[Federal Register Volume 59, Number 109 (Wednesday, June 8, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-13765]


[[Page Unknown]]

[Federal Register: June 8, 1994]


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NATIONAL FOUNDATION ON THE ARTS AND THE HUMANITIES
 

U.S. NUCLEAR REGULATORY COMMISSION

Biweekly Notice

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from May 16, 1994, through May 26, 1994. The last 
biweekly notice was published on May 25, 1994 (59 FR 27049).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11555 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC 
20555. The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By July 8, 1994, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendment requests: December 30, 1993
    Description of amendment requests: The proposed amendment would 
modify Tables 3.3-4 and 3.3-5 of Technical Specification 3/4.3.2, 
``Engineered Safety Features Actuation System Instrumentation,'' to 
provide clarification of settings for undervoltage relay trip values 
for the Class IE 4.16 kV electrical bus. The proposed amendment would 
also add Figure 3.3-1, ``LOSS OF VOLTAGE RELAY (GE IAV) TIME VS VOLTAGE 
CURVE,'' to clarify the relay setpoints and methodology.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1--Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS amendment does not significantly increase the 
probability of an accident previously evaluated. The methodology 
remains the same.
    Clarifying the minimum acceptable voltage and allowing more 
conservative values of the 4.16 kV bus undervoltage trip value will 
ensure that 4.16 kV ESF [Engineered Safety Features] bus voltages 
are sufficient to provide adequate voltage to equipment necessary 
for accident response.
    Standard 2--Create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
No new or different methodology is being proposed.
    The minimum undervoltage relay setpoints are specified and the 
new wording will allow a more conservative setpoint to ensure 
voltage levels to equipment powered by the 4.16 kV ESF bus.
    The proposed amendment also adds a figure to clarify the relay 
setpoints and methodology.
    Standard 3--Involve a significant reduction in a margin of 
safety.
    The margin of safety as defined in the TS will be increased by 
specifying the minimum and allowing more conservative values for the 
4.16 kV ESF bus undervoltage relay trip values. This clarification 
will ensure that the trip setpoint adequately protects that 
equipment powered by the 4.16 kV ESF bus from a potentially damaging 
degraded voltage condition.
    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: Theodore R. Quay

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

    Date of amendment requests: March 28, 1994
    Description of amendment requests: The proposed amendment would 
change Technical Specification 3/4.3.7.1.3 Condensate Storage Tank. The 
licensee proposed to change the minimum condensate storage tank (CST) 
indicated level from 25 feet to 29.5 feet to ensure that the CST 
contains sufficient volume of water. In addition, the licensee proposed 
to make an editorial change to the Unit 3 Technical Specification 3/
4.3.7.1.3, from ``with a level'' to ``with an indicated level,'' to be 
consistent with the Units 1 and 2 Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1--Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated based on the safety analysis for the CST [condensate 
storage tank] minimum indicated level. The proposed change increases 
the minimum indicated CST water level from 25 feet to 29.5 feet. 
Increasing the minimum indicated CST water level ensures that the 
requirements of BTP [Branch Technical Position] RSB [Reactor System 
Branch] 5-1 and UFSAR [Updated Final Safety Analysis Report] Section 
9.2.6 continue to be met. Therefore, this proposed change ensures 
that the consequences of an accident previously evaluated are not 
affected.
    Standard 2--Create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    This amendment request does not create the possibility of a new 
or different kind of accident from any accident previously analyzed 
since the minimum water volume of 300,000 gallons is maintained by 
this change. The change in the CST minimum required water level does 
not change the operation of any plant equipment while ensuring that 
the required 300,000 gallons are available. Since this change does 
not affect the operation of plant equipment and ensures that the 
minimum CST water inventory is maintained, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously analyzed.
    Standard 3--Involve a significant reduction in a margin of 
safety.
    The margin of safety presently provided is not reduced by the 
proposed change in the CST minimum required water level. The 
proposed change ensures that the CST volume of 300,000 gallons is 
available to satisfy the requirements of BTP RSB 5-1, UFSAR Section 
9.2.6 and the BASES for Technical Specification 3/4.3.7.1.3. 
Therefore, since the minimum required CST level in maintained, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensees' analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Attorney for licensees: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: Theodore R. Quay

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: May 20, 1994
    Description of amendment request: The proposed amendment will 
change the Technical Specifications (TS) to (1) add an exception to TS 
6.3.1 regarding the requirement for the Manager - Operations position 
to hold a Senior Reactor Operator's (SRO's) license, (2) add a new TS 
6.3.2 to describe the qualifications for the Manager - Operations and 
Manager - Shift Operations positions. This new section will require the 
Manager-Operations to hold or have held an SRO license for either the 
H. B. Robinson Steam Electric Plant, Unit No. 2 (Robinson), or a 
similar plant and require the Manager-Shift Operations to hold an SRO 
license at the Robinson plant, and (3) renumber the sections to allow 
for the insertion of a new TS 6.3.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The total number of senior reactor operator 
licensed personnel on shift remains unchanged. The change to provide 
a middle level of management will lessen the burden of the daily 
shift operations placed on the Manager - Operations. This will allow 
for more effective overall management of the Operations Unit. 
Requiring the Manager shift Operations to hold an SRO license will 
assure that supervision of operator activities continues to be 
performed by a senior licensed individual. The proposed changes to 
the Operations management organization do not involve physical 
alterations of the plant configuration or changes in setpoints or 
operating parameters. Therefore, there would be no increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As described above, these changes affect the organization 
of the Operations Unit. They do not represent any appreciable change 
in the current methodologies; they merely update the TS to reflect 
current personnel organization configuration and standards. The 
proposed changes to the Operations management organization do not 
involve physical alterations of the plant configuration or changes 
in setpoints or operating parameters. Therefore, the changes 
proposed do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety. The changes proposed do not 
reduce the number of senior reactor operator licensed personnel on 
shift. The changes to the Operations organization, as reflected in 
the proposed change, will enhance the overall effectiveness of plant 
operations and will serve to improve nuclear safety. There are no 
changes to the plant configuration or changes in setpoints or 
procedures. Therefore, the changes proposed have no affect on the 
facility's 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, 
Home and Fifth Avenues, 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: William H. Bateman

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

    Date of amendment request: May 10, 1994
    Description of amendment request: The proposed amendment would 
remove component lists from Technical Specification (TS) sections in 
accordance with the guidance provided in Generic Letter (GL) 91-08 
dated May 6, 1991. Related TS which reference the lists are also 
modified.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change will not result in any hardware or operating 
changes. The proposed change is based upon Generic Letter 91-08 and 
merely removes component lists, removes details relating to the 
component lists, provides clarifying information supporting the 
removal of the component listings, or removes details (which are 
considered administrative) that are no longer applicable to the TS. 
The removal of tabular component listings from the TS does not 
impact affected component OPERABILITY requirements. TS will continue 
to require the components to be OPERABLE. Action statements and 
surveillance requirements for the components will also remain in the 
TS. The tabular component lists will be relocated to plant 
procedures which will be controlled in accordance with the 
provisions specified in the Administrative Controls Section of the 
TS. Therefore, this change is administrative in nature and does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. Further, the proposed changes do 
not alter the design, function, or operation of the components 
involved and therefore, do not affect the consequences of any 
previously evaluated accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated. The proposed changes will not impose any different 
operational or surveillance requirements. The changes propose to 
relocate these component lists to plant procedures whereby adequate 
control of information is maintained. Further, as stated above, the 
proposed changes do not alter the design, function, or operation of 
the components involved and therefore, no new accident scenarios are 
created.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. The proposed change will not reduce a margin 
of safety because it has no impact on any safety analysis 
assumption. The proposed change does not alter the scope of 
equipment currently required to be OPERABLE or subject to 
surveillance testing nor does the proposed change affect any 
instrument setpoints or equipment safety functions. Therefore, the 
change does not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: July 15, 1993
    Description of amendment request: The revision proposed by 
Technical Specification Change Request (TSCR) No. 211 to the Technical 
Specifications would remove a footnote to specification 3.24, ``Reactor 
Coolant Inventory Trending System,'' and would revise that 
specification to be consistent with the revised Babcock & Wilcox 
revised Standard Technical Specifications, issued as NUREG-1430.
    The July 15, 1993, request supersedes the request of September 5, 
1991, published in the Federal Register on November 13, 1991 (56 FR 
57697).
    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 revised Limiting Condition for Operation represents no 
increase in the probability of occurrence or consequences of an 
accident previously evaluated.
    The TSCR represents no change to the physical configuration or 
operation of the Reactor Coolant Inventory Trending System.
    As stated in the Bases for the existing Technical Specification 
3.24, the system is not a system required to mitigate accidents. It 
may be useful to have the system operable, but adverse impact does 
not result if it is not operable; other useful information for 
monitoring inadequate core cooling is available. The change proposed 
is in accordance with the Limiting Condition for Operation in NUREG 
1430.
    2. The revised Limiting Condition for Operation does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    As identified above, the TSCR represents no change to the 
physical configuration or operation of the Reactor Coolant Inventory 
Trending System.
    3. The revised Limiting Condition for Operation does not involve 
a significant reduction in a margin of safety.
    The margin of safety for the proposed Limiting Condition for 
Operation is no different from that for existing Technical 
Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 11, 1994
    Description of amendment request: The purpose of the request is to 
revise the Technical Specifications (TS) by relocating the detailed 
inspection criteria, methods and frequencies of the containment tendon 
surveillance program to the Final Safety Analysis Report (FSAR) and 
providing a direct reference to the existing tendon surveillance 
program in the TS.
    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 of occurrence or the consequences of an accident 
previously evaluated. The proposed amendment only relocates the 
tendon surveillance program detailed requirements and criteria to 
the FSAR, consistent with the BWOG [Babcock & Wilcox Owners Group] 
Revised Standard Technical Specifications. The proposed amendment 
does not affect the requirement to verify the containment structural 
integrity in accordance with the inservice tendon surveillance 
program. The proposed Technical Specification specifies that the 
tendon surveillance program conforms to the recommendations of U.S. 
NRC Regulatory Guide 1.35, which ensures the effectiveness of the 
program is not reduced and containment structural integrity is not 
affected. Therefore, this change does not increase the probability 
of occurrence or the 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 tendon 
surveillance program is required to conform to the recommendations 
of U.S. NRC Regulatory Guide 1.35. Therefore, the effectiveness of 
the surveillance program is maintained, thus providing continued 
assurance of containment structural integrity.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The performance of the inservice tendon surveillance program 
is maintained in conformance with the recommendations of U.S. NRC 
Regulatory Guide 1.35, thus providing continued assurance of 
containment structural integrity. Therefore, it is concluded that 
operation of the facility in accordance with the proposed amendment 
does not involve a reduction in a margin of safety as defined in the 
basis of any Technical Specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 19, 1994
    Description of amendment request: The purpose of the request is to 
delete the quality assurance (QA) audit program frequency requirements 
from the Technical Specifications (TS) and to utilize the Operational 
Quality Assurance (OQA) Plan as the controlling document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. These changes do not affect the function of any system or 
component. Therefore, they do not increase the probability of 
occurrence or the consequences of an accident previously evaluated 
in the SAR [Safety Analysis Report].
    2. These changes do not involve a physical change to plant 
configuration and they do not affect the performance of any 
equipment. Therefore, they do not create the possibility of a new or 
different kind of accident or malfunction of a different type than 
previously identified.
    3. The shifting of the audit frequency requirements from the 
Technical Specifications to the OQA Plan and the extension of the 
maximum interval between audits of certain areas do not change the 
activities to be audited nor the scope of individual audits. 
Furthermore, audit frequencies are not associated with the marginof 
safety in the bases of any Technical Specification. Therefore, the 
margin of safety in not affected by this change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: April 14, 1994.
    Description of amendment request: The licensee proposes to revise 
Technical Specification 5.3.1, ``Fuel Assemblies.'' The amendment would 
permit fuel assembly reconstitution to restore the usefulness of fuel 
assemblies containing damaged or leaking fuel rods.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of a previously evaluated

