[Federal Register Volume 63, Number 67 (Wednesday, April 8, 1998)]
[Notices]
[Pages 17219-17242]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-9040]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 16, 1998, through March 27, 1998. The 
last biweekly notice was published on March 25, 1998.

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 Chief, Rules and 
Directives Branch, Division of Administration Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By May 8, 1998, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the

[[Page 17220]]

following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: February 11, 1998.
    Description of amendment request: The proposed amendment would 
modify the Pilgrim Nuclear Power Station (PNPS) Updated Final Safety 
Analysis Report (UFSAR) Section 10.7, Salt Service Water System, by 
identifying that certain single active failures do exist that could 
leave the Salt Service Water (SSW) system in a configuration with one 
SSW pump serving both SSW trains through open crossover (division) 
valves for the first 10 minutes of an accident.
    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.
    Operation with one (1) SSW pump supplying two (2) SSW trains is not 
an accident or transient precursor and does not prevent the [Reactor 
Building Closed Cooling Water] RBCCW system from providing adequate 
cooling during an accident. Core cooling requires no SSW for the first 
ten minutes, and no containment cooling is assumed for the first ten 
minutes. Pump testing has proved no SSW pump damage will result from 
this configuration so there will be no effect on the containment 
cooling function. The current licensing basis includes operator action 
after ten minutes to align the SSW system to achieve containment 
cooling. This amendment does not affect operator action after ten 
minutes since pump and valve manipulations are already required to 
align containment cooling. Therefore, the changes do not involve a 
significant 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.
    The SSW system operating modes are not accident precursors. They 
cannot influence the types of accidents that can occur. The SSW pumps 
can withstand operation under the full range of conditions and for the 
time periods considered under a one pump, two train system 
configuration with no adverse effects. The SSW system is properly 
designed as a common header arrangement with five (5) pumps in which 
any combination of one to five pumps may operate without damaging 
effects.
    3. The proposed amendment does not involve a significant reduction 
in the margin of safety.
    Operation with one (1) SSW pump supplying two (2) SSW trains does 
not impact the ability to provide adequate core or containment cooling 
during an accident. Although SSW system flow will be diminished during 
the first ten minutes of the accident, no system flow at all is needed 
at that time. The current licensing basis credits operator action after 
ten minutes to align the [Residual Heat Removal] RHR, RBCCW, and SSW 
systems for containment cooling.

[[Page 17221]]

Operators are expected to isolate the SSW loops or start additional SSW 
pumps as necessary given the existing specific conditions.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: February 20, 1998.
    Description of amendment request: The proposed amendment would 
change the Pilgrim Nuclear Power Station Technical Specification (TS) 
3/4.5.B and its Bases to incorporate the ultimate heat sink (UHS) 
temperature of 75 deg.F, as required by Amendment No. 173. The 
introduction of a UHS temperature restriction requires new 
specifications, actions, and surveillances for the salt service water 
system.
    The amendment would also replace existing Specification 3.5.B 
``Containment Cooling System'' with new Specification 3/4.5.B.1 
``Residual Heat Removal (RHR) Suppression Pool Cooling,'' 3/4.5.B.2 
``Residual Heat Removal (RHR) Containment Spray,'' 3/4.5.B.3 ``Reactor 
Building Closed Cooling Water (RBCCW) System,'' and 3/4.5.B.4 ``Salt 
Service Water (SSW) System and Ultimate Heat Sink (UHS).'' The proposed 
new subsections will more clearly define the various subsystems that 
comprise the containment cooling system and the operating states in 
which they are applicable. The proposed changes also provide clarity 
with respect to the application of limiting conditions of operation 
(LCOs), actions, completion times, and surveillances for the 
containment cooling subsystems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Operation of PNPS in accordance with the proposed change will not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated because of the following:
Administrative Changes
    These proposed changes (editorial rewording, reformatting, 
repagination, and renumbering) are made to restructure the section, 
accounting for the new specifications replacing Specification 3/4.5.B. 
These proposed administrative changes do not alter any existing 
requirements.
Technical Changes--More Restrictive
    The proposed changes provide more stringent requirements than 
previously existed in the Technical Specifications. The more stringent 
requirements provide greater assurance that the affected systems will 
remain capable of providing the safety functions assumed in design 
basis accidents and transients. If anything, the new requirements may 
decrease the probability or consequences of an analyzed event. The 
change will not alter assumptions relative to mitigation of an accident 
or transient event. The more restrictive requirements will not alter 
the operation of process variables, structures, systems, or components 
as described in the safety analyses.
Technical Changes--Relocations
    This proposed change relocates requirements from the Technical 
Specifications to the Inservice Testing (IST) Program. The (IST) 
Program documents containing the relocated requirements must be 
maintained using the provisions of 10 CFR 50.55a and 10 CFR 50.59. 
Since any changes to the (IST) Program documents will be evaluated per 
10 CFR 50.55a and 10 CFR 50.59, no increase in the probability or 
consequences of an accident previously evaluated will be allowed 
without NRC review.
Technical Changes--Less Restrictive
    This change relaxes the current requirements to declare the 
affected RBCCW subsystem inoperable when one of the required RBCCW 
pumps is inoperable. Since the RBCCW system is not assumed as an 
initiator of any analyzed event, the proposed change will not affect 
the probability of an accident occurring. The safety function of the 
RBCCW system is to support the operability of the RHR suppression pool 
cooling and spray functions, and component cooling for the RHR and core 
spray pumps, and area coolers. With one required RBCCW pump inoperable, 
the remaining pump in the affected subsystem is capable of supporting 
the component cooling requirements for the RHR and core spray pumps, 
and area coolers, and the remaining OPERABLE subsystem is capable of 
supporting the suppression pool cooling and spray functions.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Operation of PNPS in accordance with the proposed change will not 
create the possibility of a new or different kind of accident from any 
accident previously evaluated because of the following:
Administrative Changes
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing plant operation. The proposed changes will 
not impose any new or different requirements or eliminate any existing 
requirements.
Technical Changes--More Restrictive
    The proposed more restrictive requirements will not alter the plant 
configuration (no new or different type of equipment will be installed) 
or change methods governing plant operation. The change does impose 
different requirements. However, the changes are consistent with 
assumptions made in the safety analyses.
Technical Changes--Relocations
    This change relocates requirements to the (IST) Program. This 
change will not alter the plant configuration (no new or different type 
of equipment will be installed) or changes in methods governing plant 
operation. This change will not impose different requirements, and 
adequate control of information will be maintained. This change will 
not alter assumptions made in the safety analysis.
Technical Changes--Less Restrictive
    The proposed change will not involve any physical changes to plant 
systems, structures, or components (SSC), or the manner in which these 
systems are operated, maintained, modified, tested, or inspected.
    3. Does this change involve a significant reduction in a margin of 
safety?

[[Page 17222]]

Administrative Changes
    Operation of PNPS in accordance with the proposed change will not 
involve a significant reduction in a margin of safety because of the 
following: safety analysis margin of safety.
    The changes are administrative in nature and do not involve any 
technical changes. Since no technical changes (either actual or 
interpretational) were made, there is no impact on any safety analysis 
margin of safety.
Technical Changes--More Restrictive
    The proposed more restrictive requirements will not alter 
assumptions relative to mitigation of an accident or transient event or 
alter the operation of process variables, structures, systems, or 
components as described in the safety analyses.
Technical Changes--Relocations
    This change relocates requirements from the Technical 
Specifications to the Inservice Testing (IST) Program. The requirements 
to be transposed to the IST program are the same as the existing 
Technical Specifications. Since any changes to the (IST) Program 
documents will be evaluated per 10 CFR 50.55a and 10 CFR 50.59, no 
reduction in margin of safety previously approved will be allowed 
without NRC review.
Technical Changes--Less Restrictive
    The 7 day completion time is consistent with the completion times 
for one inoperable loop of suppression pool cooling system or 
containment spray system, and the remaining pump in the affected 
subsystem is capable of supporting the component cooling requirements 
for the RHR and core spray pumps, and area coolers.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W.S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, Massachusetts 02199.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: March 12, 1998.
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 3/4.9.12, ``Fuel Handling Building 
Emergency Exhaust System.'' Specifically, Harris Nuclear Plant (HNP) 
proposes to delete Surveillance Requirement 4.9.12.d.4, which requires 
verifying that the filter cooling bypass valve for the Fuel Handling 
Building Emergency Exhaust System is locked in the balanced position at 
least once per 18 months.
    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.
    Fuel Handling Building Emergency Exhaust System (FHBEES) is not an 
accident initiating system as described in the Final Safety Analysis 
Report. The proposed change allows the elimination of the filter 
cooling bypass flowpath for FHBEES units. Engineering calculations were 
performed which demonstrate this filter cooling path is not required to 
mitigate the consequences of a fuel handling accident.
    Therefore, the proposed change does not involve a significant 
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.
    FHBEES is a ventilation system designed to limit off-site dose 
releases in the event of a fuel handling accident. FHBEES is not an 
accident initiating system as described in the Final Safety Analysis 
Report [FSAR]. The proposed change ensures the seismic and safety 
classification is maintained while not affecting another Structure, 
System, or Component.
    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 proposed change to FHBEES does not affect any of the parameters 
that relate to the margin of safety as described in the Bases of the TS 
or the FSAR. Accordingly, NRC Acceptance Limits are not affected by 
this change.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: Pao Tsin Kuo, Acting Director.

