errors.
    All of the proposed changes are administrative in nature and do 
not affect assumptions contained in the plant safety analysis, the 
physical design and/or operation of the plant, nor do they affect 
Technical Specifications that preserve safety analysis assumptions. 
Implementation of these changes is expected to enable PSE&G [Public 
Service Electric & Gas Company] and the NRC to focus on requirements 
important to safety. Therefore, the proposed changes will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes are purely administrative and do not 
involve changes to operating procedures or physical modifications to 
the plants. Therefore, the proposed changes will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The changes discussed herein will not involve a significant 
reduction in a margin of safety since the proposed changes do not 
eliminate any existing Technical Specification requirements. All 
requirements removed from Technical Specifications are relocated to 
another licensee controlled program (i.e., the Quality Assurance 
Program). The Quality Assurance Program is controlled by existing 
regulations which provide a more appropriate vehicle for addressing 
changes and compliance. There are no administrative control 
requirements removed from the Technical Specifications which are not 
addressed by other regulations and regulatory requirements (i.e., 
10CFR50 Appendix B, 10CFR50.59, 10CFR50.54(a), and NUREG-0737).
    Prior to this proposed change it was a Technical Specification 
requirement that the Operating Engineer hold an SRO license. 
Specification 5.2.2.f of NUREG-1431 and NUREG-1433, Revision 1, 
states that an SRO license shall be held by either the Operations 
Manager or Assistant Operations Manager. The Operating Engineer and 
Assistant Operations Manager are equivalent positions at Salem and 
Hope Creek. Chapter 13 of the respective plant's Updated Final 
Safety Analysis Report, states that the Operations Manager is 
assisted by the Assistant Operations Manager (formerly the Operating 
Engineer) and other supervisory personnel. The Assistant Operations 
Manager reports directly to the Operations Manager and will assume 
the authority and responsibility of the department in the absence of 
the Operations Manager. The title change from Operating Engineer to 
Assistant Operations Manager reflects the organizational changes 
underway at Salem and Hope Creek. The duties and responsibilities 
associated with the two positions are identical. The option that 
either the Operations Manager or Assistant Operations Manager hold 
an SRO license is consistent with prior approved amendments for 
Salem and Hope Creek. These amendments [were] approved based on the 
fact that the organizational structure contained a direct report to 
the Operations Manager [who] is required to hold an SRO license. 
With the proposed change either the Operations Manager or a direct 
report (i.e., Assistant Operations Manager) is required to hold an 
SRO license. The change is also consistent with the 1993 version of 
ANSI/ANS 3.1, ``American National Standard for Selection, 
Qualification and Training of Personnel for Nuclear Power Plants'', 
and NUREG-1431 and 1433, Revision 1. This change will not involve a 
significant reduction in a margin of safety since it is still 
required that either the Operations Manager or Assistant Operations 
Manager holds an SRO license.
    The other management title changes also will not involve a 
significant reduction in a margin of safety since all organizational 
responsibilities are and will continue to be implemented in 
accordance with applicable requirements.
    The proposed changes are administrative in nature and do not 
relate to or modify a margin of safety defined and maintained by the 
Technical Specifications. Therefore, the proposed changes will 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, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri.

    Date of amendment request: January 2, 1996.
    Description of amendment request: The proposed amendment would 
revise TS 3.9.4 and its associated Bases section to allow the 
containment personnel airlock doors to be open during core alterations 
and movement of irradiated fuel in containment. In addition, TS 
Surveillance Requirement 4.9.4 would be revised to specify that each 
containment penetration should be in its ``required position'' instead 
of a ``closed/isolated condition.''
    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 to TS 3.9.4 would allow the containment 
personnel airlock to be open during fuel movement and core 
alterations. The containment personnel airlock is currently closed 
during fuel movement and core alterations to prevent the escape of 
radioactive material in the event of a fuel handling accident.
    The containment airlocks are passive components integral to the 
containment structure and are not evaluated to be accident 
initiators; therefore, the proposed amendment does not involve an 
increase in the probability of an accident previously evaluated.
    The proposed change alters assumptions previously made in 
evaluating the radiological consequences of the fuel handling 
accident inside the containment building because the containment 
personnel airlock is assumed to be open. The 