    accident.
    Fuel assemblies containing filler rods will be shown to meet the 
current nuclear, mechanical, and thermal-hydraulic design limits on 
a cycle-specific basis. Since fuel assemblies containing filler rods 
will be shown to meet the current nuclear, mechanical, and thermal-
hydraulic design limits on a cycle-specific basis, there is no 
impact on the design basis of the plant.
    Replacement of fuel rods with fuel rods containing natural 
uranium or fuel rods from fresh or burned assemblies will be 
evaluated by South Texas Project's internal 10 CFR 50.59 review 
process. Reconstituted fuel assemblies with fuel rods containing 
natural uranium and/or fuel rods from fresh or burned assemblies 
will be analyzed using the normal design methodology as described in 
Section 6.9.1.6 of the Technical Specifications.
    Since current reload core design limits will be met by fuel 
assemblies using zirconium alloy or stainless steel filler rods, the 
proposed 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 accident from any previously evaluated.
    The use of filler rods does not involve any alteration to plant 
equipment or procedures which would introduce any new or unique 
operational modes or accident precursors. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The use of filler rods does not change the reload design or 
safety analysis limits for a reload core. Their use will be 
evaluated on a cycle-specific basis using NRC-accepted fuel rod 
configurations and analysis techniques. Since the safety analysis 
limits are unaffected and since the modified fuel assemblies will be 
shown to meet existing design limits on nuclear, mechanical, and 
thermal-hydraulic parameters, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: James E. Lyons, Acting

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

    Date of amendment request: April 29, 1994
    Description of amendment request: The licensee proposes to revise 
Technical Specification 3.7.1.1, ``Turbine Cycle - Safety Valves.'' The 
amendment would change the maximum allowable power range neutron flux 
high setpoint when one or more main steam safety valve (MSSV) is 
inoperative. The Bases for Technical Specification 3.7.1.1 would also 
be changed to reflect the new algorithm used to calculate the revised 
setpoint values.
    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 a previously evaluated 
accident. The Main Steam Safety Valves [MSSVs] are only actuated 
after a transient has occurred. Therefore, revising the maximum 
power level with inoperable MSSVs would not increase the probability 
of a previously evaluated accident. Following a loss of load/turbine 
trip, the revised Technical Specification Table 3.7-1 maximum 
allowable power range neutron flux high setpoints would ensure that 
the maximum power level allowed for operation with inoperable MSSVs 
is below the heat removing capability of the operable MSSVs. This 
would ensure that the design limit of 110% overpressurization is not 
exceeded. Therefore, there is no increase in the consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change would not alter the design, configuration, 
or method of operation of STP [South Texas Project]. For this 
reason, as well as the reasons stated in response to Criterion 1 
above, the proposed change would not create the possibility of a new 
or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The revised Table 3.7-1 setpoint values were calculated using a 
conservative method where the maximum power level allowed for 
operation with inoperable MSSVs is below the heat removing 
capability of the operable MSSVs. Using the revised maximum plant 
operating power levels will ensure that the secondary system 
pressure will be limited to within 110% of its design pressure. 
Therefore, since the design criteria will continue to be met, the 
proposed change does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: Suzanne C. Black

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

    Date of amendment request: April 27, 1994
    Description of amendment request: The amendment would delete 
Technical Specifications Surveillance Requirement 4.4.1.1.1, which 
requires that the reactor recirculation pump discharge valve be 
demonstrated operable by performing a full-stroke test of the valve 
prior to reactor thermal power exceeding 25% of rated reactor thermal 
power.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed Technical Specifications (TS) change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The proposed change to delete the TS Surveillance Requirement 
(SR) does not require any modifications to the plant or equipment, 
and will not impact the operation of the reactor recirculation 
system. The reactor recirculation system will continue to function 
as designed to maintain reactor pressure boundary integrity and to 
provide sufficient flow through the reactor core to remove heat from 
the fuel. This proposed TS change does not affect the operation of 
any Emergency Core Cooling System (ECCS) or other plant equipment 
important to safety. The purpose of this TS SR is to satisfy an ECCS 
operability requirement for Boiling Water Reactor (BWRs) where the 
reactor recirculation system piping serves as the injection flowpath 
to the reactor pressure vessel for the Low Pressure Coolant 
Injection (LPCI) system, an ECCS. Each LPCI subsystem, at LGS, has 
an independent flowpath which does not rely on the reactor 
recirculation system piping for injecting to the reactor pressure 
vessel. The reactor recirculation system pump discharge valves do 
not perform an active safety-related function and are classified as 
passive safety-related components designed to maintain the reactor 
pressure boundary integrity during reactor recirculation pump 
maintenance activities. These valves are not normally used during 
plant operations except to establish normal shutdown cooling, a 
manually initiated non-safety related function. These valves are not 
used to mitigate the consequences of design bases accidents.
    Therefore, the proposed change does not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed TS change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This proposed TS does not require physical changes to the plant 
or equipment, and does not impact any design or functional 
requirements of the reactor recirculation system, LPCI system, or 
other plant systems important to safety. The purpose for this TS SR 
is to satisfy an ECCS operability requirement for BWRs where the 
reactor recirculation system piping serves as the injection flowpath 
to the reactor pressure vessel for the LPCI system. The LPCI system 
at LGS has an independent flowpath for injecting to reactor pressure 
vessel and does not rely on the reactor recirculation system piping 
as part of the injection flowpath. Since the intent of this 
requirement is to support LPCI operation, and the LPCI system design 
function is accident mitigation, eliminating this TS SR has no 
impact on the types of accidents that could occur. The reactor 
recirculation system pump discharge valves do not perform an active 
safety-related function and are classified as passive safety-related 
components designed to maintain the reactor pressure boundary 
integrity during reactor recirculation pump maintenance activities.
    3. The proposed TS change does not involve a significant 
reduction in a margin of safety.

    The proposed change to delete the TS SR does not involve a 
change to the physical design or functional requirements of the 
reactor recirculation system, LPCI system, or other plant system 
important to safety. The reactor recirculation system will continue 
to function as designed to maintain reactor pressure boundary 
integrity and to provide sufficient flow through the reactor core to 
remove heat from the fuel. This proposed TS change does not impact 
the safety-related operation of the LPCI system. The LPCI system 
will continue to function as designed to mitigate the consequences 
of an accident. These valves do not perform an active safety-related 
function and are classified as passive safety-related components 
designed to maintain the reactor pressure boundary integrity during 
reactor recirculation pump maintenance activities.
    Therefore, the proposed TS change does not involve a reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

Philadelphia Electric Company, Public Service Electric and Gas 
Company,Delmarva Power and Light Company, and Atlantic City 
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: April 27, 1994
    Description of amendment request: These technical specifications 
(TS) changes are being proposed to support the implementation of 
proposed Modification 5274 which is intended to replace the Peach 
Bottom Atomic Power Station (PBAPS), Unit 2 Containment Atmospheric 
Dilution (CAD) System and Containment Atmospheric Control (CAC) System 
analyzers. This modification was performed on PBAPS, Unit 3 during the 
previous Unit 3 outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The design function and operation of the CAC and CAD Systems, 
which are supported by the operation of the containment monitoring 
system, have not been altered as a result of these changes. The CAC 
System monitors the content of oxygen during startup and normal 
operation and the CAD System is utilized to monitor the content of 
hydrogen and oxygen during post-LOCA [loss-of-coolant accident] 
operation. The monitoring of these variables will continue to 
mitigate the consequences of accidents previously evaluated. 
Additionally, no accident precursors will be impacted by these 
changes.
    The new system meets or exceeds the design standards of the 
original system. Additionally, the decrease in warmup time will 
increase the availability and usefulness of the analyzers to 
mitigate the consequences of an accident. Therefore, the proposed 
changes will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from any accident previously evaluated; or,
    The proposed TS changes do not involve the introduction of any 
new accident initiators. The new containment monitoring system will 
enhance the ability of [the] CAD system to mitigate the consequences 
of an accident and prevent the introduction of a new or different 
type of accident previously evaluated. The new system meets or 
exceeds existing design standards and will be tested to ensure its 
reliability. The new containment monitoring system is a monitoring 
system and will not introduce new accident initiators. Therefore, 
the proposed changes will not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Although the number of analyzers is being reduced, the proposed 
modification and TS changes will enhance the ability of the 
containment monitoring system to support the operation of the CAC 
and CAD systems [through] the use of improved equipment that meets 
or exceeds the design standards of the original system. Therefore, 
the proposed changes 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: Charles L. Miller