Consumers Energy Company, Docket No. 50-255, Palisades Plant, Van Buren 
County, Michigan

    Date of amendment request: September 3, 1997, as supplemented March 
13, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to delete snubber operability 
requirements (Change A), action requirements for inoperable snubbers 
(Change B), and snubber testing requirements (Change E). The snubber 
testing requirements would be relocated to the Palisades Operating 
Requirements Manual (ORM). Each proposed change has been classified by 
the licensee as either Administrative, More Restrictive, or Less 
Restrictive.
    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:
    Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    1. Administrative Change (Change A):
    ``Administrative'' changes make wording changes which clarify 
existing TS requirements, without affecting their technical content. 
Since ``Administrative'' changes do not alter the technical content of 
any requirements, they cannot involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 17223]]

    2. More Restrictive Change (Change B):
    ``More Restrictive'' changes only add new requirements, or revise 
existing requirements to result in additional operational restrictions. 
The TS, with all ``More Restrictive'' changes incorporated, will still 
contain all of the requirements which existed prior to the changes. 
Therefore, ``More Restrictive'' changes cannot involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    3. Less Restrictive Change (Change E):
    Change E deletes the TS requirements for snubber testing, but adds 
identical requirements to a document (the ORM) controlled under 10 CFR 
50.59.
    10 CFR 50.59 specifically prohibits changes to the facility as 
described in the safety analysis report, and to procedures described in 
the safety analysis report (without prior NRC approval) ``if the 
probability of occurrence or the consequences of an accident or 
malfunction of equipment important to safety previously evaluated in 
the safety analysis report may be increased''. Since the conditions 
which limit changes performed under 50.59 are more restrictive than the 
conditions which define changes considered to involve a significant 
hazards consideration, moving of a requirement from the TS to a 
document which is controlled under 50.59 cannot involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    1. Administrative Change (Change A):
    ``Administrative'' changes make wording changes which clarify 
existing TS requirements, without affecting their technical content. 
Since ``Administrative'' changes do not alter the technical content of 
any requirements, they cannot create the possibility of a new or 
different kind of accident from any previously evaluated.
    2. More Restrictive Change (Change B):
    ``More Restrictive'' changes only add new requirements, or revise 
existing requirements to result in additional operational restrictions. 
The TS, with all ``More Restrictive'' changes incorporated, will still 
contain all of the requirements which existed prior to the changes. 
Therefore, ``More Restrictive'' changes cannot create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. Less Restrictive Change (Change E):
    Change E deletes the TS requirements for snubber testing, but adds 
identical requirements to a document (the ORM) controlled under 10 CFR 
50.59.
    10 CFR 50.59 specifically prohibits changes to the facility as 
described in the safety analysis report, and to procedures described in 
the safety analysis report (without prior NRC approval) ``if a 
possibility for an accident or malfunction of a different type than any 
evaluated previously in the safety analysis report may be created''. 
Since the conditions which limit changes performed under 50.59 are more 
restrictive than the conditions which define changes considered to 
involve a significant hazards consideration, relocation of a 
requirement from the TS to a document which is controlled under 50.59 
cannot create the possibility of a new or different kind of accident 
from any previously evaluated.
    Do the proposed changes involve a significant reduction in a margin 
of safety?
    1. Administrative Change (Changes A):
    ``Administrative'' changes make wording changes which clarify 
existing TS requirements, without affecting their technical content. 
Since ``Administrative'' changes do not alter the technical content of 
any requirements, they cannot involve a significant reduction in a 
margin of safety.
    2. More Restrictive Change (Change B):
    ``More Restrictive'' changes only add new requirements, or revise 
existing requirements to result in additional operational restrictions. 
The TS, with all ``More Restrictive'' changes incorporated, will still 
contain all of the requirements which existed prior to the changes. 
Therefore, ``More Restrictive'' changes cannot involve a significant 
reduction in a margin of safety.
    3. Less Restrictive Change (Change E):
    Change E deletes the TS requirements for snubber testing, but adds 
identical requirements to a document (the ORM) controlled under 10 CFR 
50.59.
    10 CFR 50.59 specifically prohibits changes to the facility as 
described in the safety analysis report, and to procedures described in 
the safety analysis report (without prior NRC approval) ``if the margin 
of safety as defined in the basis for any technical specification is 
reduced''. Since the conditions which limit changes performed under 
50.59 are more restrictive than the conditions which define changes 
considered to involve a significant hazards consideration, relocation 
of a requirement from the TS to a document which is controlled under 
50.59 cannot involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Cynthia A. Carpenter.

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, Monroe County, Michigan

    Date of amendment request: December 15, 1997 (Reference NRC-98-
0023).
    Description of amendment request: The proposed amendment will add a 
subpart 3 to Part 2.B of the Enrico Fermi Atomic Power Plant, Unit 1 
(Fermi 1), that would allow the licensee to receive, acquire, possess, 
use and transfer byproduct material without restriction to chemical or 
physical form for sample analysis, instrument calibration, or 
associated with radioactive apparatus, hardware, tools, and equipment, 
provided the cumulative radioactive material quantity of the byproduct 
material does not exceed the criteria contained in Section 30.72, 
Schedule C, ``Quantities of Radioactive Material Requiring 
Consideration of the Need for an Emergency Plan for Responding to a 
Release.''
    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 using the standards in 10 CFR 50.92(c). The licensee's 
analysis is presented below:
    (1) Does the proposed change significantly increase the probability 
or consequences of an accident previously evaluated?
    The proposed amendment does not involve a significant increase in 
the probability or consequences of an accident. Using slightly 
contaminated apparatus or a small non-exempt radioactive source cannot 
affect the probability of the analyzed sodium or liquid waste 
accidents. The ability to possess such equipment does not in itself 
change any methods of handling liquid waste or sodium. Use of

[[Page 17224]]

contaminated equipment could potentially increase the consequences of 
an accident if it was in use or in the vicinity if an accident occurs. 
However, the increase in consequences would not be significant due to 
the limitations on radioactivity content of such equipment. The limit 
was selected to be that in 10 CFR Part 30.72, Schedule C, as the 
threshold beyond which offsite emergency plans are required. Since the 
quantity is below that requiring an offsite emergency plan, even if all 
the byproduct material allowed to be possessed by the proposed 
amendment were released during a postulated accident, the consequences 
would be significantly increased. The quantity contained in any 
specific piece of contaminated apparatus or a source would be expected 
to be even less. Therefore, this amendment does not involve a 
significant increase in the probability or consequences of an accident.
    (2) Will the proposed amendment 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 type of accident from any previously evaluated. Allowing 
possession of contaminated apparatus, tools, or equipment does not 
change methods of monitoring the facility or operation or surveillance 
of any system at Fermi 1. While possession of a different source will 
permit other instruments to be calibrated, source checked, or tested at 
Fermi 1, testing of instrumentation is routine, ordinary activity. It 
is not an activity which creates the possibility of a new or different 
type of accident.
    (3) Will the proposed change significantly reduce the margin of 
safety at the facility?
    The proposed amendment does not involve a significant reduction in 
the margin of safety at Fermi 1. No change to any system or the status 
of any systems or structures, are created by this amendment. Being able 
to have limited amounts of additional radioactive material at Fermi 1 
in the form of contaminated apparatus, tools, equipment or hardware or 
non-exempt radioactive sources will not significantly reduce the margin 
of safety because a 10 CFR Part 20 program is already in place and the 
amount of radioactive material is being limited below the amount in 10 
CFR Part 30.72, Schedule C. For these reasons, this amendment will not 
significantly reduce the margin of safety at Fermi 1.
    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, 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, Esquire, Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Branch Chief: John W. N. Hickey.

Florida Power and Light Company, et al., Docket No. 50-389, St. Lucie 
Plant, Unit No 2, St. Lucie County, Florida

    Date of amendment request: March 3, 1998.
    Description of amendment request: The amendment request proposes to 
revise the applicability of the St. Lucie Unit 2 technical 
specifications (TSs) to be consistent with St. Lucie Unit 1 TSs for 
reactor coolant system (RCS) chemistry. In addition, the amendment 
request proposes to modify the St. Lucie Unit 2 TSs by making 
administrative changes to the TS discussion of the criticality design 
features for fuel storage, and administrative changes to the technical 
review responsibilities under the cognizance of the Company Nuclear 
Review Board.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change to TS 3.4,7 will replace the existing 
applicability statement of ``At all times'' with ``All MODES.'' This 
revision will obviate the burden and personnel radiation exposures 
associated with sampling the RCS for chloride and fluoride 
concentrations during low temperature, defueled conditions. The 
existing limits, corrective actions for above limit conditions, and 
sampling requirements will be applicable for all operational MODES 
defined in the TS. The proposed applicability will continue to assure 
consistency with the bases for the RCS chemistry specification, and the 
potential for occurrence, initial conditions, or consequences of events 
considered in the safety analyses are not changed. The revisions 
proposed for TS 5.6.1.a.1 and 6.5.2.9.d are administrative in nature, 
and assure consistency with the bases for previously approved license 
amendments. Therefore, operation of the facility in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed amendment will not change the physical plant or the 
operational MODES defined in the facility license. The changes do not 
involve the addition of new equipment or the modification of existing 
equipment, nor do they alter the design of St. Lucie plant systems. 
Therefore, operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    The proposed revision to TS 3.4.7 will not change the existing RCS 
chemistry requirements that are applicable to the operational MODES 
defined in the technical specifications. However, the change will allow 
the chloride and fluoride concentrations to go unmonitored during 
certain refueling operations when there is no fuel in the reactor 
vessel. For the limited time intervals associated with this defueled 
condition, the RCS is depressurized, coolant temperature is near 
ambient, it is unlikely that the chloride and fluoride concentrations 
could be significantly increased above the concentrations that existed 
during MODE 6 prior to the core off-load, and susceptibility to 
corrosive attack from these halides is, therefore, significantly 
reduced. The existing bases for the RCS chemistry limiting conditions 
for operation are not changed, and both the bases and the proposed 
specification are consistent with the corresponding TS at St. Lucie 
Unit 1. The proposed revisions to TS 5.6.1.a.1 and TS 6.5.2.9.d are 
administrative in nature and ensure that descriptions contained therein 
are consistent with the bases for previously approved license 
amendments. Therefore, operation of the facility in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.

