[Federal Register Volume 62, Number 242 (Wednesday, December 17, 1997)]
[Notices]
[Pages 66133-66151]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-32763]


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NUCLEAR REGULATORY COMMISSION


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

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 November 21, 1997, through December 5, 1997. 
The last biweekly notice was published on December 3, 1997 (62 FR 
63970).

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 Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By January 16, 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

[[Page 66134]]

document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: 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.

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

    Date of amendment request: February 28, 1997.
    Description of amendment request: The proposed amendments would 
revise Byron and Braidwood Technical Specifications (TS) Sections 3/
4.4.5, ``Steam Generators,'' and 3/4.4.8, ``Reactor Coolant System 
Specific Activity,'' for both the Byron Station, Units 1 and 2, and the 
Braidwood Station, Units 1 and 2. The intent of these proposed 
revisions is to restore for both Byron, Unit 1, and Braidwood, Unit 1, 
the original TS related to steam generator (SG) inspections and the 
primary coolant dose equivalent iodine-131 (DEI) concentrations. These 
amendments will become effective when the original steam generators 
(OSG) which are Westinghouse Model D4 SGs, are removed and the 
replacement steam generators (RSG) made by Babcock and Wilcox, 
International (BWI), are installed. The RSGs are presently being 
installed at Byron, Unit 1, while the RSGs will be installed at 
Braidwood, Unit 1, in fall 1998.
    The SG inspection methodology, inspection frequency, reporting 
requirements and acceptance criteria for the RSGs in both Byron, Unit 
1, and Braidwood, Unit 1, will revert to the TSs for the OSGs before 
several prior license amendments incorporated into the TSs: (1) The 
interim plugging criteria (IPC) consistent with Generic Letter (GL) 95-
05; (2) the F* criteria for the SG tube expansions into the tubesheet; 
and (3) the criteria for repairing SG tubes using either Westinghouse 
laser welded sleeves or Combustion Engineering tungsten inert gas (TIG) 
welded sleeves. The TSs applicable to Byron, Unit 2, and Braidwood, 
Unit 2, both of which have Westinghouse Model D5 SGs, remain unchanged 
except for designating them in the TSs as model D5 SGs.
    With respect to the limiting value of the DEI primary coolant 
concentration, both the Byron, Unit 1, TSs and the Braidwood, Unit 1, 
TSs will revert from their present TS limit of 0.35 to 1.0 microcuries 
per gram. A license amendment request to lower the Byron, Unit 1, TS 
DEI limit from 0.35 to 0.20 microcuries per gram was submitted on 
January 31, 1997, but this request was

[[Page 66135]]