Power Authority of The State of New York, Docket No. 50-286, Indian 
PointNuclear Generating Unit No. 3, Westchester County, New York

    Date of amendment request: May 10, 1994
    Description of amendment request: The proposed amendment would 
revise Section 3.1.C.3 and Table 4.1-1 of Appendix A of the Operating 
License. This change would require that the reactor coolant average 
temperature (Tavg) be restored to greater than or equal to 
540 deg. F within a 15-minute period or be in hot shutdown within the 
following 15 minutes. The proposed change in Table 4.1-1 entitled, 
``Minimum Frequencies for Checks, Calibrations and Tests,'' will add 
the requirement for Tavg instrument check frequency to be reduced 
to 30 minutes when the Tavg deviation and low Tavg alarms are 
not reset and the control banks are above zero steps. This application 
also proposes revision to the Bases to reflect that the minimum 
temperature for criticality provides assurance that the reactor is 
operated within the bounds of the safety analyses. In addition, the 
proposed application also includes an administrative change to correct 
some typographical errors on page 3.1-25 of the Technical 
Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?

Response:  

    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The probability or the consequences of an 
accident previously evaluated will not be affected because the 
proposed changes will make the Minimum Temperature for Criticality 
Specification (540 deg. F) more restrictive than the current 
specification which allows reactor criticality at a temperature as 
low as 450 deg. F. The proposed changes will also make the minimum 
temperature for criticality consistent with the licensing basis 
safety analyses. In addition, critical operation at Tavg less 
than 540 deg. F will require operator response to restore Tavg 
to [greater than or equal to] 540 deg. F within 15 minutes or be in 
hot shutdown within the following 15 minutes. [***] [T]he minimum 
temperature for criticality when defined to be within 7 deg. F below 
the no-load Tavg value of 547 deg. F does not adversely affect 
pressurizer operability, reactor vessel nil-ductility temperature, 
the reactor protection system operability, nor the plant design 
basis analyses and is supported by the current licensing basis 
safety analyses. The presence of two separate alarms, each 
annunciating on a 1-out-of-4 Tavg signal, will provide 
assurance that constant Tavg monitoring is available during 
approaches to criticality. The proposed change also increases the 
surveillance frequency for Tavg instrument check when the 
Tavg deviation and low Tavg alarms are not reset and the 
control banks are above zero steps. Therefore, the proposed changes 
have no effect on the probability or consequences of an accident 
previously evaluated.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?

Response:  

    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed changes do not involve the addition of any 
new or different type of equipment, nor do they involve the 
operation of equipment required for safe operation of the facility 
in a manner different from those addressed in the Final Safety 
Analysis Report. The safety analyses, which assume a critical 
temperature of 547 deg. F, are applicable for critical temperatures 
as low as 540 deg. F. The proposed changes will ensure that the 
plant parameters are within their analyzed ranges and will increase 
the surveillance frequency for the Tavg instrument check when 
the Tavg deviation and low Tavg alarms are not reset and 
the control banks are above zero steps. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?

:Response:  

    The proposed license amendment does not involve a significant 
reduction in a margin of safety. The proposed changes do not affect 
any safety related system or component operation or operability, 
instrument operation, or safety system setpoints and do not result 
in increased severity of any of the accidents considered in the 
safety analyses. Operator response to a drop in temperature after 
reaching criticality for a specified period of time will place the 
reactor in the hot shutdown condition where the LCO [limiting 
condition for operation] does not apply. The proposed changes are 
being made to make the Technical Specifications consistent with the 
licensing basis safety analyses and increase the surveillance 
frequency. These changes have no effect on any margin of safety and, 
therefore, do not create 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Michael L. Boyle, Acting

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

    Date of amendment request: March 31, 1994
    Description of amendment request: This amendment request would 
revise the diesel fuel oil storage operability requirements and the 5 
minute diesel hot restart test conditions. In addition, the amendment 
request also revises the Surveillance Requirements to allow the 24-hour 
diesel generator endurance test to be conducted during any operational 
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. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The addition of a 48 hour period to complete restoration of the 
required fuel oil level prior to declaring the diesel generator 
inoperable does not significantly increase the probability or 
consequences of an accident previously evaluated. PSE&G believes 
that the attendant risk of maintaining the diesel generator OPERABLE 
status under temporary conditions where fuel oil supply is below 
48,800 gallons (but greater than 44,709 gallons) is less than the 
attendant risk of initiation and completion of shutdown actions 
currently required by Technical Specifications under these 
conditions. Since a minimum 44,709 gallon [6 day] supply of oil will 
be maintained for these 48 hours, and procedures are implemented to 
obtain replenishment fuel oil when the level falls below 48,800 
gallons of fuel, and the probability of an event requiring the 
onsite power sources during this brief period are low (as stated in 
NUREG-1433), PSE&G concludes that this change does not increase the 
likelihood of accidents occurring nor significantly affect the 
performance of any system involved in the occurrence or mitigation 
of the accidents.
    The proposed amendment to allow the 24 hour diesel generator 
endurance run to be conducted during any mode of operation does not 
significantly increase the probability or consequences of an 
accident previously evaluated in Chapter 15 of the UFSAR since the 
capability to safely shutdown the plant following a Loss of Offsite 
Power (LOP), LOCA or LOCA/LOP coincident with a single failure is 
maintained throughout the surveillance test. The 24 hour endurance 
test does not disable any of the automatic actuations and interlocks 
of the diesel generator control functions, nor prevent the 
satisfactory completion of the LOP or LOCA/LOP loading sequence if a 
LOP or LOCA signal is received at any time during the test. Required 
Class-1E onsite power OPERABILITY during normal operation, shutdown 
cooling, loss of off-site power, and accident conditions will be the 
same.
    In addition, the performance of proposed Surveillance 
Requirement 4.8.1.1.2.k.1 during Operational Conditions 1 or 2 will 
not significantly increase the consequences of perturbations to any 
of the electrical distribution systems that could result in a 
challenge to steady state operation or to plant safety systems. 
Performance of proposed Surveillance Requirement 4.8.1.1.2.k.1 
during Operational Conditions 1 or 2, or failure of the 
surveillance, will not cause, or result in, an anticipated 
operational occurrence with attendant challenges to plant safety 
systems that has not been previously analyzed for the existing 
monthly surveillances.
    Therefore, PSE&G concludes that this above change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The establishment of the new Surveillance Requirement 
4.8.1.1.2.k.2, allows the five minute diesel hot restart test to be 
performed at times other than after the 24 hour endurance run test, 
but does not alter nor modify the test requirements currently 
required by Surveillance Requirement 4.8.1.1.2.h.4.b. The proposed 
Surveillance Requirement 4.8.1.1.2.k.2 maintains the test conditions 
required by current Specifications, and satisfies the intent of 
Regulatory Guide 1.9, Rev. 3, paragraph 2.2.10. Therefore, PSE&G 
believes that this change 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 accident previously evaluated.
    The proposed amendment does not involve any physical changes to 
plant structures, systems or components, or change the manner in 
which the plant is operated. Therefore, the proposed changes will 
not increase the probability of accidents of a different type, nor 
will they create malfunctions of a different type than any 
previously evaluated in the SAR.
    3. Will not involve a significant reduction in a margin of 
safety. The basis for this statement is outlined in Item 1 above.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Charles L. Miller

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

    Date of amendment request: March 31, 1994
    Description of amendment request: The amendment request would 
establish an allowed out-of-service time (AOT) of 72 hours for any one 
low pressure Emergency Core Cooling System injection subsystem 
inoperable in addition to an inoperable High Pressure Coolant Injection 
system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below