[[Page 17225]]

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
    Attorney for licensee: M. S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: March 12, 1998
    Description of amendment request: The licensee proposed to amend 
Turkey Point Unit 3 Facility Operating License DPR-31 to delete license 
conditions 3.I, ``Steam Generator Repair Program,'' 3.K, ``Integrated 
Schedule,'' and Section 4 of the Operating License Conditions and 
renumber Section 5 to Section 4; and to amend Turkey Point Unit 4 
Facility Operating License DRP-41 to delete license conditions 3.H, 
``Steam Generator Repair Program,'' and 3.K, ``Integrated Schedule''. 
In addition, the proposed amendments would modify Appendix A of 
Facility Operating Licenses DPR-31 and DPR-41 of the Turkey Point Units 
3 and 4 Technical Specifications (TS) to delete outdated references 
from TS Figure 5.1-2, ``Plant Area Map'' and to incorporate a recent 
organization change in TS 6.5.1.2, and 6.5.3.1.a.
    The proposed changes are administrative in nature because they 
would remove fulfilled license conditions and outdated TS references, 
and incorporate an organizational change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed changes are administrative in nature removing 
fulfilled license conditions, outdated Technical Specification 
referenced material, and reflecting an organizational change. These 
amendments will not involve a significant increase in the probability 
or consequences of an accident previously evaluated because they do not 
affect assumptions contained in plant safety analyses, the physical 
design and/or operation of the plant, nor do they affect Technical 
Specifications that preserve safety analysis assumptions. Therefore, 
the proposed changes do not affect the probability or consequences of 
accidents previously analyzed.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The use of the modified specifications cannot create the 
possibility of a new or different kind of accident from any previously 
evaluated since the proposed amendments will not change the physical 
plant or the modes of plant operation defined in the facility operating 
license. No new failure mode is introduced due to the administrative 
changes since the proposed changes do not involve the addition or 
modification of equipment nor do they alter the design or operation of 
affected plant systems, structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components are unchanged by the proposed 
amendments. The organizational change from Services Manager to 
Protection Services Manager maintained the associated level of 
management controls and the required qualifications. The proposed 
changes to the Facility Operating License Conditions and to the 
Technical Specifications are administrative and do not significantly 
reduce any of the margins 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: Library, Florida International 
University, University Park Campus, Miami, Florida 33199.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: March 2, 1998.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) 4.5.2.b.1 to delete the requirement to 
vent the operating chemical volume and control system (CVCS) 
centrifugal charging pump casing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change does not affect accident initiators or 
precursors and does not alter the design assumptions affecting the 
ability of the ECCS [emergency core cooling system] pumps to mitigate 
the consequences of an accident.
    The proposed change will align the surveillance requirements with 
the installed system design and normal operating conditions. The intent 
of the surveillance requirement ensures operability of the CVCS 
centrifugal charging pumps by verifying that the ECCS pumps and piping 
is full of water and not subjected to gas binding or hydraulic 
transients.
    Excluding the venting of the operating CVCS centrifugal charging 
pump will not effect pump operation nor subject the high head safety 
injection portion of the ECCS to potential hydraulic transients. 
Venting the operating pump under a dynamic condition at high system 
pressure is ineffective.
    The design and installation of the CVCS centrifugal charging pumps 
is such that significant non-condensable gasses do not collect in the 
pumps, whether they are running or not. Therefore, it is unnecessary to 
require periodic pump casing venting to ensure the pumps will remain 
operable. Venting of the non-operating centrifugal charging pump will 
continue to be performed, as required by TS 4.5.2b.1.
    Therefore, the proposed change does not involve a significant 
increase in the

[[Page 17226]]

probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any previously analyzed.
    The proposed change will not result in new failure modes because no 
new components or physical changes are involved with this change nor 
are the components operated in a new or different manner. The proposed 
change does not alter the ability of the CVCS centrifugal charging 
pumps to perform their intended function to mitigate the consequences 
of an initiating event within the acceptance limits assumed in the 
Updated Final Safety Analysis Report (UFSAR). The proposed change has 
no impact on component or system interactions, or the plant design 
basis. Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    There is no impact on equipment design or operation and there are 
no changes being made to the Technical Specification required safety 
limits or safety system settings that would adversely affect plant 
safety. The CVCS centrifugal charging pumps are designed and installed 
to be self-venting, such that, accumulation, if any, of non-condensable 
gasses would have no significant impact on pump operation. Since the 
proposed change will not result in new failure modes, then, the 
designed margins of safety to minimize/preclude the consequences of a 
radiological event resulting from a design basis accident remain 
unchanged. Therefore, the proposed change to eliminate the requirement 
to vent the operating CVCS centrifugal charging pump casing does 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: January 30, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) by relocating pressure-
temperature (P-T) curves, predicted radiation induced NDTT shift 
curves, and the low temperature overpressure protection (LTOP) limits 
and values from the TS to an OPPD controlled document.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes relocate the reactor coolant system (RCS) 
pressure-temperature (P-T) curves, the predicted radiation induced NDTT 
shift curve and the low temperature overpressure protection (LTOP) 
limits to the Fort Calhoun Station Unit No. 1 RCS Pressure-Temperature 
Limits Report (PTLR).
    Compliance with these curves and limits continues to be required by 
the Technical Specifications. Changes to the curves and limits will be 
controlled by TS 5.9.6, and must be in accordance with the NRC and ASME 
approved methodologies listed there and with 10 CFR 50.59.
    The FCS PTLR in combination with the limitations imposed by the TS, 
will ensure the integrity of the reactor vessel pressure boundary. 
Therefore, the proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    There will be no physical alterations to the plant configuration 
(no new or different equipment is being installed). No changes in 
operating modes or limits are proposed. The TS retain requirements to 
maintain the RCS within acceptable operational limits established in 
accordance with NRC and ASME approved methodologies and assure 
operability of the LTOP system. As such, the TS will continue to 
require compliance with the limitations being relocated to the FCS 
PTLR. Therefore, these proposed changes do not create the possibility 
of a new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    This proposed change to the FCS TS is administrative in nature 
relocating the P-T curves, NDTT curve, LTOP limits and associated TS 
requirements to the FCS PTLR in accordance with GL 96-03. Future 
updates of the FCS PTLR will be conducted under the 10 CFR 50.59 
process utilizing NRC and ASME approved methodologies (as described in 
FCS Unit No. 1 PTLR, Rev. 0 and CEOG Task 942, Report CE NPSD-683, Rev. 
02). Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: January 30, 1998.
    Description of amendment request: The proposed amendment would 
revise Facility Operating License No. DPR-40 to delete the License Term 
based on a reevaluation of the end of license fluence.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The previously evaluated accidents affected by this change are 
limited to the pressurized thermal shock (PTS) events. Vessel 
embrittlement due to fast neutron associated damage to the limiting 
beltline region reactor vessel material, which for Fort Calhoun Station 
is the lower course axial welds, is a

[[Page 17227]]

component in the PTS analysis. The fast neutron, thermal neutron and 
dpa values of the FCS reactor vessel were recalculated using actual 
power history values for Cycles 1 through 14 rather than conservative 
estimates, with the revised BUGLE-93 cross sections from the ENDF/B-VI 
cross section library to appropriately account for the iron atoms in 
the thermal shield and a methodology that the NRC has previously 
approved for neutron fluence calculations performed by Westinghouse. 
The evaluation included data from the three surveillance capsules (W-
225, W-265, and W-275) previously removed and analyzed. The evaluation 
results indicate that the FCS reactor vessel is able to reach current 
licensed life without exceeding the 10 CFR 50.61 screening criteria for 
RTPTS of 270 deg.F for limiting axial welds.
    In accordance with 10 CFR 50.61, this assessment must be updated 
whenever there is a significant change in projected values of 
RTPTS or upon request for a change in the expiration date of 
the facility. Since these requirements are contained in 10 CFR 50.61, 
Section 3.E can be deleted from Operating License No. DPR-40 without 
resulting in a significant increase in the probability or consequences 
of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change does not physically alter the configuration of 
the plant and no new or different mode of operation is proposed. 
Increasing the long term load factor from 0.77 to 0.85 more accurately 
projects RTPTS by accounting for improvement in FCS 
operating cycle efficiency. Requirements for assessing and reporting 
RTPTS are contained in 10 CFR 50.61 and therefore, the 
proposed change does not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The margin of safety is defined by the draft regulatory guide DG-
1053 for neutron fluence calculations which requires the methodology to 
be capable of providing best estimate fluence evaluations within plus 
or minus 20 percent (1). The analysis shows that the 
applicable regulatory criteria are met and therefore, the proposed 
change does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: March 18, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specifications by changing the title of the Shift 
Supervisor to Shift Manager.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    OPPD proposes to change the title of the Shift Supervisor to Shift 
Manager. The qualifications required of these individuals and the 
duties they perform are unchanged. The title of Shift Manager better 
conveys the appropriate level of responsibility and authority required 
of the position. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    There will be no physical alterations to the plant configuration 
(no new or different equipment is being installed). No changes in 
operating modes or limits are proposed. The qualifications required of 
these individuals and the duties they perform are unchanged. Therefore, 
these proposed changes do not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed change in the title of the Shift Supervisor to Shift 
Manager is strictly administrative. The qualifications required of 
these individuals and the duties they perform are unchanged. The title 
of Shift Manager better conveys the appropriate level of responsibility 
and authority required of the position. Therefore, this change does not 
involve a significant increase in the probability or consequences of an 
accident previously evaluated.
    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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne Count, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for: (1) the standby liquid control (SLC) 
system that ensures that there is a functioning flow path from the 
boron injection tank to the reactor pressure vessel, and (2) the scram 
discharge volume (SDV) that verifies system performance of the vent and 
drain valves. Specifically, the interval for SLC testing is being 
increased from once every 18 months to once every 24 months for a 
maximum interval of 30 months including the 25 percent grace period; 
and, from once every 36 months to once every 48 months for those 
surveillances on a staggered test basis. The frequency for testing the 
SDV vent and drain valves would be increased from once every 18 months 
to once every 24 months for a maximum interval of 30 months including 
the 25 percent grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or