[[Page 5820]]
radiological consequences described in this change are bounded by the 
Loss of Coolant Accident and General Design Criteria 19. All doses 
for the proposed change are less than the acceptance criteria, 
therefore, there is no significant increase in the consequences of 
an accident previously analyzed.
    In evaluating the consequences of this accident, NRC states in 
Section 15.4.6. of the Callaway Plant Safety Evaluation Report 
(NUREG-0830) that: ``The potential doses for the fuel handling 
accident are well within the guideline values given in 10 CFR Part 
100.'' Section II.1 of the Standard Review Plan defines ``well 
within'' to be 25% or less of the 10 CFR Part 100 exposure guideline 
values. NSAC 125, Guidelines for 10 CFR 50.59 Safety Evaluations, 
Section 3.6, states: ``If in licensing the plant the NRC explicitly 
found that the plant's response to a particular event was acceptable 
because the dose was less than the SRP guidelines (without further 
qualification), then the NRC implicitly accepted the SRP guideline 
as the licensing basis for the plant and the particular event, and 
the licensee may make changes that increase the consequences for the 
particular event, up to this value without prior NRC approval.'' 
Therefore, in the case of the fuel handling accident, NRC has 
implicitly accepted 25% of the 10 CFR Part 100 exposure guidelines 
as the acceptance limit.
    Since the probability of a fuel handling accident is unaffected 
by the airlock door positions, and the increased doses do not exceed 
acceptance limits, operation of the facility in accordance with the 
proposed amendment would not involve a significant increase in the 
probability or consequences of any 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 change to allow the containment personnel airlock 
to be open during core alteration and movement of irradiated fuel 
affects a previously evaluated accident (e.g., a fuel handling 
accident inside containment). The existing accident analysis has 
been modified to account for the containment personnel airlock doors 
being opened at the time of the accident. It does not represent a 
significant change in the configuration or operation of the plant. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is reduced when the offsite and control 
room doses exceed the acceptance criteria in General Design Criteria 
19 and the Standard Review Plan. As previously discussed in the 
response to Item 1, the offsite and control room doses are below the 
acceptance criteria. Therefore, operation of the facility in 
accordance with the proposed amendment would not involve a 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: January 19, 1996.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) for leak tests of containment 
isolation valves. The proposed amendment replaces the current specified 
surveillance intervals for containment leak testing with new 
surveillance requirements to conduct containment leak testing based on 
a performance-based containment leak test program. The licensee 
proposed use of performance-based testing in accordance with the 
revised 10 CFR Part 50 Appendix J (60 FR 49495), which would establish 
surveillance intervals based on the historical performance of the 
tested penetrations. In addition, the proposed amendment would extend 
the surveillance interval for leak testing of main steam isolation 
valves from the current 18 months to 30 months, consistent with 
Regulatory Guide 1.163.
    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.
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes modify the interval at which the containment 
leak rate testing is performed. The proposed change does not affect the 
containment leakage limits currently in the plant licensing basis and 
specified in the existing TS. Consequently, the radiological 
consequences of containment leakage during and after an accident are 
unchanged. The frequency of testing and the test methodology for 
containment leak rate testing are not identified as factors in the 
initiation, progression, or mitigation of any accident previously 
evaluated. The proposed change, therefore, does 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 change potentially affects the current surveillance 
intervals for conducting containment leak rate testing. A change in the 
length of the surveillance interval does not change the design or 
performance mode of structures, systems, or components, and thus 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 margin of safety for containment leakage is based on meeting 
the potential radiation exposure for occupational or postulated post-
accident conditions. The margin for WNP-2 is established by ensuring 
these exposures do not exceed 10 CFR Parts 20 and 100, respectively. 
Basing the surveillance intervals on containment leak rate performance 
is expected to lengthen the surveillance interval, thus the proposed 
change is expected to lower the cumulative occupational radiation 
exposure to conduct the leak rate testing.
    The performance criteria for the containment is based on ensuring 
that postulated post-accident radiation exposures remain within 10 CFR 
Part 100 limits. The proposed containment leak rate test program is 
based on ensuring that containment leakage is maintained below the 
level that will assure that radiation exposures resulting from 
postulated accident scenarios will remain below the regulatory limits. 
The length of time between tests will be based on historical 
performance of the tested penetrations. The change in test interval 
does not modify the current TS acceptance limits for containment 
leakage, and thus the proposed change does not involve a significant 
reduction in a margin of safety.
    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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352. 