    :1. Will not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change establishes a 72 hour Allowed Out-of-Service 
Time (AOT) for coincidental High Pressure Coolant Injection system 
(HPCI) and one low pressure injection/spray subsystem inoperability, 
in accordance with NUREG-1433. In addition, PSE&G will justify 
retaining the existing HCGS AOTs for one, two, and three Low 
Pressure Coolant Injection (LPCI) subsystems inoperable.
    The HCGS electrical distribution system that supplies the ECCS 
(Emergency Core Cooling System) equipment with power contains four 
(4) Emergency Diesel Generator (EDGs) while the electrical 
distribution system of the NUREG-1433 plant contains only three (3) 
EDGs. This additional, redundant system adds reliability to the 
overall electrical distribution system. In NUREG-1433, the loss of 
one EDG (Emergency Diesel Generator) results in the inoperability of 
two (2) Low Pressure Injection (LPCI) pumps or one (1) LPCI and one 
(1) 100% capacity CS subsystem, whereas the loss of one EDG at HCGS 
would only result in the inoperability of one LPCI pump and one-half 
of a CS subsystem (50% capacity). Another design feature of the HCGS 
LPCI system that makes it more reliable than the plant discussed in 
NUREG-1433 is that the LPCI system injects directly into the reactor 
vessel via water boxes mounted inside the core shroud. The LPCI 
system of the NUREG-1433 plant injects into the reactor vessel via 
the recirculation loops, 2 LPCI pumps inject into each recirculation 
loop. Therefore, following a rupture of a
    recirculation line, a Design Base Accident Loss of Coolant 
Accident, the NUREG-1433 plant would lose 2 LPCI subsystems, while HCGS 
would maintain all 4 LPCI subsystems available for injection. The end 
result is that, following assumed loss of a single EDG electrical 
system, the HCGS plant is left with much more injection capability than 
the NUREG-1433 plant. For these reasons:
    *the HCGS AOT for one LPCI subsystem inoperable, given at least 
one CS subsystem is operable, is 30 days as opposed to 7 days as 
specified in NUREG-1433, and
    *the HCGS AOT for two LPCI subsystems inoperable, given at least 
one CS subsystem is operable, is 7 days. This AOT is the same as the 
AOT in NUREG-1433 for one (1) LPCI subsystem inoperable. However, 
this AOT is less restrictive than the action required in NUREG-1433 
for one (1) LPCI subsystem inoperable with one (1) CS subsystem 
inoperable, which is immediate entry into Technical Specification 
3.0.3., and
    *the HCGS AOT for three LPCI subsystems inoperable, given both 
CS subsystems are operable, is 72 hours as opposed to immediate 
entry into Technical Specification 3.0.3 as specified in NUREG-1433.
    NUREG-1433 would be overly conservative for HCGS, since HCGS has 
four (4) EDGs and four (4) physically separate channels of 
electrical power, while the NUREG-1433 plant only has 3 EDGs. HCGS
    has greater capability to supply power to its ECCS in an emergency 
than the NUREG-1433 plant. Similarly, the incorporation of an AOT of 72 
hours for HPCI and one low pressure injection/spray subsystem as 
proposed in NUREG-1433 would also be conservative for HCGS.
    In the HCGS safety analysis, the small break LOCA is the design 
basis event for the HPCI system. The most limiting single failure 
coincident with a large or small break LOCA is the failure of the DC 
source (Channel A) common to the HPCI system, one CS subsystem, and 
one LPCI subsystem. The change made by this submittal would 
establish a 72 hour AOT for either one LPCI subsystem or one CS 
subsystem to be inoperable coincident with HPCI being inoperable. As 
discussed above, if an initiating event occurs while in the proposed 
AOT for the ECCS, the resulting scenario would be within the design 
basis of the HCGS. Even with a failure of one CS subsystem while 
HPCI and LPCI are in a 72 hour LCO or a failure of one LPCI 
subsystem while HPCI and CS are in a 72 hour LCO, HCGS would still 
be within its design basis if an initiating event occurred. Adequate 
core cooling is ensured by the operability of the ADS and the 
remaining low pressure injection/spray subsystems.
    The loss of feedwater flow transient analyzed in Chapter 15 of 
the HCGS UFSAR assumes both the HPCI and RCIC systems to be 
operable. The existing TS for HPCI allows the system to be 
inoperable for up to 14 days. Sufficient injection capability is 
assured because the RCIC system will be required to be operable if 
the HPCI is inoperable. For this transient, the proposed 72 hour AOT 
for the HPCI and one low pressure injection/spray subsystem 
inoperable presents no challenge greater than the existing 14 day 
HPCI AOT.
    For the steam line breaks outside of containment, the HPCI 
system is assumed unavailable. For feed line breaks outside of 
containment, either the HPCI or RCIC systems are capable of 
providing adequate cooling to the vessel to prevent cladding damage. 
The RCIC system will be required to be operable during the proposed 
72 hour AOT.
    The Core Damage Frequency (CDF) at HCGS is calculated based on 
Probabilistic Risk Assessment (PRA) models using present Technical 
Specifications that do not allow simultaneous outages of either HPCI 
and one LPCI subsystem or HPCI and one CS subsystem. The proposed 
action statement, which would allow 72 hours of plant operation with 
the HPCI system inoperable coincident with one low pressure ECCS 
injection/spray subsystem (LPCI or CS) inoperable, was determined to 
have a potential impact on 28 of the core damage sequences. These 28 
core damage sequences were requantified with a new model (based on 
the proposed action statement that would allow the simultaneous 
outage), and none of the frequencies of the 28 sequences were 
significantly affected.
    In addition, the probability of an accident is not affected 
because no physical modifications are being made to the plant.
    Finally, implementation of this proposed change would reduce 
unnecessary plant shutdowns without commensurate effects on safety 
and minimize associated challenges to safety systems.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Establishing an AOT of 72 hours for the HPCI system inoperable 
in addition to any one (1) low pressure injection/spray subsystem 
does not alter the function of the equipment nor involve any type of 
plant modification. Additionally, no new modes of plant operation 
are involved with these changes. The proposed change therefore will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The basis for this statement is outlined in item 1 above.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Charles L. Miller

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

    Date of amendment request: April 25, 1994
    Description of amendment request: The proposed amendment would 
change the minimum level in the Emergency Diesel Generator (EDG) Fuel 
Oil Day Tank from 200 gallons to 360 gallons. The change is necessary 
to support a required run time of 55-60 minutes from the day tank. The 
Bases would also be changed to include reference to Amendment 59, which 
authorized a change to the EDG surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The operation of the Hope Creek Generating Station (HCGS) in 
accordance with the proposed change will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment to the Technical Specifications will 
bring the minimum allowable EDG fuel oil day tank levels into 
agreement with the levels corresponding to EDG run times reviewed 
and approved by the NRC in Amendment 59 to the Hope Creek Generating 
Station Facility Operating License. Increasing the minimum level 
requirement is conservative and when compared to the current level 
requirement, will act to reduce the consequences of any accident or 
malfunction of equipment important to safety previously evaluated in 
the Updated Final Safety Analysis Report (FSAR). The proposed change 
will have no impact on the probability of any accident.
    2. The operation of the Hope Creek Generating Station (HCGS) in 
accordance with the proposed change will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The physical changes to the plant and to the manner in which the 
plant is operated in the proposed revisions add conservatism to the 
current requirements in the Technical Specifications. No new or 
different kind of accident is created by the proposed change.
    3. The operation of the Hope Creek Generating Station (HCGS) in 
accordance with the proposed change does not involve a significant 
reduction in a margin of safety.

    The proposed revision, by maintaining a higher (more 
conservative) Emergency Diesel Generator Day Tank level, will result 
in an increase in margins of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Charles L. Miller

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

    Date of amendment request: April 25, 1994
    Description of amendment request: The proposed amendment would 
eliminate the requirement from the Hope Creek technical specifications 
to perform Type C leak rate tests in accordance with 10 CFR 50, 
Appendix J, of identified containment isolation valves that penetrate 
the primary containment and terminate below the minimum water level in 
the suppression chamber (torus). The valves would still be subject to 
testing in accordance with the American Society of Mechanical Engineers 
Boiler and Pressure Vessel Code and Type A leak rate tests (Integrated 
Leak Rate Test) of the primary containment.
    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 Containment Isolation Valves (CIVs) for which Appendix J, 
Type C leak rate testing will no longer be performed are all on 
lines which penetrate the Torus and terminate below the Torus 
minimum water level. Since the Torus is designed and operated to be 
filled with water during and following any postulated Design Basis 
Accident, the CIVs will remain water sealed during these conditions.
    Type C testing of individual CIVs per the requirements of 
10CFR50 Appendix J is not necessary since no potential containment 
bypass leakage path exists due to the water seal and closed system 
piping. The CIVs will, however, continue to be tested pursuant to 
the applicable requirements of Section XI of the American Society of 
Mechanical Engineers Boiler and Pressure Vessel (ASME B&PV) Code.
    In addition, the elimination of 10CFR50 Appendix J, Type C 
testing for the subject CIVs will not affect the overall leak-tight 
integrity of the Primary Reactor Containment, which will be 
demonstrated during 10CFR50 Appendix J Type A Integrated Leak Rate 
Testing as described in the Hope Creek Generating Station Updated 
Final Safety Analysis Report (HCGS UFSAR) Section 6.2.6.1.
    Consequently, radiological releases and their consequences due 
to leakage of the subject CIVs will be minimized and within the 
existing plant licensing basis.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This proposal does not involve any hardware or logic changes, 
nor does it alter the way in which any plant systems operate. Post-
accident Containment isolation features, boundaries and system 
interfaces are not affected by the changes. Therefore, there are no 
new possibilities or types of accidents considered.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed elimination of 10CFR50 Appendix J, Type C testing 
for certain CIVs in lines which penetrate the Torus and terminate 
below the minimum water level will not adversely affect the margins 
of safety associated with the plant's licensing bases.
    The water seal provided by the Torus, in conjunction with the 
closed system piping, precludes post-accident bypass leakage. 
Individual CIVs will be tested in accordance with the ASME B&PV Code 
Section XI - Division 1, Article IWV-3000, as required. 10CFR50 
Appendix J, Type A testing will ensure that the overall Containment 
leakage rate is consistent with the plant's licensing bases. In 
addition, any leakage associated with the subject CIVs will have 
little radiological significance since it will involve liquid 
releases (i.e., minimum fission products). CIV seat leakage will be 
confined within the closed system piping downstream of the CIVs. 
Existing Containment isolation features, boundaries and system 
interfaces are not affected by the changes.
    Since the CIVs and the systems they serve are all located in the 
Reactor Building, any leakage (e.g., packing gland leakage) which 
escapes the confines of the closed system piping, will be contained 
within the Reactor Building. The Reactor Building is a 
radiologically controlled area which is served by the Filtration, 
Recirculation, and Ventilation System. This assures that all 
radioactive releases to the environment are within the existing 
plant licensing bases. The elimination of Type C testing for the 
subject isolation valves will not affect the existing radiological 
release evaluations currently described in the HCGS-UFSAR.
    The proposed Amendment will not affect the functional capability 
of any plant safety-related structures, systems or components nor 
will it result in any relaxation of existing plant licensing bases.
    In addition, the implementation of the proposed Amendment will 
result in a reduction of radiological exposure to plant personnel. 
Therefore, the proposed revision will not reduce 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: Charles L. Miller