[[Page 17228]]

consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the SAR. Therefore, this change will have no impact on the 
probability of an accident previously evaluated. By changing the 
surveillance Frequency from 18 months plus grace to a maximum of 30 
months, the consequences of an accident previously evaluated in the SAR 
are not significantly increased. This is based on the fact that the 
evaluation of the subject changes demonstrated that the overall impact, 
if any, on the systems availability is minimal. Since the impact on the 
systems is minimal, it can be concluded that the overall impact on the 
plant accident analysis is negligible. Furthermore, it is shown that 
the performance history for the subject systems does not indicate any 
failures which would invalidate the conclusions reached in this 
evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for performance of channel calibrations on: 
(1) the reactor protection system (RPS) instrumentation, (2) the source 
range monitor (SRM) instrumentation, (3) the feedwater-main turbine 
high-water-level trip instrumentation, (4) the post accident monitoring 
(PAM) instrumentation, (5) the remote shutdown system instrumentation, 
(6) the end-of-cycle recirculation pump trip (EOC-RPT) instrumentation, 
(7) the anticipated transient without scram recirculation pump trip 
(ATWS-RPT) instrumentation, (8) the emergency core cooling system 
(ECCS) instrumentation, (9) the rector core isolation cooling (RCIC) 
system instrumentation, (10) the primary containment isolation 
instrumentation, (11) secondary containment isolation instrumentation, 
(12) the control room emergency outside air supply (CREOAS) system 
instrumentation, (13) the loss of power (LOP) instrumentation, and (14) 
the RPS electric power monitoring instrumentation. Specifically, the 
intervals for the associated channel calibration would be increased 
from either once every 18 months or refueling cycle to once every 24 
months for a maximum interval of 30 months including the 25 percent 
grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the SAR. Furthermore, the instrument drift has been 
evaluated and found to be acceptable for the extended operating 
cycle[.] Therefore, this change will have no impact on the probability 
of an accident previously evaluated. By changing the Surveillance 
Frequency from 18 months plus grace to a maximum of 30 months, the 
consequences of an accident previously evaluated in the SAR are not 
significantly increased. This is based on the fact that the evaluation 
of the subject changes demonstrated that the overall impact, if any, on 
the systems availability is minimal and instrument drift over the 
extended operating cycle has been evaluated and found to be acceptable. 
Since the impact on the systems and from instrument drift is minimal, 
it can be concluded that the overall impact on the plant accident 
analysis is negligible. Furthermore, it is shown that the performance 
history for the subject systems does not indicate any failures which 
would invalidate the conclusions reached in this evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library,

[[Page 17229]]

Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for: (1) the integrated leak test of each 
system listed as a primary coolant source outside containment, and (2) 
the engineered safety feature filter ventilation systems in the 
ventillation filter testing program. Specifically, the interval for 
these tests would be increased from once every 18 months to once every 
24 months for a maximum interval of 30 months including the 25 percent 
grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the SAR [safety analysis report]. Therefore, this change 
will have no impact on the probability of an accident previously 
evaluated. By changing the Surveillance Frequency from 18 months plus 
grace to a maximum of 30 months, the consequences of an accident 
previously evaluated in the SAR are not significantly increased. This 
is based on the fact that the evaluation of the subject changes 
demonstrated that the overall impact, if any, on the systems 
availability is minimal. Because the impact on the systems is minimal, 
it can be concluded that the overall impact on the plant accident 
analysis is negligible. Furthermore, it is shown that the performance 
history for the subject systems does not indicate any failures which 
would invalidate the conclusions reached in this evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for the AC and DC electrical power system 
sources. Specifically, the intervals for various functional tests would 
be increased from once every 18 months to once every 24 months for a 
maximum interval of 30 months including the 25 percent grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the [safety analysis report] SAR. Therefore, this change 
will have no impact on the probability of an accident previously 
evaluated. By changing the Surveillance Frequency from 18 months plus 
grace to a maximum of 30 months, the consequences of an accident 
previously evaluated in the SAR are not significantly increased. This 
is based on the fact that the evaluation of the subject changes 
demonstrated that the overall impact, if any, on the systems 
availability is minimal. Because the impact on the systems is minimal, 
it can be concluded that the overall impact on the plant accident 
analysis is negligible. Furthermore, it is shown that the performance 
history for the subject systems does not indicate any failures which 
would invalidate the conclusions reached in this evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 17230]]

standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would lower the 
minimum allowable low power setpoint for the control rod block 
instrumentation rod worth minimzer (RWM) from less than or equal to 20 
percent rated thermal power (RTP) to less than or equal to 10 percent 
RTP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    This change establishes the minimum allowable low power setpoint of 
the RWM as less than or equal to 10% RTP. This change will not result 
in a significant increase in the probability of an accident previously 
evaluated because the Operability of the RWM not considered an 
initiator for any accidents previously analyzed. This change will not 
result in a significant increase in the consequences of an accident 
previously evaluated because, as documented in Amendment 17 to NEDE-
24011-P-A (GESTAR-II) and the associated NRC SER [safety evaluation 
report], if core power level exceeds 10% RTP, no control rod pattern 
can generate rod worths such that the fuel enthalpy would exceed the 
280 cal/gm fuel enthalpy limit during the worst RDA [rod drop 
accident].
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SSC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The proposed change does not involve a significant reduction in a 
margin of safety because, as documented in Amendment 17 to NEDE-24011-
P-A (GESTAR-II) and the associated NRC SER, if core power level exceeds 
10% RTP, no control rod pattern can generate rod worths such that the 
fuel enthalpy would exceed the 280 cal/gm fuel enthalpy limit during 
the worst RDA.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for the: (1) drywell-to-suppression chamber 
vacuum breaker leakage test, (2) the primary containment isolation 
valves functional tests, (3) each reactor instrumentation line excess 
flow check valve (EFCV) functional tests, (4) the suppression chamber-
to-drywell vacuum breaker opening setpoint test, (5) the system 
functional test, visual examination, and heater phase resistance to 
ground tests for the drywell and suppression chamber hydrogen 
recombiners, (6) the secondary containment vacuum tests of the standby 
gas treatment (SGT) subsystem, (7) the seconday containment isolation 
valves (SCIVs) functional tests, and (8) the SGT subsytem functional 
tests. Specifically, the intervals for these tests would be increased 
from once every 18 months to once every 24 months for a maximum 
interval of 30 months including the 25 percent grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the [Safety Analysis Report] SAR. Therefore, this change 
will have no impact on the probability of an accident previously 
evaluated. By changing the Surveillance Frequency from 18 months plus 
grace to a maximum of 30 months, the consequences of an accident 
previously evaluated in the SAR are not significantly increased. This 
is based on the fact that the evaluation of the subject changes 
demonstrated that the overall impact, if any, on the systems 
availability is minimal. Since the impact on the systems is minimal, it 
can be concluded that the overall impact on the plant accident analysis 
is negligible. Furthermore, it is shown that the performance history 
for the subject systems does not indicate any failures which would 
invalidate the conclusions reached in this evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the