[[Page 5821]]

    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502.
    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 
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.

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

    Date of amendment request: December 28, 1995.
    Brief description of amendment request: The proposed amendment 
would change Hope Creek Generating Station Technical Specification (TS) 
1.4, ``Channel Calibration'', to define actions required for channel 
calibration of instrument channels containing resistance temperature 
detector or thermocouple sensors.
    Date of publication of individual notice in Federal Register: 
January 5, 1996 (61 FR 420).
    Expiration date of individual notice: February 5, 1996.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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.

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

    Date of application for amendment: July 14, 1995, as supplemented 
September 12 and December 8, 1995.
    Brief description of amendment: The amendment changes the scram 
insertion times, Section 3.3.C, Minimum Critical Power Ration section, 
Section 4.11.C and, the associated Bases in Sections 2.1.1 and 3/4.4.3.
    Date of issuance: January 23, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 165.
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39443) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 23, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Plymouth Public Library, 132 
South Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, et al., Docket No. 50-324, Brunswick 
Steam Electric Plant, Unit 2, Brunswick County, North Carolina

    Date of amendment request: August 4, 1995.
    Brief description of amendment: The amendment changes the Technical 
Specifications to (1) reflect the use of a new type of fuel (GE13) and 
(2) modify the minimum critical power ratio safety limit and the 
standby liquid control system sodium pentaborate limits to accommodate 
the GE13 fuel.
    Date of issuance: January 31, 1996.
    Effective date: January 31, 1996.
    Amendment No.: 212.
    Facility Operating License No. DPR-62: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49931).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 1996.
    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 Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: September 13, 1995, as amended 
on November 27, 1995, and January 29, 1996.
    Brief Description of amendments: The amendments revise the 
Brunswick Steam Electric Plant, Units 1 and 2, Technical Specifications 
to permit the use of 10 CFR Part 50, Appendix J, Option B, Performance-
Based Containment Leakage Rate Testing.
    Date of issuance: February 1, 1996.
    Effective date: February 1, 1996.
    Amendment Nos.: 181 and 213.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications.
    Date of initial notice in Federal Register: December 12, 1995 (60 
FR 63739); repeated on January 3, 1996 (61 FR 188). The January 29, 
1996, amendment to the application provided supplemental information 
that was not outside the scope of the December 12, 1995 notice. 

[[Page 5822]]

    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 1, 1996.
    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, Docket No. 50-261, H. B. Robinson Steam 
Electric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: September 11, 1995.
    Brief description of amendment: Changes Technical Specification to 
add an allowance for Rod Insertion Limits (RILs) to be exceeded for a 
time no greater than the time criteria established by the axial power 
distribution methodology or 1 hour, whichever is sooner. An action is 
also added for the reactor to be placed in the hot shutdown condition 
within 6 hours if compliance with the RILs cannot be restored within 
the specified time period.
    Date of issuance: January 26, 1996.
    Effective date: January 26, 1996.
    Amendment No.: 167.
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54716).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 26, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant, Middlesex County, Connecticut

    Date of application for amendment: March 31, 1995, as supplemented 
November 14, 1995.
    Brief description of amendment: The amendment revises the Haddam 
Neck Technical Specifications (TS) to delete TS Sections 1.38 and 1.39, 
``Definitions, Fuel Assembly Types,'' revise TS Sections 3/4.9.3, 
``Refueling Operations, Decay Time'' and 3/4.9.14, ``Refueling 
Operations, Spent Fuel Pool--Reactivity Condition,'' replace TS 
Sections 5.6.1.1, ``Spent Fuel,'' and 5.6.3, ``Capacity,'' and add a 
new TS Section 3/4.9.15, ``Refueling Operations, Spent Fuel Pool 
Cooling.'' These changes support a rerack of the spent fuel pool to 
expand the spent fuel pool's storage capacity from 1168 assemblies to 
1480 assemblies so as to accommodate a full-core-discharge through the 
current validity date of the Haddam Neck Operating License (2007).
    Date of Issuance: January 22, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 6 months.
    Amendment No.: 188.
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1995 (60 FR 
25740).
    The November 14, 1995, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated January 22, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.