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: March 28, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.3.2, Table 3.3-4, Item 7b, Sustained 
Degraded Voltage for Salem Generating Station Units 1 and 2. The 
amendment proposes to change the Sustained Degraded Voltage Trip 
Setpoint from greater than or equal to 91.6% of bus voltage for less 
than or equal to 13 seconds to greater than or equal to 94.6% of bus 
voltage for less than or equal to 13 seconds, and the Allowable Value 
for Sustained Degraded Voltage from greater than or equal to 91% of bus 
voltage for less than or equal to 15 seconds to greater than or equal 
to 94% of bus voltage for less than or equal to 15 seconds.
    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 or malfunction of equipment important to 
safety previously evaluated.
    The proposed amendment will increase the minimum voltage 
available at the vital buses and maintain vital loads within their 
voltage ratings under degraded voltage conditions.
    The change to the Sustained Degraded Voltage Trip Setpoint also 
considered the minimum bus recovery voltage following a transient 
that would reset the undervoltage relay to prevent unnecessary 
transfers to the emergency diesel generators. Therefore, the 
proposed amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated in 
the UFSAR.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed amendment to the Sustained Degraded Voltage Trip 
Setpoint ensures a higher minimum voltage is available to all vital 
loads during any electrical transient, and is sufficient to operate 
these loads within their voltage rating. Therefore, the proposed 
amendment does 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 amendment will increase the minimum voltage 
available at the vital buses and maintain vital loads within their 
voltage ratings. This ensures that the minimum voltage for any vital 
load will continue to be available including under degraded voltage 
conditions. Calculations have determined that a Sustained Degraded 
Voltage Trip Setpoint of greater than or equal to 94.6% of bus 
voltage for less than or equal to 13 seconds is the optimum setpoint 
with an Allowable Value of greater than or equal to 94% of bus 
voltage for less than or equal to 15 seconds to meet the current 
design requirements. These values provide sufficient margin based on 
the minimum voltage of 93.2% required to ensure that vital loads 
will operate within their designed voltage range. Therefore, the 
proposed amendment does not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Charles L. Miller

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, and 
Docket No. 50-354, Hope Creek Generating Station, Salem County, New 
Jersey

    Date of amendment request: April 13, 1994
    Description of amendment request: The proposed changes revise the 
Quality Assurance audit frequencies in the Hope Creek and Salem Unit 
Nos. 1 and 2 Technical Specifications. These revisions will permit a 
biennial audit frequency with a 25% extension to the frequency and 
transfer subsequent control over the audit program to the Updated Final 
Safety Analysis Report. 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 likelihood that an accident will occur is neither increased 
or decreased by the proposed Technical Specification changes which 
affect review and audit frequencies. The proposed changes will not 
impact the function or method of operation of plant equipment. Thus, 
there is not a significant increase in the probability of a 
previously analyzed accident due to the changes. Also, the 
consequences of a malfunction of equipment important to safety 
previously evaluated in the UFSAR is not increased by the changes.
    The proposed changes affect review and audit frequencies. As 
such, the proposed changes have no impact on accident initiators or 
plant equipment, and thus, do not affect the probabilities or 
consequences of an accident.
    The audit program verifies that functions and methods of plant 
operation have not been altered or degraded. This verification 
process will continue, but on a more flexible basis. The proposed 
changes will not alter the function or diminish the quality of the 
audits. Frequency may only be decreased for those programs which 
demonstrate acceptable performance. The changes provide added 
flexibility in scheduling audits and facilitate resource allocation 
to areas with perceived weaknesses. In this regard, the proposed 
changes offer an opportunity for possible decreases in the 
likelihood that an accident would occur.
    Therefore, we conclude that the proposed changes do not 
significantly increase the probabilities or consequences of an 
accident.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes do not involve changes to the physical 
plant or operations. Since program audits do not contribute to 
accident initiation, changes related to audit functions cannot 
produce a new accident scenario or produce a new type of equipment 
malfunction. Also, the [changes do] not alter any existing accident 
scenarios. The proposed changes do not affect equipment or its 
operation, and, thus, do not create the possibility of a new or 
different kind of accident.
    The proposed changes revise the audit frequency to biennial and 
permit a 25% extension to this frequency. However, audit frequency 
may only be decreased for those programs exhibiting acceptable 
performance. The 25% extension is a provision to provide added 
flexibility in scheduling audits and is not applicable to the audits 
of the Facility Emergency and Security Plans.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed changes concerning conduct of reviews and audits do 
not directly affect plant equipment or operation. Safety limits and 
limiting safety system settings are not affected.
    The proposed changes will not alter the function or diminish the 
quality of the reviews and audits. For the audit program, the 
changes propose a biennial audit frequency and a subsequent transfer 
of control over the audit program to the Updated Final Safety 
Analysis Report. This will not result in a loss of regulatory 
control. 10 CFR 50.54(a)(3) requires that changes to the quality 
assurance description report which reduce commitments be submitted 
to the NRC prior to implementation. In addition, the audit frequency 
may only be decreased for those programs which demonstrate 
acceptable performance. The changes provide added flexibility in 
scheduling audits and facilitate resource allocation to areas with 
perceived weaknesses. The proposed changes should result in a more 
effective audit program that will contribute to an improvement in 
plant safety.
    Therefore, use of the proposed Technical Specification changes 
would not involve any 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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079 (Salem Generating Station, Units 
1 and 2) and Pennsville Public library, 190 S. Broadway, Pennsville, 
New Jersey 08070 (Hope Creek Generating Station)
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Charles L. Miller

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: May 3, 1994
    Description of amendment request: The proposed amendments would 
modify the Technical Specification (TS) for Combustible Gas Control (3/
4.6.4.1) by changing the surveillance frequency for performing the 
channel functional test to once-per-quarter and the channel calibration 
to once-per-refueling. Also, the TS for the Auxiliary Feedwater System 
(3/4.7.1.2) would be changed to reduce the surveillance frequency for 
performing pump operability tests to once-per-quarter on a staggered 
test basis. These changes are consistent with the provisions of Generic 
Letter 93-05, ``Line-Item Technical Specifications Improvements to 
Reduce Surveillance Requirements For Testing During Power Operation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes to the surveillance requirements for the 
hydrogen analyzers and the Auxiliary Feedwater pumps are consistent 
with the intent of Generic Letter 93-05, Line-Item Technical 
Specification Improvements to Reduce Surveillance Requirements for 
Testing During Power Operation, and NUREG-1366, Improvements to 
Technical Specification Surveillance Requirements. The proposed 
changes will modify surveillance frequency for both the hydrogen 
analyzers and the Auxiliary Feedwater Pumps. Changing the 
surveillance frequency for the hydrogen analyzers and the Auxiliary 
Feedwater pumps does not affect the probability of occurrence or the 
consequences of accidents identified in the UFSAR. No accident 
precursors are being generated by the proposed increase in 
surveillance frequency. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
a previously analyzed accident.
    2. Create the possibility of a new or different kind of 
accident.
    The proposed changes to the surveillance requirements for the 
hydrogen analyzers and the Auxiliary Feedwater pumps are consistent 
with Generic Letter 93-05 and NUREG-1366. There are no modifications 
or changes in operating conditions associated with the proposed 
changes. Therefore, the proposed changes will not increase the 
possibility of a new or different kind of accident from any accident 
previously identified.
    3. Involve a significant reduction in a margin of safety.
    The Technical Specification operability requirements for the 
hydrogen analyzers and the Auxiliary Feedwater pumps are not being 
changed. Surveillance testing will still be performed on a routine 
frequency. The proposed frequency will be capable of performing its 
intended function and ensuring a consistent degree of reliability. 
Therefore, the changes to the surveillance frequencies do not 
involve a significant reduction in any 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: Charles L. Miller

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of amendment request: September 29, 1993 (TS 333)
    Description of amendment request: The proposed amendment would 
increase the amount of Boron-10 required to be available in the Standby 
Liquid Control System (SLCS).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The Standby Liquid Control System (SLCS) is designed to make the 
reactor subcritical from rated power to a cold shutdown at any time 
in core life with the control rods remaining withdrawn in the rated 
power pattern. The increase in the amount of Boron-10 that is 
required to be stored in the SLCS Solution Tank does not affect the 
precursors for any accident or transient analyzed in Chapter 14 of 
the BFN Final Safety Analysis Report (FSAR). Since there is no 
change to an accident precursor, there is no increase in the 
probability of any accident previously evaluated.
    The increase in the amount of Boron-10 required to be stored in 
the SLCS Solution Tank restores the ability of the SLCS to maintain 
the Boron concentration required to ensure cold shutdown for future 
anticipated core configurations. The minimum concentration of 
natural Boron in the reactor vessel required to achieve cold 
shutdown for the currently anticipated future core configurations is 
660 ppm. The 660 ppm concentration was recommended by General 
Electric in Service Information Letter (SIL) No. 325, June 1980.
    The increase in the amount of Boron-10 required to be stored in 
the SLCS Solution Tank restores the ability of the SLCS to maintain 
the Boron concentration required to ensure cold shutdown for future 
anticipated core configurations. The increased Boron-10 storage 
requirements does not represent a change to the designed Boron 
concentration capability of the SLCS. Since this change will ensure 
the ability of the SLCS to mitigate the consequences of an accident 
for future anticipated core designs, the change does not involve a 
significant increase in the consequences of any accident previously 
evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The increase in the amount of Boron-10 required to be stored in 
the SLCS Solution Tank does not affect the function or operation of 
any other system. The proposed change does not introduce any new 
modes of operation or modify existing equipment design. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The SLCS shutdown margin is determined for each core 
configuration by using the BWR simulator code to calculate the core 
multiplication for the cold, xenon-free, all-rods-out condition at 
the exposure point of maximum cold reactivity. The resulting k-
effective is subtracted from the critical k-effective of 1.0 to 
obtain the SLCS shutdown margin. Increasing the amount of Boron-10 
stored in the SLCS Solution Tank increases the SLCS shutdown margin. 
Therefore, the proposed amendment results in an increased margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone County, 
Alabama