[[Page 17231]]

subject systems, it can be concluded that the assumptions in the 
licensing basis are not impacted by the changes in the subject 
requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for: (1) the system functional test of the 
core spray and low pressure coolant injection system, and (2) the high 
pressure coolant injection (HPCI) and the low pressure HPCI flow test. 
Specifically, the intervals for system functional tests and response 
time tests would be increased from once every 18 months to once every 
24 months for a maximum interval of 30 months including the 25 percent 
grace period. Additionally, the surveillance test intervals for: (1) 
the system functional test of the automatic depressurization system 
(ADS), and (2) the system functional test and low pressure flow test of 
the reactor core isolation cooling (RCIC) 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the SAR. Therefore, this change will have no impact on the 
probability of an accident previously evaluated. By changing the 
Surveillance Frequency from 18 months plus grace to a maximum of 30 
months, the consequences of an accident previously evaluated in the SAR 
are not significantly increased. This is based on the fact that the 
evaluation of the subject changes demonstrated that the overall impact, 
if any, on the systems availability is minimal. Since the impact on the 
systems is minimal, it can be concluded that the overall impact on the 
plant accident analysis is negligible. Furthermore, it is shown that 
the performance history for the subject systems does not indicate any 
failures which would invalidate the conclusions reached in this 
evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test interval for the channel calibration of the reactor 
coolant system leakage detection instrumentation. The surveillance test 
interval would be increased from once every 18 months to once every 24 
months for a maximum interval of 30 months including the 25 percent 
grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the [S]urveillance 
Frequency from 18 months to 24 months. The change in [S]urveillance 
Frequency is not assumed to be an accident initiator for any accidents 
previously evaluated in the SAR [safety analysis report]. Furthermore, 
the instrument drift has been evaluated and found to be acceptable for 
the extended operating cycle. Therefore, this change will have no 
impact on the probability of an accident previously evaluated. By 
changing the Surveillance Frequency from18 months plus grace to a 
maximum of 30 months, the consequences of an accident previously 
evaluated in the SAR are not significantly increased. This is based on 
the fact that the evaluation of the subject changes demonstrated that 
the overall impact, if any, on the systems availability is minimal and 
instrument drift over the extended operating cycle has been evaluated 
and found to be acceptable. Since the impact on the systems and from 
instrument drift is minimal, it can be concluded that the overall 
impact on the plant accident analysis is negligible. Furthermore, it is 
shown that the performance history for the subject systems does not 
indicate any failures which would invalidate the conclusions reached in 
this evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant

[[Page 17232]]

systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would remove the 
operability requirement for the 480 volt engineered safeguards systems 
bus 0565 undervoltage relay (degraded voltage 65 percent and 92 
percent) in the loss of power instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes remove from the SSES CTS [Susquehanna Steam 
Electric Station current technical specifications] items that are 
informational or implementing details that are adequately and more 
appropriately controlled by the licensee. Additionally, the proposed 
changes remove from the SSES CTS items that are contained in the Code 
of Federal Regulations or other regulatory documents and, therefore, do 
not need to be repeated in the SSES ITS [improved technical 
specifications]. These requirements being moved to another controlled 
document or removed from Technical Specifications are not deleted or 
changed. Therefore, these changes will not result in any changes to the 
requirements specified in the SSES CTS, but will reduce the level of 
regulatory control on the identified requirements. The level of 
regulatory control has no impact on the probability or the consequences 
of an accident previously evaluated, therefore, these changes have no 
impact on the probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes will not involve any physical changes to plant 
systems, structures, or components (SSC), or the manner in which these 
SSC are operated, maintained, modified, tested, or inspected. The 
proposed changes will not impose or eliminate any requirements. 
Therefore, these changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety as defined in the bases of any Technical 
Specification is not reduced. The requirements being moved to another 
controlled document or removed from Technical Specifications remain the 
same as stated in the SSES CTS. Therefore, no reduction in a margin of 
safety will be permitted.
    Removal of these items from SSES CTS eliminates the requirement for 
NRC review and approval of revisions in accordance with 10 CFR 50.92. 
Elimination of this administrative process does not have a margin of 
safety that can be evaluated. However, the proposed changes are 
consistent with the BWR [Boiling-Water Reactor] Standard Technical 
Specification, NUREG-1433, Rev. 1, which was approved by the NRC. 
Revising the Technical Specifications to reflect the approved level of 
detail ensures no significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996.
    Description of amendment request: The change would increase the 
surveillance test intervals for: (1) the reactor protection system 
(RPS) instrumentation, (2) the feedwater-main turbine high-water-level 
trip instrumentation, (3) the end of cycle recirculation pump trip 
(EOC-RPT) instrumentation, (4) the anticipated transient without scram 
recirculation pump trip (ATWS-RPT) instrumentation, (5) the emergency 
core cooling system (ECCS) instrumentation, (6) the reactor core 
isolation cooling (RCIC) system instrumentation, (7) RPS electric power 
monitoring system instrumentation, (8) primary containment isolation 
instrumentation, (9) secondary containment isolation instrumentation, 
(10) the control room emergency outside air supply (CREOAS) system 
instrumentation, and (11) the loss of power (LOP) instrumentation. 
Specifically, the intervals for various logic system functional tests 
and response time tests would be increased from once every 18 months to 
once every 24 months for a maximum interval of 30 months including the 
25 percent grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

[[Page 17233]]

    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the SAR. Therefore, this change will have no impact on the 
probability of an accident previously evaluated. By changing the 
Surveillance Frequency from 18 months plus grace to a maximum of 30 
months, the consequences of an accident previously evaluated in the SAR 
are not significantly increased. This is based on the fact that the 
evaluation of the subject changes demonstrated that the overall impact, 
if any, on the systems availability is minimal. Since the impact on the 
systems is minimal, it can be concluded that the overall impact on the 
plant accident analysis is negligible. Furthermore, it is shown that 
the performance history for the subject systems does not indicate any 
failures which would invalidate the conclusions reached in this 
evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996, and March 2, 1998.
    Description of amendment request: The change would increase the 
surveillance test interval for the: (1) emergency service water (ESW) 
system functional test, (2) the control room emergency outside air 
supply (CREOAS) system functional test and control room pressurization 
test, and (3) the main turbine bypass system functional and response 
time tests. Specifically, the interval for these tests would be 
increased from once every 18 months to once every 24 months for a 
maximum interval of 30 months including the 25 percent grace period.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes involve a change in the surveillance Frequency 
from 18 months to 24 months. The change in surveillance Frequency is 
not assumed to be an accident initiator for any accidents previously 
evaluated in the [safety analysis report] SAR. Therefore, this change 
will have no impact on the probability of an accident previously 
evaluated. By changing the Surveillance Frequency from 18 months plus 
grace to a maximum of 30 months, the consequences of an accident 
previously evaluated in the SAR are not significantly increased. This 
is based on the fact that the evaluation of the subject changes 
demonstrated that the overall impact, if any, on the systems 
availability is minimal. Since the impact on the systems is minimal, it 
can be concluded that the overall impact on the plant accident analysis 
is negligible. Furthermore, it is shown that the performance history 
for the subject systems does not indicate any failures which would 
invalidate the conclusions reached in this evaluation.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    This proposed change will not involve any physical changes to plant 
systems, structures, or components (SCC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The margin of safety has not been significantly reduced. Although, 
there will be an increase in the interval between the subject 
surveillance tests, the evaluation of the changes demonstrates that 
there is no evidence of any failures which would impact the subject 
systems availability. Based on the fact that the increased testing 
interval has a minimal impact on the subject systems, it can be 
concluded that the assumptions in the licensing basis are not impacted 
by the changes in the subject requirements and commitments.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996, and March 2, 1998.
    Description of amendment request: The change would add a 
surveillance requirement and acceptance criteria to verify the source 
range monitor (SRM) count rate versus the signal to noise ratio of the 
SRMs. This change also incorporates a new SRM count rate to signal to 
noise ratio curve which is based on General Electric Service 
Information Letter (SIL) 478.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 17234]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes provide requirements determined to be more 
conservative than the existing requirements for operation of the 
facility.
    Therefore, these changes establish or maintain adequate assurance 
that components are operable when necessary for the prevention or 
mitigation of accidents or transients and that plant variables are 
maintained within limits necessary to satisfy the assumptions for 
initial conditions in the safety analysis. Therefore, these changes do 
not involve any increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes will not involve any physical changes to plant 
systems, structures, or components (SSC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The imposition of more restrictive requirements either has no 
impact on or increases the margin of plant safety. As provided in the 
discussion of each of the changes, each change in this category 
provides additional requirements designed to enhance plant safety. Each 
of the changes maintains requirements within the safety analyses and 
licensing basis. Therefore, these changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388, 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: August 1, 1996, as supplemented March 2, 
1998.
    Description of amendment request: The change would reduce the 
allowable values for the reactor protection system instrumentation 
scram discharge volume water level--high scram setpoints: (1) for the 
level transmitter from less than or equal to 88 gallons to less than or 
equal to 66 gallons, and (2) for the float switch from less than or 
equal to 88 gallons to less than or equal to 62 gallons in order to be 
consistent with the design setpoint calculations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes provide requirements determined to be more 
conservative than the existing requirements for operation of the 
facility. Therefore, these changes establish or maintain adequate 
assurance that components are operable when necessary for the 
prevention or mitigation of accidents or transients and that plant 
variables are maintained within limits necessary to satisfy the 
assumptions for initial conditions in the safety analysis. Therefore, 
these changes do not involve any increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes will not involve any physical changes to plant 
systems, structures, or components (SSC). The changes in normal plant 
operation are consistent with the current safety analysis assumptions. 
Therefore, these changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin of 
safety?
    The imposition of more restrictive requirements either has no 
impact on or increases the margin of plant safety. As provided in the 
discussion of each of the changes, each change in this category 
provides additional requirements designed to enhance plant safety. Each 
of the changes maintains requirements within the safety analyses and 
licensing basis. Therefore, these changes do not involve a reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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: December 30, 1997.
    Description of amendments request: The proposed amendments would 
revise the Technical Specification surveillance requirements for the 
Auxiliary Building and Service Water Building batteries to remove the 
existing 1.75 volt minimum individual cell voltage associated with the 
``service test'' acceptance criterion and replace it with a reference 
to the battery load profile specified in the Final Safety Analysis 
Report, Section 8.3.2.
    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 changes to remove and replace specific acceptance 
criterion in the Technical Specifications with a reference to more 
detailed and bounding criteria in the FSAR [Final Safety Analysis 
Report] for service tests on the batteries do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated in the Farley FSAR. The AB [Auxiliary Building] and SWB 
[Service Water Building] batteries do not initiate any accident. 
Clarification of testing acceptance criteria does not adversely affect 
the batteries ability to mitigate the consequences of any accident in 
the