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: August 8, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification Table 4.4-4, ``Reactor Coolant Specific Activity Sample 
and Analysis Program,'' to allow reactor coolant system gross specific 
activity measurement method to be changed from the current degassed 
method to a non-degassed, or pressurized dilution, method.
    Date of issuance: January 22, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--141--Unit 2--135.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58400).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 22, 1996.
    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, 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: August 17, 1995.
    Brief description of amendments: The amendments revise Technical 
Specification Surveillance Requirement (SR) 4.2.5.2 to delete the 
requirement to calibrate the reactor coolant system (RCS) flowrate 
measurement instrumentation within 7 days prior to the performance of 
the flow measurement. Catawba Units 1 and 2 now utilize an RCS flowrate 
measurement method based on a one-time calibration of the cold leg 
elbow differential pressure taps as requested in the licensee's January 
10, 1994, application and as approved in License Amendments 128 and 122 
for Units 1 and 2, respectively. The January 10, 1994, application did 
not include a proposal to delete that portion of SR 4.2.5.2 which 
specifies that the measurement instrumentation shall be calibrated 
within 7 days prior to the performance of the flowrate measurement. 
This portion of the SR is now deleted since it only applies to the 
precision calorimetric heat balance method of RCS flowrate measurement.
    Date of issuance: January 23, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--142--Unit 2--136.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65676).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 1996.
    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: August 20, 1992, as 
supplemented by letter dated December 5, 1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications related to the 60-month 

[[Page 5823]]
125-volt surveillance requirement (SR). The change is to delete the 
words ``during shutdown'' from SR 4.8.2.1.2.e.
    Date of issuance: February 1, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--163--Unit 2--145.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65677).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 1, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of amendment request: May 30, 1995, as supplemented by letters 
dated November 20 and December 12, 1995.
    Brief description of amendment: The amendment revised the technical 
specifications for the drywell to permit bypass testing on a 10-year 
frequency with increased testing if performance degrades, changes the 
drywell air lock testing and surveillance requirements, deletes action 
notes for the drywell air lock and drywell isolation valves when the 
bypass leakage is not met, and deletes the specific leakage limits for 
the drywell air lock seal.
    Date of issuance: January 29, 1996.
    Effective date: January 29, 1996
    Amendment No.: 87.
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62490).
    The additional information contained in the supplemental letter 
dated December 12, 1995, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 29, 1996.
    No significant hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York

    Date of application for amendment: January 24, 1995
    Brief description of amendment: The amendment revises Technical 
Specification 3.4.1, ``Leakage Rate,'' and the associated Bases 
section. Specifically, the TS allowable Reactor Building leakage rate 
is reduced from 2000 cfm to 1600 cfm.
    Date of issuance: January 22, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 156.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11134)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 22, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Northeast Nuclear Energy Company, et al., Docket Nos. 50-245, 50-336, 
and 50-423, Millstone Nuclear Power Station, Unit Nos, 1, 2, and 3 New 
London County, Connecticut

    Date of application for amendments: August 4, 1995.
    Brief description of amendments: The amendments revise the 
Administrative Controls sections of the Technical Specifications for 
Millstone 1, 2 and 3 to allow the implementation of a Station Qualified 
Reviewer Program (SQRP) for the review and approval of selected 
procedures, programs and changes thereto.
    Date of issuance: January 17, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment Nos.: 91, 193, and 125.
    Facility Operating License Nos. DPR-21, DPR-65 and NPF-49: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45181)
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 17, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

PECO Energy Company, Public Service Electric and Gas Company, Delmarva 
Power and Light Company, and Atlantic City Electric Company, Docket 
Nos. 50-277 and 50-278, Peach Bottom Atomic Power Station, Unit Nos. 2 
and 3, York County, Pennsylvania

    Date of application for amendments: December 19, 1995.
    Brief description of amendments: These amendments change the 
ventilation filter test program bypass and penetration leakage test 
acceptance criteria from less than 0.05 percent to less than 1.0 
percent. The change corrects an administrative error that occurred 
during the development of the Peach Bottom Improved Technical 
Specifications which were issued as Amendments 210 and 214 to the Peach 
Bottom licenses on August 30, 1995.
    Date of issuance: January 16, 1996.
    Effective date: Unit 2, effective as of date of issuance, to be 
implemented concurrently with Amendment 210, issued August 30, 1995; 
Unit 3, effective as of date of issuance, to be implemented 
concurrently with Amendment 214, issued August 30, 1995.
    Amendments Nos.: 213 and 218.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (60 FR 66997, December 27, 1995). That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by January 26, 1996, 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 amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
January 16, 1996
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105. 