    Date of amendment request: September 30, 1993 (TS 312)
    Description of amendment request: The proposed amendment removes 
Technical Specification requirements for scram discharge volume air 
header pressure trip switches in the Reactor Protection System (RPS) 
for the Browns Ferry Nuclear Plant (BFN) Unit 2. The amendment also 
clarifies the description of the RPS scram discharge volume high water 
level bypass in the Technical Specifications for BFN Units 1, 2, and 3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The Unit 2 scram pilot air header pressure trip switches perform 
the same function as the high water level switches in the scram 
discharge instrument volume. They automatically initiate control rod 
insertion (scram) in the event that degraded air conditions are 
detected in the BWR control air supply system. Since the scram pilot 
air header pressure trip function is to ensure that the Control Rod 
Drive (CRD) System is available to mitigate the consequence of an 
accident or transient, and the removal of the scram pilot air header 
pressure e trip switches scram function does not effect the 
precursors for any accident or transient analyzed in Chapter 14 of 
the BFN Final Safety Analysis Report (FSAR), there is no increase in 
the probability of any accident previously evaluated.
    The scram system has been analyzed, based on a plant-specific 
maximum inleakage, and the removal of the scram pilot air header 
pressure trip switches scram function would still result in a 
successful scram, provided that the CRD leakage rate and the water 
level instrumentation response characteristics remain within the 
success criteria region. Administrative controls and periodic 
measurements of the CRD and scram discharge system performance 
ensure acceptable response time delays and stall flow rates. Since 
the scram function would be successfully performed, the removal of 
the scram pilot air header pressure trip switches scram function 
does not involve a significant increase in the consequences of any 
accident previously evaluated.
    The clarification of the description of the scram discharge 
volume high water level bypass in the Reactor Protection System does 
not reflect a modification to plant equipment, maintenance 
activities, or operating instructions. The revised description does 
not effect the precursors for any accident or transient analyzed in 
Chapter 14 of the BFN FSAR or equipment used in the mitigation of 
these accidents or transients. Therefore, there is no increase in 
the probability of any accident previously evaluated nor an increase 
in the consequences of any accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The design criteria for the scram system is contained in the 
generic Safety Evaluation Report (SER), which was transmitted by NRC 
letter to All BWR Licensees, dated December 9, 1980, BWR Scram 
Discharge System. Section 4.2.4.1 of the generic SER states that the 
CRD system shall be analyzed based on a plant-specific maximum 
inleakage to ensure that the system function is not lost prior to 
initiation of automatic scram. This analysis has been performed and 
it was concluded that the removal of the scram pilot air header 
pressure trip switches scram function would still allow a successful 
scram, provided that the CRD leakage rate and the water level 
instrumentation response characteristics remain within the success 
criteria region. Administrative controls and periodic measurements 
of the CRD and scram discharge system performance ensure acceptable 
response time delays and stall flow rates.
    The overall scram system design is in conformance with the 
generic SER. No new system failure modes are created as a result of 
removing the scram pilot air header pressure trip switches scram 
function since, assuming the maximum CRD inleakage due to a degraded 
air supply system, redundant and diverse level switches will 
initiate a successful scram prior to this function becoming 
disabled. The removal of the scram pilot air header pressure trip 
switches scram function does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The clarification of the description of the scram discharge 
volume high water level bypass in the Reactor Protection System does 
not reflect a modification to plant equipment, maintenance 
activities, or operating instructions. No new external threats, 
system interactions, release pathways, or equipment failure modes 
are created. Therefore, the clarification of this description does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    The water level in the scram discharge instrument volume is 
monitored by redundant Magnetrol and heated reference resistive 
temperature devices (RTD) level switches. Since the level switches 
are redundant and diverse, redundancy and diversity in the 
instrumentation that initiates the scram signal is maintained even 
with the removal of the scram pilot air header pressure trip 
switches scram function.
    The scram system has been analyzed, based on a plant-specific 
maximum inleakage, and the removal of the scram pilot air header 
pressure trip switches scram function would still result in a 
successful scram, provided that the CRD leakage rate and the water 
level instrumentation response characteristics remain within the 
success criteria region. Since the scram system would successfully 
operate to mitigate the consequences of accidents and transients 
previously analyzed, the proposed amendment does not involve a 
significant reduction in the margin of safety.
    The clarification of the description of the scram discharge 
volume high water level bypass in the Reactor Protection System does 
not reflect a modification to plant equipment, maintenance 
activities, or operating instructions. There is no change to the 
licensing or design basis of the Reactor Protection System. 
Therefore, the revised description does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority (TVA), Docket Nos. 50-259, 50-260 and 
50-296, Browns Ferry Nuclear Plant, Units 1, 2 and 3, Limestone 
County, Alabama

    Date of amendment request: April 4, 1994 (TS322, Revision 1)
    Description of amendment request: The proposed amendment supersedes 
an amendment request dated March 25, 1993. Similar to the March 25, 
1993, proposal, the revised amendment request for the Browns Ferry 
Nuclear Plant (BFN), Units 1, 2, and 3, continues to seek elimination 
of Technical Specifications (TS) for the reactor scram and main steam 
line isolation functions associated with the main steam line radiation 
monitors (MSLRMs). Additionally the revised amendment request proposes 
elimination of TS for the remaining isolation functions associated with 
the MSLRMs as well. The remaining isolation functions include main 
steam line drain valve closure, reactor recirculation loop sample valve 
closure, and main condenser mechanical vacuum pump (MVP) trip and 
isolation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The probability of occurrence of previously evaluated accidents 
is based on initial conditions and assumptions which are not 
dependent directly or indirectly on the functions of the Main Steam 
Line Radiation Monitors (MSLRMs). Elimination of the MSLRM scram and 
isolation functions does not affect the operation of the other 
Reactor Protection System or Primary Containment Isolation System 
functions required to mitigate a Rod Drop Accident (RDA). Also, 
eliminating the MSLRM functions does not affect the Control Rod 
Drive System and, hence, cannot increase the probability of RDA. The 
proposed change does not involve any increase in the probability of 
an [any] previously evaluated accident.
    There is no significant increase in the consequences of any 
previously evaluated accident. Elimination of the MSLRM reactor 
scram and isolation functions could potentially increase the amount 
of radioactivity released following an RDA. However, the potential 
increase in release from those previously determined is not 
significant. Both the NEDO-31400 analysis and the additional BFN 
dose calculations determined that the resulting doses (See Table B 
[of this request]) remain well below the limits of Standard Review 
Plan (SRP) 15.4.9 and 10 CFR 100.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This proposed change deletes the reactor scram and isolation 
functions of the MSLRMs. The sole purpose of these functions is to 
assist in mitigating the consequences of an RDA, a previously 
analyzed event. This event is terminated by a high flux scram.
    The NEDO-31400 RDA analysis without the MSLRM scram and MSIV 
[Main Steam Isolation Valve] closure functions has been reviewed and 
accepted by the NRC. An RDA without the MSLRM MVP isolation function 
has been previously reviewed by the NRC as documented in Section 
14.6.2 of the Updated Final Safety Analysis Report (UFSAR). An RDA 
without the MSLRM recirculation sample line isolation coupled with a 
sample line break is no different than that if the normal RWCU 
[Reactor Water Cleanup] [system] sample path is in service (see 
Section III.C) [of this request]. However, TVA has evaluated the 
consequences of this event and determined the MSLRM sample line 
isolation function is not required to maintain radioactive releases 
within the acceptable limits.
    The NEDO-31400 RDA analysis and the additional BFN offsite dose 
calculations show that the elimination of the MSLRM functions does 
not create the possibility of a new or different kind of
    accident from any accident previously evaluated.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    A reliability assessment of the elimination of the MSLRM scram 
function on reactivity control failure frequency and core damage 
frequency was performed as part of the NEDO-31400 analysis. The 
results of the NEDO-31400 analysis indicated a negligible increase 
in reactivity control failure frequency associated with deletion of 
the MSLRM scram function. However, this increase is offset by a 
reduction in the frequency of transient initiating events 
(inadvertent scrams). This reduction in transient initiating events 
represents a net reduction in core damage frequency of 0.3 percent.
    BFN TS Bases 3.2 states that the MSLRMs are provided to detect 
gross fuel failures as in an RDA and provide main steam isolation 
valve closure to maintain radiological releases below the 10 CFR 100 
limits. The BFN offsite dose calculations and the NRC's Safety 
Evaluation Report approving the NEDO-31400 RDA analysis document 
that the potential radiological release consequences following an 
RDA without the MSLRMs reactor scram and isolation functions are 
still well within the 10 CFR 100 limits. Thus, this change will not 
result in a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: May 16, 1994 (TS 93-18)
    Description of amendment request: The proposed change would revise 
Surveillance Requirement 4.8.1.1.1.b to reflect installation of common 
station service transformers with auto load tap changers and the 
resulting change to the alignment of this portion of the distribution 
system. The proposed amendment would replace the terms ``normal'' and 
``alternate'' with ``unit generator supported'' and ``preferred power 
(GDC 17),'' respectively, circuits that must be demonstrated operable.
    Basis for proposed no significant hazards consideration 
determination: Asrequired by 10 CFR 50.91(a), the licensee has provided 
its analysis of the issue of no significant hazards consideration, 
which is presented below:

    TVA has evaluated the proposed technical specification (TS) 
changeand has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The description clarification for the circuits in TS 4.8.1.1.1.b 
does not change any TS requirement and will not alter plant 
functions or components as a result of normal alignment to the 
common station service transformers (CSSTs). By implementing normal 
CSST alignment to unit power, SQN has reduced the required transfers 
to obtain alignment to the preferred power circuit under accident 
conditions. The verifications provided by TS Surveillance 
4.8.1.1.1.b will continue to ensure the capability of this transfer 
during occasional use of the unit station service transformer (USST) 
for maintenance or repair activities. The accidents described in 
Chapter 15 of the SQN Updated Final Safety Analysis Report assume 
the unit power supply to be from the USST circuit that requires 
transfer to the CSST after turbine trip. This analysis is bounding 
for normal alignment to the CSST that does not require a transfer. 
These changes minimize the impact of malfunctions because transfers 
will not be required as often. The consequences of an accident are 
therefore not increased by this change. The proposed unit power 
alignment to the CSSTs has always been used for unit start-up; and 
since no plant functions are changed as a result of normal alignment 
to the CSSTs, other than a lower potential for transfers, the 
probability of an accident is not increased.