[[Page 17235]]

Farley FSAR. No new accident initiators are identified as a result of 
this proposed revision. No new performance requirements for any system 
that is used to mitigate dose consequences have been imposed by this 
proposed change. No input assumptions to any dose consequence 
calculations are affected by this proposed change. All previously 
reported dose consequences remain bounding. Therefore, the radiological 
consequences resulting from any accident previously evaluated in the 
FSAR are not increased.
    2. The proposed changes to remove and replace specific acceptance 
criterion in the Technical Specifications with a reference to more 
detailed and bounding criteria in the FSAR for service tests on the 
batteries do not create the possibility of a new or different kind of 
accident from any previously evaluated in the Farley FSAR. No new 
accident scenarios, failure mechanisms or limiting single failures are 
introduced as a result of the clarifications to the battery service 
test acceptance criteria. No new challenges to the safety-related AB or 
SWB 125VDC Distribution Systems have been identified. The 125VDC 
Systems including the batteries have not been modified. Farley will 
continue to perform service discharge surveillance tests in accordance 
with the frequency requirements of the Technical Specifications to 
demonstrate battery operability. Previously identified accident 
scenarios remain bounding because the performance requirements of the 
batteries have not been changed. Therefore, the possibility of a new or 
different kind of accident is not created.
    3. The proposed changes to remove and replace specific acceptance 
criterion in the Technical Specifications with a reference to more 
detailed and bounding criteria in the FSAR for service tests on the 
batteries do not involve a significant reduction in the margin of 
safety. All previously established acceptance limits continue to be met 
for all events since the battery function is to provide power during 
the time between LOSP [loss of offsite power] & D/G [diesel generator] 
start and in the event of battery charger failure to mitigate the 
consequences of any accident scenario. Relocating and clarifying 
service test acceptance criteria will not invalidate the battery 
function. There are no physical modifications required to the AB or SWB 
125VDC Distribution Systems or the batteries. This change will not 
affect the operation of the batteries or any other safety-related 
equipment. Applicable values, reflected in the governing electrical 
design calculations, will be incorporated into the FSAR and will remain 
or be included in the surveillance test procedures. Since current 
battery performance acceptance limits will continue to be met, there is 
no 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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama.
    NRC Project Director: Herbert N. Berkow.

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 13, 1998 (TS 97-04).
    Brief description of amendments: The amendments change the Sequoyah 
(SQN) Technical Specifications (TS) by relocating the mechanical 
snubber requirements from Section 3.7.9 of the TS to the SQN Technical 
Requirements Manual.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TS, does not involve a 
significant hazards consideration. TVA's conclusion is based on its 
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    The proposed revision to the TS relocates the requirements for SQN 
snubbers without changing the current requirements and deletes an 
obsolete License Condition. TVA does not consider the snubbers to be 
the source of any accident; therefore, this administrative relocation 
of the requirements and License Condition deletion will not increase 
the possibility of an accident. The capability of the snubbers will 
continue to provide the same function in support of accident 
mitigation. Changes to the relocated requirements will be processed, in 
accordance with 10 CFR 50.59, to ensure the snubber functions will be 
properly maintain[ed]. Therefore, the proposed relocation of the 
snubber requirements and License Condition deletion will not increase 
the consequences of an accident.
    B. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The SQN safety-related snubbers provide support for mitigation 
functions associated with previously evaluated accidents and are not 
the initiator of any accident. The proposed change does not alter the 
current functions of the snubbers; therefore, it will not create the 
possibility of a new or different kind of accident.
    C. The proposed amendment does not involve a significant reduction 
in a margin of safety.
    The requirements for SQN safety-related snubbers are unchanged by 
the proposed relocation of the requirements to the SQN TRM [Technical 
Requirements Manual] and the License Condition deletion. The function 
of the snubbers and surveillances to ensure operability will remain the 
same as currently required by the TS. Changes to these requirements 
will be evaluated, in accordance with 10 CFR 50.59, to ensure 
acceptability and NRC review as required. Therefore, the proposed 
change will not result in a reduction in a margin of safety.
    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, 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 application for amendments: February 25, 1998 (TS 97-06).
    Brief description of amendments: The amendments change the Sequoyah

[[Page 17236]]

(SQN) Technical Specifications (TSs) for the emergency diesel 
generators (D/Gs) by 1) incorporating vendor-recommended changes to the 
D/G inspection program, 2) revising the D/G surveillance program, and 
3) changing the allowable D/G steady-state voltage range.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:
    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TSs (or operating 
license[s]), does not involve a significant hazards consideration. 
TVA's conclusion is based on its evaluation, in accordance with 10 CFR 
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c).
    Part 1--Vendor Recommended Inspections:
    The proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed revision to the TS deletes the requirements for 18-
month inspections from the TS. TVA does not consider the inspections to 
be the source of any accident; therefore, this deletion will not 
increase the possibility of an accident. The D/Gs come within the 
purview of 10 CFR 50.65, which monitors the effectiveness of 
maintenance at nuclear power plants. The capability of the D/Gs to 
provide the required safety function in support of accident mitigation 
will be unaffected. Therefore, the proposed deletion of the inspection 
requirements will not increase the consequences of an accident.
    The proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The emergency D/Gs provide support for mitigation functions 
associated with previously evaluated accidents and are not the 
initiator of any accident. The proposed change does not alter the 
current functions of the D/Gs; therefore, it will not create the 
possibility of a new or different kind of accident.
    The proposed amendment does not involve a significant reduction in 
a margin of safety.
    The requirements for emergency D/Gs are unchanged by the proposed 
deletion of the requirements from TSs. The function of the emergency D/
Gs and surveillances to ensure operability will remain the same as 
currently required by the TS. NRC will continue to monitor the 
effectiveness of D/G maintenance as required by 10 CFR 50.65. 
Therefore, the proposed change will not result in a reduction in a 
margin of safety.
    Part 2--D/G Online Testing:
    The proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed amendment to allow the load rejection tests and the 
24-hour D/G 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 Final Safety 
Analysis Report (FSAR) since the capability to safely shutdown the 
plant following a LOOP [loss of offsite power], LOCA [loss of coolant 
accident] or LOCA/LOOP coincident with a single failure is maintained 
throughout the surveillance test. Other aspects of D/G parallel testing 
(protective devices, risks interactions with offsite power 
capabilities, and operation) are unaffected by the proposed TS change. 
Required Class-lE onsite power operability during normal operation, 
shutdown cooling, LOOP, and accident conditions will be the same.
    Performance of the new SR [Surveillance Requirement] 4.8.1.1.2.g.4 
requires the D/Gs to be at the same system conditions prior to the test 
(stabilized operating temperature) as previously required. The LOOP 
start will continue to be performed as required by SR 4.8.1.1.2.d.4.b.
    In addition, the performance of proposed SRs 4.8.1.1.2.g.1, 
4.8.1.1.2.g.2, 4.8.1.1.2.g.3, or 4.8.1.1.2.g.4 during Modes 1, 2 or 3 
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 SR 4.8.1.1.2.g.1, 4.8.1.1.2.g.2, or 
4.8.1.1.2.g.3 during Modes 1, 2 or 3 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, TVA concludes that the above change does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The requested changes do not result in a new or different kind of 
accident from that previously analyzed in SQN's FSAR. The changes 
propose to eliminate restrictions of the plant operating modes in which 
standby D/G system testing may be performed, but does not change the 
type of testing performed and are not due to modification of the system 
design. NRC's assessment of the testing of the D/Gs in the 
configuration proposed is documented in Section 8.3.1, Supplement 1 of 
the SER (NUREG-0011).
    The proposed amendment does not involve a significant reduction in 
a margin of safety.
    As previously stated, performance of proposed SRs 4.8.1.1.2.g.1, 
4.8.1.1.2.g.2, 4.8.1.1.2.g.3, or 4.8.1.1.2.g.4 during Modes 1, 2 or 3 
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. It also 
does not change any setpoints or limits established for accident 
mitigation. Therefore, implementation of the proposed amendment will 
not reduce the margin of safety for this system.
    Part 3--D/G Steady State Allowable Voltage Range:
    The proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed revisions to the SRs conservatively restrict the 
allowable range of the D/G steady state voltage. The capability of the 
D/Gs to provide the required safety function, in support of accident 
mitigation, will be unaffected or enhanced. Therefore, the proposed 
revision of the SRs will not increase the consequences of an accident.
    The proposed amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not alter the current functions of the D/
Gs; therefore, they will not create the possibility of a new or 
different kind of accident.
    The proposed amendment does not involve a significant reduction in 
a margin of safety.
    The requirements for emergency D/Gs are unchanged by the 
conservative revision of the allowable range of the D/G steady state 
voltage or clarification of the required voltage and frequency after 10 
seconds. The function of the emergency D/Gs and surveillances to ensure 
operability will remain the same as currently required by the TS. 
Therefore, the proposed changes will not result in a reduction in a 
margin of safety.