[[Page 5824]]


Philadelphia Electric Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: December 9, 1993, as 
supplemented by letters dated July 5, September 9, October 19, November 
15, and December 2, 1994, January 6, and January 23, 1995.
    Brief description of amendment: The amendment changes the Operating 
License and the corresponding Appendix A to reflect the planned 
implementation of the Power Rerate Program at the Limerick Generating 
Station, Unit 1, and the corresponding increase in the authorized 
maximum reactor core power level by five percent to 3458 megawatts 
thermal (MWt) from the current limit of 3293 MWt.
    Date of issuance: January 24, 1996.
    Effective date: As of date of issuance and to be implemented prior 
to startup in Cycle 7.
    Amendment No. 106.
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications and the licensee.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7695).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 24, 1996.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: June 20, 1995.
    Brief description of amendments: These amendments revise the 
Technical Specifications to reference 10 CFR Part 50, Appendix J, for 
the 1) Type A (Integrated Leakage Rate Test), and 2) Drywell-to-
Suppression Chamber (bypass) leakage tests instead of providing 
explicit requirements in the TS.
    Date of issuance: January 25, 1996.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos. 108 and 71.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42605).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 25, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendments: July 28, 1995.
    Brief description of amendments: The amendments modify Technical 
Specifications (TS) Surveillance Requirements 4.9.1.1, 4.9.1.2, 4.9.3, 
4.9.5, and 4.9.8 to delete specific requirements to perform 
surveillances just prior to beginning or resuming core alterations or 
control rod withdrawal associated with refueling activities. The 
amendments also delete the phrase ``incore instrumentation'' from the 
footnote in TS Section 3/4.9.5, ``Communication.''
    Date of issuance: January 31, 1996.
    Effective date: As of its date of issuance, to be implemented 
within 30 days.
    Amendment Nos.: 109 and 72.
    Facility Operating License Nos. NPF-39 and NPF-85: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49943).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 31, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: December 28, 1995.
    Brief description of amendment: This amendment changes Hope Creek 
Generating Station Technical Specification 1.4, ``Channel 
Calibration,'' to define actions required for channel calibration of 
instrument channels containing resistance temperature detector or 
thermocouple sensors.
    Date of issuance: January 25, 1996.
    Effective date: As of date of issuance.
    Amendment No.: 90.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (61 FR 420, January 20, 1996). That notice provided 
an opportunity to submit comments on the Commission's proposed no 
significant hazards consideration determination. No comments have been 
received. The notice also provided for an opportunity to request a 
hearing by February 5, 1996, 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 is contained 
in a Safety Evaluation dated January 25, 1996.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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

    Date of application for amendments: October 11, 1994, as 
supplemented December 13, 1994, September 6, 1995, and December 28, 
1995.
    Brief description of amendments: The amendments make two changes to 
Technical Specification 3/4.4.4 concerning pressurizer heaters. The 
first change adds the phrase ``capable of being powered from an 
emergency power supply'' to the Limiting Condition for Operation. The 
second change alters the frequency of surveillance requirement 4.4.4.2 
from 92 days to every refueling outage.
    Date of issuance: January 24, 1996.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos. 179 and 160.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60386).
    The December 13, 1994, September 6, 1995, and December 28, 1995, 
letters provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 24, 1996.
    No significant hazards consideration comments received: No. 
    