    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The TS 4.8.1.1.1.b clarification does not change plant functions 
and only provides a more appropriate description of the circuits 
involved in the transfer function. The transfer from the USST 
circuit to the CSST circuit will still be available when SQN is 
using the USST circuit, but the need for these transfers will be 
reduced by normal alignment of unit power to the preferred CSST 
circuit. These changes will not create a new or different kind of 
accident because the plant functions remain unchanged and only the 
CSST alignment for normal unit power and circuit descriptions is 
altered. This CSST alignment has always been used for unit start-up 
and is being expanded to apply to normal power operation in addition 
to start-up and accident conditions.
    3. Involve a significant reduction in a margin of safety.
    The clarification for the unit power circuit descriptions does 
not alter the intent of requirements for this surveillance. Plant 
functions and setpoints remain unchanged for automatic transfers 
from the USST circuit to the CSST circuit. This circuit description 
change provides a clearer explanation of how the transfer functions. 
The margin of safety has not been altered for unit power alignment 
to the USST or CSST, and this change will support SQN's use of the 
CSST for the normal unit power circuit. Therefore, these changes do 
not involve a significant reduction in a margin of safety.

    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: December 6, 1993
    Description of amendment request: The proposed amendment would 
remove the requirement to perform jet pump integrity and operability 
surveillances in the idle loop during operation with one recirculation 
pump. It also provides consistency between the Bases and Section 4.6.F 
on jet pump surveillance requirements, and includes some administrative 
changes within the Bases.
    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:

    Removing the surveillance requirement to examine the jet pump 
flow patterns within the idle loop during single loop operation does 
not increase the probability or consequences of an accident 
previously evaluated. The probability of loss of the ability of such 
a jet pump to maintain a refloodable volume of two thirds core 
height is not increased. The stresses undergone without forced 
recirculation flow are very small and therefore a jet pump within an 
idle recirc loop should not physically degrade during this time. The 
ability of the core standby cooling systems to provide their 
intended function is not degraded by removal of this surveillance 
requirement; therefore the consequences of the recirculation line 
break accident are not increased.
    This change will remove a surveillance requirement that has been 
determined to be unnecessary based upon the operating conditions 
during single recirculation loop operation. Other surveillance 
requirements within the technical specifications adequately 
determine jet pump operability when stresses exist which could cause 
a jet pump failure to occur. The change will not affect any plant 
hardware, plant design, safety limit settings, or plant system 
operation, and therefore does not modify or add any initiating 
parameters that would significantly increase the probability or 
consequences of any previously analyzed accident.
    As discussed above, there is no increase in the probability of a 
jet pump failing and being undetectable with the proposed technical 
specification surveillances. Therefore, there is no increase in the 
possibility of not being able to maintain a floodable volume of two 
thirds core height during the design basis LOCA [loss-of-coolant 
accident]. As a result, there is no 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624
    NRC Project Director: Walter R. Butler

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.

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 application for amendment: January 14, 1994
    Brief description of amendment request: The proposed amendment 
would revise the technical specifications (TSs) by removing component 
lists from the TSs in accordance with NRC Generic Letter (GL) 91-08 and 
by removing the schedule for withdrawal of reactor vessel material 
specimen capsules from the TSs in accordance with GL 91-01.
    Date of individual notice in Federal Register: May 23, 1994 (59 FR 
26675)
    Expiration date of individual notice: June 22, 1994
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

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

    Date of application for amendment: December 6, 1993, supplemented 
by letter May 6, 1994
    Brief description of amendment request: The proposed amendment 
would revise Technical Specifications (TS) 3/4.4.2 and 3/4.5.1 to 
require main steam system and automatic depressurization system safety/
relief valve (SRV) surveillance testing within 12 hours after steam 
pressure and flow have been found to be adequate to do the testing. 
Also, TS Table 4.3.7.5-1 would be revised to require SRV position 
indicator surveillance testing within 12 hours after steam pressure and 
flow have been found to be adequate to do the testing.
    Date of individual notice in Federal Register: May 13, 1994 (59 FR 
25131)
    Expiration date of individual notice: June 13, 1994
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendment: May 5, 1994
    Brief description of amendment request: The proposed amendment 
would change the plant operating license to rename three primary 
containment isolation check valves listed in the technical 
specifications. The licensee is making an administrative change to 
rename valve PI-EFC-X29d to make its number consistent with other 
similar valves in the technical specifications. The licensee is 
renaming excess flow check valves PI-EFCX-72f and PI-EFCX-73e because 
they are replacing them with swing check valves that have a different 
numbering nomenclature.
    Date of individual notice in Federal Register: May 12, 1994 (59 FR 
24762)
    Expiration date of individual notice: June 13, 1994
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Notice Of Issuance Of Amendments To Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendments: January 20, 1994
    Brief description of amendments: The amendments increase the 
departure from nucleate boiling ratio limit from 1.24 to 1.30 to 
accommodate the uncertainties in core inlet flow distribution. The 
amendments also add the analytical method supplement entitled ``System 
80TM Inlet Flow Distribution'' to the list of methods used to 
determine core operating limits.
    Date of issuance: May 26, 1994
    Effective date: May 26, 1994
    Amendment Nos.: 76, 62 and 48
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12358) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 26, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004

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 application for amendments: April 1, 1993
    Brief description of amendments: The amendments make 
administrative, editorial, and format changes to the Operating 
Licenses. These changes include the deletion or incorporation, as 
appropriate, of all handwritten or ``pasted-up'' changes and the 
removal of all previous license conditions that have been completed to 
the satisfaction of the Commission. The changes and reformatting result 
in the Operating Licenses containing only those license conditions that 
are currently applicable.
    Date of issuance: May 20, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 189 and 166
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28052) The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated May 20, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: February 9, 1994
    Brief description of amendment: This amendment changes the 
Technical Specifications and the License. These changes consist of 
revised wording for the License, clarifying wording to aid operators in 
selecting the correct pressure/temperature curve during startup and 
shutdown operations, and removal of certain obsolete mechanical snubber 
acceptance criteria.
    Date of issuance: May 16, 1994
    Effective date: May 16, 1994
    Amendment No.: 153
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12359) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 16, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

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: March 1, 1993
    Brief description of amendment: The amendment revises the Technical 
Specifications to provide specific actions to take if primary 
containment leakage limits are exceeded and cannot be restored when 
reactor coolant temperature is greater than 200 deg. F.
    Date of issuance: May 26, 1994
    Effective date: May 26, 1994
    Amendment No. 60
    Facility Operating License No. NPF-58. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14896) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 26, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: January 17, 1994
    Brief description of amendment: The amendment removes Technical 
Specification 3/4.4.12, ``Failed Fuel Rods'' and its associated Bases 
Section 3/4.4.12.
    Date of Issuance: May 17, 1994Effective date: As of the date of 
issuance to be implemented within 30 days.
    Amendment No.:  171
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 16, 1994 (59 
FR 7687) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated May 17, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, Connecticut 06457.

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

    Date of amendment request: March 19, 1993
    Brief description of amendment: The amendment changed Technical 
Specification 4.18.6 and Table 4.18-2 to make the requirements for C-3 
reports on steam generator tube inspections consistent with the Babcock 
& Wilcox Standard Technical Specifications.Date of issuance: May 19, 
1994Effective date: May 19, 1994
    Amendment No.:  172
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 21, 1993 (58 FR 
39049) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 19, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801

Florida Power and Light Company, Docket Nos. 50-250 and 50-251, 
Turkey Point Plant Units 3 and 4, Dade County, Florida

    Date of application for amendments: September 3, 1993
    Brief description of amendments: These amendments eliminate the 
reliance on the five ``cranking'' diesel generators by replacing the 
motor driver on one of the standby steam generator feedwater pumps with 
a dedicated diesel driver.
    Date of issuance: May 20, 1994
    Effective date: May 20, 1994
    Amendment Nos. 164 and 158Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52985) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 20, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199

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

    Date of application for amendment: November 30, 1993
    Brief description of amendment: The amendment revises the plant 
Technical Specifications (TS) to remove the protective and maximum 
allowable setpoint limits for axial power imbalance and the trip 
setpoint for nuclear overpower based on reactor coolant system (RCS) 
flow (flux-to-flow) from the TS and relocate them to the existing TMI-1 
Core Operating Limits Report (COLR).
    Date of Issuance: May 23, 1994
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 184
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2867). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 23, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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

    Date of amendment request: March 3, 1994
    Brief description of amendment: The amendment revises the technical 
specifications in accordance with the guidance provided by Generic 
Letter 93-08, ``Relocation of Technical Specification Tables of 
Instrument Response Time Limits.''
    Date of issuance: May 19, 1994
    Effective date: May 19, 1994
    Amendment No.: 73
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12380) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 19, 1994.No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803

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

    Date of application for amendment: January 21, 1994
    Brief description of amendment: The amendment revised the Technical 
Specifications by changing the name of one operator and one of the 
owners of the Duane Arnold Energy Center from Iowa Electric Light and 
Power Company to IES Utilities Inc. The title of the position 
responsible for the management of the Nuclear Division has also been 
changed to Vice President, Nuclear from Manager-Nuclear Division. One 
blank page was deleted and several other editorial changes were made. 
Pages 1 through 4 of the Facility Operating License were revised to 
reflect the corporate name change and a spelling error.
    Date of issuance: May 13, 1994
    Effective date: May 13, 1994
    Amendment No.: 198
    Facility Operating License No. DPR-49. Amendment revised the 
license and Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10008) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 13, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: December 22, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications by correcting several typographical and administrative 
errors.
    Date of issuance: May 18, 1994
    Effective date: May 18, 1994
    Amendment No.:  199
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10008) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 18, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: April 14, 1994, as 
supplementedApril 20, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to change the laboratory testing protocol for the 
charcoal absorbers for the Control Room Emergency Ventilation System 
(TS 3.7.6.1), the Enclosure Building Filtration System (TS 3.6.5.1) and 
the Storage Pool Ventilation System (TS 3.9.15).
    Date of issuance: May 23, 1994
    Effective date: As of the date of issuance to be implemented 
within30 days.
    Amendment No.: 175
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (59 FR 23085, May 4, 1994) That 
notice provided an opportunity to submit comments on the Commission's 
proposed no significant hazards consideration determination. No 
comments have been received. The notice also provided for an 
opportunity to request a hearing by June 3, 1994, but indicated that if 
the Commission makes a final no significant hazards consideration 
determination any such hearing would take place after issuance of the 
amendment.The Commission's related evaluation of the amendment, finding 
of exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated May 
23, 1994.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, Connecticut 06360.