[[Page 17237]]

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

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

    Date of amendment request: February 25, 1998.
    Description of amendment request: Requests Technical Specifications 
changes to permit use of Option B of 10 CFR 50, Appendix J, for 
containment leakage testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed TS changes do not involve any physical or operational 
changes to structures, systems or components. The current safety 
analysis and design basis for the accident mitigation functions of the 
containment, the airlocks, and the containment isolation valves are 
maintained. On-site and off-site dose consequences remain unaffected. 
Containment leakage rate testing is not an accident initiator.
    2. The proposed license amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The accidents considered are found in the Safety Analysis, Section 
14 of the USAR. The proposed change does not involve a change to the 
plant design (structures, systems or components) or operation. No new 
failure mechanisms beyond those already considered in the current plant 
Safety Analysis are introduced. No new accident is introduced and no 
safety-related equipment or safety functions are altered. The proposed 
change does not affect any of the parameters or conditions that 
contribute to initiation of any accidents.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    The implementation of Option B potentially affects the frequency of 
Type A, B, and C containment testing. Except for the determination of 
test frequency, the methods for performing the actual tests are not 
changed. NUREG-1493, ``Performance-Based Containment Leak-Test 
Program'', dated September, 1995, which forms the basis for the 
Appendix J revision, concludes that adoption of performance-based 
testing will not significantly reduce the margin of safety. Therefore, 
the proposed TS amendment will not involve a significant reduction in a 
margin of safety and will continue to support the design and licensing 
basis of ensuring an essentially leak-tight containment boundary.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Project Director: Richard P. Savio.

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

    Date of amendment request: March 4, 1998.
    Description of amendment request: Requests Technical Specifications 
changes to provide a one hour Limiting Condition for Operation (LCO) 
that will permit a safety injection pump to be used for addition of 
make-up fluid to safety injection accumulators 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. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    While filling a safety injection (SI) accumulator, the large break 
loss of coolant accident (LOCA) would be the bounding accident for pump 
runout concerns. The proposed LCO would allow relaxation of a single 
failure being assumed during the short duration of the accumulator 
fill. The SI pump filling the SI accumulator will be considered to be 
operable while filling the accumulator.
    Using current KNPP PRA methods, this configuration results in a 
core damage frequency (CDF) of 5x10-5/year during the five 
minutes it exists. The increased core damage probability (CDP) due to 
an accumulator fill is 8x10-11. Conservatively assuming that 
the accumulator fill occurs every three weeks, the total CDP increase 
is 1.3x10-9 in a year. The configuration specific DF and CDP 
increase are well below the limits of 1.0x10-3/year and 
1.0x10-6, respectively, in the Electric Power Research 
Institute's PRA Applications Guide. The increase in probability is 
extremely low and well within industry PRA limits.
    With entry into a one hour action statement, the single failure 
criterion is relaxed (i.e., a postulated failure of an SI pump is not 
required) and both SI pumps will provide the required flow to ensure 
accident mitigation and prevent pump run out. By assuming both SI pumps 
are available, there is no impact on the accident analysis.
    By remaining within the bounds of the accident analysis and the 
extremely low increase in the probability of a LOCA concurrent with an 
accumulator fill, WPSC concludes that this change does not 
significantly increase the probability or consequences of an accident 
previously evaluated.
    2. The proposed license amendment requests does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The change allows relaxation of single failure criteria during the 
short time an SI accumulator would be filled. The SI pump filling the 
accumulator will be available during the short filling period.
    With entry into a one hour action statement, the single failure 
criterion is relaxed (i.e., a postulated failure of an SI pump is not 
required) and both SI pumps will provide the required flow to ensure 
accident mitigation and prevent pump runout.
    The proposed change is not a result of a hardware change, and with 
one SI pump considered to be available during an accumulator fill, all 
the accident analysis requirements are satisfied. Therefore, WPSC 
concludes that this

[[Page 17238]]

proposed change does not create the possibility of a new or different 
kind of accident.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.
    With both SI pumps available during an accumulator fill, there is 
not an SI pump runout concern and all the requirements of the accident 
analysis are met. Due to the infrequent occurrence, short duration and 
extremely low probability of LOCA occurring during an accumulator fill, 
WPSC concludes there is not significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Project Director: Richard P. Savio.

Previously Published Notices of Consideraton 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.

Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power 
Station, Unit 2, Grundy County, Illinois

    Date of amendment request: March 19, 1998.
    Description of amendment request: The proposed amendment would 
reflect a change in the Dresden, Unit 2, minimum critical power ratio 
(MCPR) Safety Limit and revise footnotes in Technical Specifications 
(TS) Section 5.3, to allow the use of Siemens Power Corporation (SPC) 
ATRIUM-9B fuel.
    Date of publication of individual notice in Federal Register: March 
26, 1998 (63 FR 14735).
    Expiration date of individual notice: April 27, 1998.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Duquesne Light Company, et al., Docket Nos. 50-334 and 50-412, Beaver 
Valley Power Station, Unit Nos. 1 and 2, Shippingport, Pennsylvania

    Date of amendment request: March 16, 1998.
    Brief description of amendment request: These amendments add a new 
Limiting Condition for Operation (LCO) 3.0.6 to TS Section 3/4.0, 
``APPLICABILITY.'' The new LCO 3.0.6 provides specific guidance for 
returning equipment to service under administrative control to perform 
testing required to demonstrate OPERABILITY.
    Date of publication of individual notice in Federal Register: March 
24, 1998 (63 FR 14142).
    Expiration date of individual notice: Comment period April 7, 1998, 
and hearing period April 23, 1998.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Northern States Power Company, Docket No. 50-263, Monticello Nuclear 
Generating Plant, Wright County, Minnesota

    Date of amendment request: March 13, 1998.
    Description of amendment request: The proposed amendment would 
revise Section 2.1.A of the Technical Specifications (TS) to change the 
safety limit minimum critical power ratio (SLMCPR) values from 1.08 to 
1.10 for two recirculation pump operation, and from 1.09 to 1.11 for 
single loop operation. The amendment would also revise pages 6 and 249b 
of the TS to indicate that the revised SLMCPR values are applicable 
only to operating cycle 19.
    Date of individual notice in the Federal Register: March 20, 1998 
(63 FR 13704).
    Expiration date of individual notice: April 20, 1998.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.

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

    Date of amendment request: March 12, 1998, TXX-98076.
    Description of amendment request: The proposed amendment would 
provide a temporary Technical Specification change for SRs 
4.8.1.1.2f.4)b) and 4.8.1.1.2f.6)b) to allow the verification of the 
auto connected shut-down loads through the load sequencer to be 
performed at power for fuel cycle 6 on Unit 1 and fuel cycle 4 on Unit 
2.
    Date of individual notice in the Federal Register: March 27, 1998 
(63 FR 14974).
    Expiration date of individual notice: April 13, 1998.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of amendment request: March 20, 1997.
    Description of amendment request: The licensee requested to modify 
their licensing basis by limiting the time the large (18'') purge and 
vent valves may be open to containment.
    Date of publication of individual notice in Federal Register: March 
27, 1998. (63 FR 14976).
    Expiration date of individual notice: April 27, 1998.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

[[Page 17239]]

License, Proposed No Significant Hazards Consideration Determination, 
and Opportunity for A Hearing in connection with these actions was 
published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendment: December 17, 1997.
    Brief description of amendment: These amendments modify the 
technical specifications (TS) to remove the reference to Exide 
batteries with a generic reference to low specific gravity cell 
batteries.
    Date of issuance: March 16, 1998.
    Effective date: March 16, 1998.
    Amendment No.: Unit 1--116; Unit 2--109; Unit 3--88.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2272).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, 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: October 22, 1997.
    Brief description of amendments: The amendments change the 
Technical Specifications (TSs) to incorporate both steady state and 
transient degraded voltage setpoints as opposed to the current single 
degraded voltage setpoints. Additionally, the TS decreases the 4 kV 
voltage range of the emergency diesel generators to assure that the new 
steady state degraded voltage relays are not actuated during testing.
    Date of issuance: March 17, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 226 and 200.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61838).
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated March 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: November 6, 1997, as 
supplemented by letters dated January 27, March 3, March 6, March 13, 
and March 18, 1998.
    Brief Description of amendments: The amendments change the 
Technical Specifications (TS) for the Brunswick Steam Electric Plant 
(BSEP) Units 1 and 2 to allow three 18-month diesel generator (DG) 
surveillance requirements (SR) to be performed during both plant 
operation (Operational Conditions 1 and 2) and shutdown (Operational 
Conditions 3, 4, and 5) rather than, as currently required, only during 
shutdown. The first SR is an inspection of the DG involving a partial 
disassembly. The second ensures that non-critical DG protective 
functions are bypassed on an Emergency Core Cooling system actuation 
signal. The third verifies that the DG operates for greater than or 
equal to 60 minutes while loaded to at least 3500 kw, which bounds the 
maximum expected post-accident DG loading. The proposed amendments 
additionally remove an expired footnote from the BSEP Unit 2 DG TS.
    Date of issuance: March 26, 1998.
    Effective date: March 26, 1998
    Amendment Nos.: 192 and 223.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
authorize changes to the facility's Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63971). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendment: April 23, 1997.
    Brief description of amendment: This amendment changes the 
Technical Specifications Surveillance Requirements for TS 4.3.2.1.1.a, 
4.3.2.1.4.b, 4.3.2.1.10.a, 4.3.2.1.10.b, and 4.7.3.b.3. to provide more 
specific information about the tests performed and the components 
tested.
    Date of issuance: March 18, 1998.
    Effective date: March 18, 1998.
    Amendment No.: 76.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33119).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 18, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendment: September 29, 1997 (NRC-97-
0089), as supplemented on March 10, 1998 (NRC-98-0036).
    Brief description of amendment: The amendment revises the technical 
specifications by relocating the requirements for selected