[[Page 5825]]

    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendment: May 1, 1995, supplemented 
December 20, 1995.
    Brief description of amendment: The amendment revises the technical 
specifications to eliminate selected response time testing requirements 
as described in the Boiling Water Reactor Owners' Group topical report, 
NEDO-32291, ``System Analyses for Elimination of Selected Response Time 
Testing Requirements,'' and to incorporate Generic Letter 93-08 
guidance regarding relocation of technical specification tables dealing 
with instrument response time limits.
    Date of issuance: January 11, 1996.
    Effective date: January 11, 1996, and implemented not later than 90 
days after issuance.
    Amendment No.: 77.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 23, 1995 (60 FR 
27345).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 11, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak Steam 
Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: January 5, 1996 (TXX-96007).
    Brief description of amendments: The amendments were processed as 
exigent amendments following issuance of a notice of enforcement 
discretion (NOED) by NRC letter dated January 11, 1996. The NOED and 
exigent Technical Specification (TS) amendments authorize the licensee 
to continue operating the Comanche Peak Steam Electric Station, Unit 2 
reactor at power with less than the minimum channels operable for Wide 
Range RCS (Reactor Coolant System) Temp. (Temperature)-Th remote 
shutdown indication. The minimum number of channels required is being 
revised from one per RCS Loop for each RCS Loop to one per RCS Loop for 
three of the four RCS Loops. These changes are only applicable to CPSES 
Unit 2 and are being submitted on the CPSES Unit 1 docket for 
administrative purposes only because the CPSES TSs is a single document 
which applies to both units.
    Date of issuance: February 2, 1996.
    Effective date: February 2, 1996.
    Amendment Nos.: Unit 1--Amendment No. 45; Unit 2--Amendment No. 31.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed significant hazards 
consideration: Yes (61 FR 1651, dated January 22, 1996). 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 February 21, 1996, but stated that any such hearing would 
take place after issuance of the amendment. The Commission's related 
evaluation of the amendments, finding of exigent circumstances, and 
final determination of no significant hazards consideration is 
contained in a Safety Evaluation dated February 2, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendment: October 18, 1995.
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant Technical Specification (TS) 3.4, ``Steam and Power 
Conversion System,'' by modifying and clarifying the operability 
requirements for the main steam safety valves (MSSVs), the auxiliary 
feedwater (AFW) System, and the condensate storage tank system. The 
amendment also eliminates inconsistencies within TS Section 3.4 and 
provides the basis for acceptable operation of the Auxiliary Feedwater 
System below 15% reactor power.
    Date of issuance: January 3, 1996.
    Effective date: January 3, 1996.
    Amendment No.: 123.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58407).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 3, 1996.
    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.

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

    Date of amendment request: October 18, 1995.
    Brief description of amendment: This amendment replaces the current 
fuel oil volume requirement in the emergency diesel generator (EDG) day 
tank in Technical Specifications 3.8.1.1.b.1) and 3.8.1.2.b.1) with a 
fuel oil level requirement. Associated Surveillance Requirement 
4.8.1.1.2.a.1) is also changed to replace the visual check requirement 
on fuel oil level in the day tank with a requirement to verify that the 
fuel oil transfer pump starts on low level in the day tank standpipe. 
The associated Bases section is also revised to reflect the above 
changes.
    Date of issuance: January 19, 1996.
    Effective date: January 19, 1996, to be implemented prior to 
startup from the eighth refueling outage currently scheduled to begin 
in March 1996.
    Amendment No.: 94.
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58049).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 19, 1996.
    No significant hazards consideration comments received: No.
    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.

[[Page 5826]]


Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By March 15, 1996, 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 

[[Page 5827]]
those facts or expert opinion. Petitioner must provide sufficient 
information to show that a genuine dispute exists with the applicant on 
a material issue of law or fact. Contentions shall be limited to 
matters within the scope of the amendment under consideration. The 
contention must be one which, if proven, would entitle the petitioner 
to relief. A petitioner who fails to file such a supplement which 
satisfies these requirements with respect to at least one contention 
will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of application for amendment: January 10, 1996.
    Brief description of amendment: The amendment granted a one-time 
extension for surveillances relating to the main steam isolation valve 
leakage control system, the reactor mode switch and manual scram of the 
reactor protection system, and the scram discharge vent and drain 
valves in order for the plant to operate for six more days until its 
planned shutdown date for refueling outage.
    Date of issuance: January 19, 1996.
    Effective date: January 19, 1996.
    Amendment No.: 78.
    Facility Operating License No. NPF-58: This 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 
January 19, 1996.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.
    Attorney for licensee: Jay E. Silberg, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, D.C. 20037.
    NRC Project Director: Gail H. Marcus.

    Dated at Rockville, Maryland, this 8th day of February 1996.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 96-3124 Filed 2-13-96; 8:45 am]
BILLING CODE 7590-01-P