Northern States Power Company, Docket Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County,Minnesota

    Date of application for amendments: February 14, 1994
    Brief description of amendments: The amendments revise Technical 
Specifications to reflect the new configuration for the Unit 1 480V 
safeguards bus arrangement (two 480V safeguards buses fed by each 4160V 
safeguards bus). These changes make the specifications the same for 
both units since the configuration for the two units will become the 
same at the completion of the current outage.
    Date of issuance: May 17, 1994
    Effective date: May 17, 1994
    Amendment Nos.: 110 & 103
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14892) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 17, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment: January 6, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications (TS) to incorporate the following changes:
    (1) The residual heat removal (RHR) pump flow calibration frequency 
(specified in TS Table 4.1-1) was changed to accommodate operation on a 
24-month cycle.
    (2)The RHR loop isolation valve automatic isolation and interlock 
testing frequency (specified in TS Table 4.1-3) was changed to 
accommodate operation on a 24-month cycle.
    (3)The RHR system leakage testing frequency (specified in TS 
Section 4.4.I.4) was changed to accommodate operation on a 24-month 
cycle.
    (4)The recirculation pump testing frequency (specified in TS 
Section 4.5.B.1.a) was changed to accommodate operation on a 24-month 
cycle.
    (5)The accumulator check valve operability testing frequency 
(specified in TS Section 4.5.B.2.b) was changed to accommodate 
operation on a 24-month cycle.
    (6)The safety injection (SI)/RHR check valve gross leakage testing 
frequency (specified in TS Section 4.5.B.2.c) was changed to 
accommodate operation on a 24-month cycle.
    These changes followed the guidance provided in Generic Letter 91-
04, ``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle,'' as applicable.
    In addition, the gross leakage surveillance requirements for 
certain SI/RHR system check valves (specified in TS Section 4.5.B.2.d) 
was changed to implement requirements as set forth in NRC generic 
letter, dated February 23, 1980, regarding testing of low pressure 
injection (LPI)/RHR check valves. Therefore, Item A.5 of the February 
11, 1980, Confirmatory Order is considered rescinded.
    Date of issuance: May 20, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.:  148
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 17, 1993 (58 
FR 8777) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 20, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: October 18, 1993, and suplement 
dated March 7, 1994
    Brief description of amendment: This amendment revised TS 3/4.3.2 
and associated Bases to increase surveillance test intervals and add 
out-of-service times for isolation actuation instrumentation. The 
changes are in accordance with General Electric Company's Licensing 
Topical Reports which have been previously reviewed and approved by the 
NRC staff.
    Date of issuance: May 25, 1994
    Effective date: As of date of issuance and shall be implemented 
within 60 days of the date of issuance.
    Amendment No.: 70
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64615) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 25, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: October 26, 1993 and 
supplemented by letter dated December 14, 1993.
    Brief description of amendment: The amendment request revised the 
MINIMUM CHANNELS OPERABLE requirement for suppression pool water 
temperature instruments, Accident Monitoring ACTION STATEMENTS, and 
removes ACTIONS and Surveillance Requirements for suppression chamber 
temperature and level instruments.
    Date of issuance: May 25, 1994
    Effective date: May 25, 1994
    Amendment No.: 71
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67861) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 25, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of amendments request: October 14, 1993
    Brief description of amendments: The requested changes incorporate 
changes allowing longer surveillance test intervals (STIs) and allowed 
outage times (AOTs) for the reactor trip system (RTS) and engineered 
safety features actuation system (ESFAS) instrumentation into the 
Technical Specifications. The proposed changes also revise certain RTS/
ESFAS functions, minimum channels operable, channel calibration, and 
channel functional test requirements to ensure they are in concert with 
the Westinghouse Standard Technical Specifications and WCAP-10271, 
``Evaluation of Surveillance Frequencies and Out-of-Service Times for 
Reactor Protection Instrumentation Systems.''
    Date of issuance: May 16, 1994
    Effective date: May 16, 1994
    Amendment Nos.: 107 and 99
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (57 FR 
17605) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 16, 1994.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Tennessee Valley Authority, Docket Nos. 50-259 and 50-296, Browns 
Ferry Nuclear Plant, Units 1 and 3, Limestone County, Alabama

    Dates of application for amendments: April 6, 1992 (TS 308), and 
September 28, 1992 (TS 326)
    Brief description of amendments: The amendments change the 
Technical Specifications (TS) to add Automatic Depressurization System 
(ADS) high drywell pressure bypass timer requirements, revise the ADS 
timer trip level setting, increase the number of ADS valves required to 
be operable for startup, and revise the limiting conditions for 
operation with inoperable ADS valves. The ADS bases have also been 
revised for consistency with these TS changes.
    Date of issuance: May 19, 1994
    Effective date: May 19, 1994
    Amendment Nos.: 205 and 178
    Facility Operating License Nos. DPR-33 and DPR-68: Amendments 
revised the Technical Specifications.
    Dates of initial notice in Federal Register: May 27, 1992 (57 FR 
22269) and November 25, 1992 (57 FR 55593).The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 19, 1994.No significant hazards consideration comments received: 
None
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611
    Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
    Date of application for amendments: February 9, 1994; supplemented 
April 13, 1994 (TS 93-21)
    Brief description of amendments: The amendments change Technical 
Specification Table 3.3-11 to reflect the addition of ionization fire 
detectors to Fire Zones 184, 185, 186, and 187.
    Date of issuance: May 23, 1994
    Effective date: May 23, 1994
    Amendment Nos.:  181 and 173
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: March 16, 1994 (59 FR 
12369) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated May 23, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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

    Date of application for amendments: October 1, 1993; supplemented 
March 29, 1994 (TS 93-09)
    Brief description of amendments: The amendments revise the 
setpoints and time delays for the Auxiliary Feedwater loss of power and 
6.9 kv shutdown board loss-of-voltage and degraded-voltage 
instrumentation. In addition, the description, total number of 
channels, channels to trip, minimum channels operable, actions, trip 
setpoints, allowable values, channel checks, and channel functional 
test requirements for loss-of-power instrumentation have been revised.
    Date of issuance: May 24, 1994
    Effective date: May 24, 1994
    Amendment Nos.: 182 and 174
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: February 2, 1994 (59 FR 
4947) The Commission's related evaluation of the amendments are 
contained in a Safety Evaluation dated May 24, 1994.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

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: March 30, 1994
    Brief description of amendments: The proposed amendments revise the 
technical specifications by adding an alternative method for verifying 
that the emergency diesel generator fuel oil meets requirements.
    Date of issuance: May 13, 1994
    Effective date: May 13, 1994, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 24; Unit 2 - Amendment No. 
10
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17606) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 13, 1994 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P.O. 
Box 19497, Arlington, Texas 76019.

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

    Date of application for amendments: February 25, 1994
    Brief description of amendments: The amendments revise the 
surveillance frequency of the nozzles in the Quench Spray and 
Recirculation Spray Systems from 5 to 10 years. The change is in 
accordance with NRC Generic Letter 93-05, ``Line-Item Technical 
Specifications Improvements to Reduce Surveillance Requirements at 
Power Operation,'' dated September 27, 1993.
    Date of issuance: May 16, 1994
    Effective date: May 16, 1994
    Amendment Nos.: 182 and 163
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17608) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 16, 1994No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendments: October 4, 1993
    Brief description of amendments: These amendments revise the NA-1&2 
Technical Specifications requirements to allow the use of ZIRLO 
material for fuel cladding.
    Date of issuance: May 26, 1994
    Effective date: May 26, 1994
    Amendment Nos.: 183 and 164
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57859) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 26, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: February 25, 1994
    Brief description of amendments: The amendments revise the 
surveillance frequency of the nozzles in the containment spray and 
recirculation spray systems from 5 to 10 years.
    Date of issuance: May 20, 1994
    Effective date: May 20, 1994
    Amendment Nos. 191 and 191
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 13, 1994 (59 FR 
17608) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 20, 1991.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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 20555, 
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 July 8, 1994, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC 20555 and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. 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 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Commonwealth Edison Company, Docket No. STN 50-457, Braidwood 
Station, Unit No. 2, Will County, Illinois

    Date of application for amendment: April 21, 1994
    Brief description of amendment: The amendment effects a one-time 
only change to Technical Specification (TS) Surveillance Requirement 
4.7.1.1 by adding a note which relieves Braidwood, Unit 2, from 
compliance with the provisions of TS 4.0.4 until initial entry into 
Mode 2. This will permit Braidwood, Unit 2, to reach Mode 3 to reset 
Main Steam Safety Valves (MSSVs) and proceed with a startup. The 
amendment is applicable only until entry into Mode 2 following forced 
outage A2F27.
    Date of issuance: May 16, 1994
    Effective date: May 16, 1994
    Amendment No.: 51
    Facility Operating License No. NPF-77. This amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
siginificant hazards consideration: Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration and requested that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by the close of business on May 12, 1994. The notice was 
published in the Joliet News Herald and the Morris Daily Herald on May 
9, 1994. No comments have been received.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, consultation with the state of Illinois and 
final no significant hazards consideration determination are contained 
in a Safety Evaluation dated May 16, 1994.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481
    NRC Project Director: James E. Dyer
    Dated at Rockville, Maryland, this 1st day of June 1994.
    For the Nuclear Regulatory Commission.
Gus C. Lainas,
Acting Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation.
[FR Doc. 94-13765 Filed 6-7-94; 8:45 am]
Billing Code 7590-01-F