[[Page 17240]]

instrumentation and the associated Bases from the technical 
specifications (TS) to the updated final safety analysis report. The 
affected instrumentation is seismic monitoring (TS 3.7.2), 
meteorological monitoring (TS 3.7.3), the traversing in-core probe 
system (TS 3.7.7), the chlorine detection system (TS 3.7.8), and the 
loose-parts detection system (TS 3.7.10). The TS index and list of 
tables are also revised to reflect the relocation of these TS and 
associated Bases. NRC Generic Letter 95-10, ``Relocation of Selected 
Technical Specification Requirements Related to Instrumentation,'' 
dated December 15, 1995, provided information concerning relocation of 
the requirements for these instruments.
    Date of issuance: March 17, 1998.
    Effective date: March 17, 1998, with full implementation within 90 
days.
    Amendment No.: 115.
    Facility Operating License No. NPF-43: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54870). The March 10, 1998, supplement requested a change in the 
implementation period and was not outside the scope of the initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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

    Date of amendment request: May 24, 1997.
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 3/4.7.4, Ultimate Heat Sink, Table 3.7-3, by 
incorporating more restrictive dry cooling tower fan requirements, and 
changes the wet cooling tower water consumption in the TS Bases.
    This amendment modifies the TS to be consistent with revised 
design-basis calculations.
    Date of issuance: March 23, 1998.
    Effective date: March 23, 1998, to be implemented within 60 days.
    Amendment No.: 139.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33123).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date application for amendment: July 31, 1997.
    Brief description of amendment: This amendment changes Action 
Statement 36 to TS Table 3.3.3-1, ``Emergency Core Cooling System 
Actuation Instrumentation,'' to include actions to be taken if more 
than one channel per trip function should be inoperable in the high-
pressure core spray drywell pressure and reactor water level 
instrumentation.
    Date of issuance: March 16, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 79.
    Facility Operating License No. DPR-63: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 27, 1997 (62 FR 
45460).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 16, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendment request: October 16, 1996.
    Description of amendment request: The amendment revises the 
Technical Specifications (TSs) relating to the requirements for AC 
power sources. The amendment changes certain requirements stated in TS 
3/4.8.1, ``AC Sources.'' The requirements are related to the emergency 
diesel generators.
    Date of issuance: March 17, 1998.
    Effective date: As of the date of issuance, with full 
implementation within 60 days.
    Amendment No.: 54.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66711).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of amendment request: February 12, 1997.
    Description of amendment request: The amendment modifies Technical 
Specification (TS) Section 6.0 ``Administrative Controls,'' to reflect 
recent organizational changes and changes to the approval title for the 
Station Qualified Reviewer Program and corrects an incorrect reference 
in TS 6.4.3.9.b.
    Date of issuance: March 26, 1998.
    Effective date: As of its date of issuance, to be implemented 
within 60 days.
    Amendment No.: 55.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27797).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station, 
Unit No. 1, Washington County, Nebraska

    Date of amendment request: July 25, 1997, as supplemented by 
letters dated November 21, 1997, and March 3, 1998.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 3.5(2), 3.5(3) through 3.5(7), 5.19 and associated 
Basis to implement Option B of 10 CFR 50 Appendix J.
    Date of issuance: March 23, 1998.
    Effective date: March 23, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 185.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59919).

[[Page 17241]]

    The November 21, 1997, and March 3, 1998, supplemental letters 
provided additional clarifying information that did not change the 
original no significant hazards determination consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102

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

    Date of application for amendments: February 26, 1997, as 
supplemented by letters dated December 23, 1997, January 30, 1998, and 
February 9, 1998.
    Brief description of amendments: The amendments revised the 
combined Technical Specifications (TS) for the Diablo Canyon Power 
Plant (DCPP) Unit Nos. 1 and 2 to change TS 3/4.4.5 and 3.4.6.2, 
including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the 
implementation of steam generator (SG) tube voltage based repair 
criteria for outside diameter stress corrosion cracking (ODSCC) 
indications at tube-to-tube support plant (TSP) intersections. The 
allowed primary-to-secondary operational leakage from any one SG would 
be reduced from 500 gpd to 150 gpd.
    Date of issuance: March 12, 1998.
    Effective date: March 12, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment Nos.: Unit 1-124; Unit 2-122.
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 4, 1997 (62 FR 
17239).
    The December 23, 1997, January 30, 1998, and February 9, 1998, 
supplemental letters provided additional clarifying information and did 
not change the staff's initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendments: March 4, 1997.
    Brief description of amendments: These amendments revise the 
emergency core cooling system surveillance test acceptance criteria in 
Technical Specification 3/4.5.2 for the centrifugal charging and safety 
injection pumps. Specifically, the change would reduce the maximum 
specified flow rate values for system alignments that affect the 
suction pressure to the pumps. In the recirculation mode, increased 
system flow occurs when the charging and safety injection pumps take 
suction from the discharge of the residual heat removal pumps.
    Date of issuance: March 12, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos: 208 and 189.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19834).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 12, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendment: October 29, 1997, as 
supplemented on January 27, 1998.
    Brief description of amendment: The amendment provides a one-time 
change to Technical Specification 3/4.4.6, ``Steam Generators,'' to 
require that the next inspection be performed within 24 months from 
initial criticality for fuel cycle 10, or during the next refueling 
outage, whichever is first for fuel cycle 10. In addition, the 
amendment eliminates a description of an alternate steam generator tube 
sampling plan that was applicable only during the fourth refueling 
outage.
    Date of issuance: March 19, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No: 190.
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66142).
    The January 27, 1998, supplemental letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 19, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendments: November 4, 1997.
    Brief description of amendments: These amendments revise the 
containment systems surveillance test acceptance criteria in Technical 
Specification 3/4.6.2 for the containment spray pumps. Specifically, 
the change would replace the Salem Unit 2 minimum specified discharge 
pressure requirement with an acceptance criterion based on pump 
differential pressure, and add this surveillance as a new requirement 
on Salem Unit 1.
    Date of issuance: March 24, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days Amendment Nos.: 209 and 191.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66141).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

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

    Date of application for amendment: February 9, 1998.
    Brief description of amendment: The amendment revises the Virgil C. 
Summer Nuclear Station Technical Specifications (TS) to remove

[[Page 17242]]

emergency diesel generator (1) accelerated testing requirements (TS 3/
4.8.1, Table 4.8-1), and (2) special reporting requirements (TS 
Surveillance Requirement 4.8.1.1.3) in accordance with NRC Generic 
Letter (GL) 94-01, ``Removal of Accelerated Testing and Special 
Reporting Requirements for Emergency Diesel Generators.''
    Date of issuance: March 30, 1998.
    Effective date: March 30, 1998.
    Amendment No.: 139.
    Facility Operating License No. NPF-12: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9614) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.

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

    Date of application for amendment: April 24, 1997, as supplemented 
by letters dated June 6, 1997, and June 27, 1997.
    Brief description of amendment: The amendment revises Section 6.0 
of the Callaway Plant, Unit 1 Technical Specifications to change the 
title ``Senior Vice President Nuclear'' to ``Vice President and Chief 
Nuclear Officer.''
    Date of issuance: March 23, 1998.
    Effective date: March 23, 1998.
    Amendment No.: 122.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40859).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 23, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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

    Date of application for amendment: October 10, 1997, as 
supplemented on October 31, 1997.
    Brief description of amendment: The amendment revises and clarifies 
the offsite power requirements.
    Date of Issuance: March 24, 1998.
    Effective date: March 24, 1998, to be implemented within 60 days.
    Amendment No.: 155.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68319).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendments: January 21, 1997, as 
supplemented on December 15, 1997.
    Brief description of amendments: These amendments revise TS Section 
15.6.11, ``Radiation Protection Program,'' references to Title 10, Code 
of Federal Regulations, Part 20.
    Date of issuance: March 17, 1998.
    Effective date: March 17, 1998, with full implementation within 45 
days.
    Amendment Nos.: 182 and 186.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19837) The December 15, 1997, supplement provided clarifying 
information and modified proposed language within the scope of the 
original application and did not change the staff's initial proposed no 
significant hazards considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 17, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

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

    Date of application for amendments: November 17, 1995 (TSCR 182), 
as supplemented on July 29, 1996, and December 15, 1997.
    Brief description of amendments: These amendments revise Technical 
Specifications 15.6.3.2, 15.6.3.3, and 15.6.5 designation of health 
physics manager to health physicist.
    Date of issuance: March 24, 1998.
    Effective date: March 24, 1998, with full implementation within 45 
days.
    Amendment Nos.: 183 and 187.
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47983).
    The December 15, 1997, letter provided additional clarifying 
information within the scope of the original application and did not 
change the staff's initial proposed no significant hazards 
considerations determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241.

    Dated at Rockville, Maryland, this 1st day of April 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-9040 Filed 4-7-98; 8:45 am]
BILLING CODE 7590-01-P