subsequently withdrawn on November 11, 1997, because the RSGs were 
being installed in the Byron, Unit 1, refueling outage which started in 
early November 1997. A license amendment request to lower the 
Braidwood, Unit 1, TS DEI limit from 0.35 to 0.10 microcuries per gram 
was submitted on September 2, 1997. Action on this request is still 
pending but in any case, will not affect the subject license amendment 
request for Braidwood, Unit 1, because the September 2, 1997, request 
is only applicable to the OSGs which are presently using the IPC that 
were originally incorporated into the TSs on November 9, 1995. The 
applicable bases sections of the Byron, Unit 1, TSs and Braidwood, Unit 
1, TS will also be revised to reflect the TS changes discussed above.
    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.
    Due to design differences between the replacement Steam 
Generators (RSGs) and OSGs, the analyses supporting the application 
of the F* and voltage-based repair criteria do not apply to the 
RSGs. Also, the analyses supporting sleeving repair by the 
Westinghouse laser welded or Combustion Engineering Tungsten Inert 
Gas (TIG) welded sleeving methodologies do not apply to the RSGs due 
to the design differences. The RSG and OSG tube bundle 
configurations are similar, however, the RSG tubes are smaller in 
diameter, constructed of Inconel Alloy 690 instead of Alloy 600, and 
supported by stainless steel lattice grids instead of the drilled 
carbon steel plates used in the OSGs. The RSG tubes are 
hydraulically expanded into the tube sheet during initial assembly. 
The RSG upper tube bundle shape consists of tubes with continuous, 
smooth, long radius bends.
    The structural analysis demonstrates that the tube integrity is 
maintained for a Main Steamline Break (MSLB) occurring during normal 
full power operation. The structural evaluation of the tubing for 
faulted conditions was performed in accordance with the ASME Boiler 
and Pressure Vessel Code Section III requirements. The tube material 
selection and size exceed the strength requirements of the existing 
steam generators. Comparison of the Alloy 690 tube material used in 
the RSGs with the Alloy 600 tube material in the OSGs show that the 
RSG material strength characteristics are as good as or better than 
those of the existing design. A comparison of the stress margins of 
the RSG and OSG show that the stress margin in the RSG tubes exceed 
the stress margin in the OSG tubes.
    RSG portions of the reactor coolant pressure boundary are 
designed to permit periodic inspection and testing of important 
areas and features to assess structural and leak-tight integrity. 
ASME Section XI, provides the depth of an allowable outside diameter 
(O.D.) flaw for tubes in service. The RSG has tubing fabricated from 
SB-163 material (Inconel Alloy 690) which is examined by eddy 
current methods to the requirements of ASME Section III, NB-2550. 
The tubing has a radius to thickness (r/t) ratio less than 8.70. In 
accordance with ASME Section XI, for tubing having an r/t ratio of 
less than 8.70, the depth of an allowable O.D. flaw shall not exceed 
40% of the nominal tube wall thickness.
    The potential for tube rupture is not increased from the OSGs as 
demonstrated in the qualification analysis and testing for the RSGs. 
The program for periodic inservice inspection of the steam 
generators monitors the integrity of the SG tubing to ensure that 
there is sufficient time to take proper and timely corrective action 
if any tube degradation is detected. Therefore, installation of the 
RSGs will not increase the probability of the occurrence of primary-
to-secondary leakage or a steam generator tube rupture (SGTR) during 
normal or accident conditions.
    The design basis doses calculated for postulated accidents 
involving degradation of SG tubes, such as SGTR and MSLB accidents, 
as presented in UFSAR [Updated Final Safety Analysis Report] Chapter 
15 accident analysis have been evaluated and are decreased by 
installation of the RSGs and restoration of the RCS activity limit 
to 1.0 microcuries/gm. The decrease in offsite dose is primarily due 
to the smaller RSG tube diameter and less primary-to-secondary 
transfer during the event. The dose calculations are performed 
consistent with NUREG-0800, ``Standard Review Plan'' and ensure site 
boundary doses are within a small fraction of the Title 10 Code of 
Federal Regulations Part 100 (10 CFR 100) requirements. Therefore, 
the change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    Limiting the applicability of TS provisions to a specific cycle 
or SG type are administrative changes in that they provide 
clarification consistent with current analyses and do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Restricting application of IPC, F* and sleeving methodologies to 
the OSGs and reinstating an RCS activity limit of 1.0 microcuries/gm 
upon installation of the RSGs will not introduce significant or 
adverse changes to the plant design basis that could lead to a new 
or different kind of accident being created. The RSG tubing meets 
the requirements of General Design Criteria (GDC) 14, 15, 30, 31, 
and 32 of 10 CFR 50, Appendix A. The RSG tubing has been designed 
and evaluated consistent with ASME Code Section III criteria and the 
inspection criteria for the RSGs is consistent with ASME Code 
Section XI criteria. The RSGs have thermally treated Inconel Alloy 
690 tubes which are hydraulically expanded into the tube sheet 
during initial assembly. Alloy 690 is more resistant to stress 
corrosion cracking (SCC) than Alloy 600 which is used in the OSG 
tubing. Overall tube bundle structural and leakage integrity is 
maintained at a level consistent with or better than the originally 
supplied tubing during all plant conditions.
    ComEd will continue to apply the TS maximum primary-to-secondary 
leakage limit of 150 gpd (0.1 gpm) through any one SG at Byron and 
Braidwood to help preclude the potential for excessive leakage 
during all plant conditions. The EPRI recommended 150 gpd limit 
provides for leakage detection and plant shutdown in the event of an 
unexpected tube leak and precludes the potential for excessive 
leakage or tube burst in the event of a Main Steam Line Break or 
under Loss of Coolant Accident conditions.
    Limiting the applicability of TS provisions to a specific cycle 
or SG type are administrative changes in that they provide 
clarification consistent with current analyses.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Restricting application of IPC, F*, and sleeving methodologies 
to the OSGs for which the supporting analyses apply, does not 
involve a reduction in a margin of safety. The RSG tubing has been 
shown to retain adequate structural and leakage integrity during 
normal, transient, and postulated accident conditions consistent 
with GDC 14, 15, 30, 31, and 32 of 10 CFR 50 Appendix A. The RSG 
tubing has been designed and evaluated consistent with the margins 
of safety specified in ASME Code Section III. The proposed program 
for periodic inservice inspection of the replacement steam 
generators monitors the integrity of the SG tubing to ensure that 
there is sufficient time to take proper and timely corrective action 
if any tube degradation is present. The proposed program is 
consistent with the Standard Technical Specifications.
    The Unit 1 RCS dose equivalent I-131 limit is being raised upon 
installation of the RSGs to eliminate the compensatory lower limit 
that was adopted in conjunction with IPC for the existing 
Westinghouse D4 SGs. With the RCS activity limit returned to the 
Standard Technical Specification value of 1.0 [mu]Ci/gm, the 
assessment of postulated UFSAR Chapter 15 accidents (including SGTR 
and MSLB) has concluded that the calculated design basis doses 
presented in Chapter 15 are not adversely impacted by the RSGs. This 
ensures that the resulting 2-hour dose rates at the Byron and 
Braidwood site boundaries will not exceed an appropriately small 
fraction of 10 CFR 100 dose guideline values.
    Limiting the applicability of TS provisions to a specific cycle 
or SG type are

[[Page 66136]]

administrative changes in that they provide clarification consistent 
with current analyses.
    Therefore, it is concluded that this change does not involve a 
significant reduction in a margin of safety with respect to plant 
safety as defined in the UFSAR or the Technical Specification.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: March 26, 1997.
    Description of amendment request: The proposed amendment would 
revise the containment system technical specifications (TS) contained 
in TS Sections 3.6 and 4.5. The licensee has classified the changes as 
``More Restrictive,'' ``Less Restrictive,'' and ``Administrative.'' 
``More Restrictive'' changes include reduction of the allowable 
containment pressure, addition of an action statement defining action 
to be taken when the containment pressure limit is exceeded, addition 
of a restriction on containment temperature, and revision of the 
applicable conditions for the containment purge valves to require that 
the valves be operable above 210 degrees F versus the current 
requirement that they be operable above 525 degrees F. ``Less 
Restrictive'' changes include addition of an allowance to enter an air 
lock through a locked door to perform maintenance, addition of an 
allowance to open containment isolation valves under administrative 
control, revision of the applicable conditions for containment pressure 
to exclude the cold shutdown operating condition, and addition of an 
exception to the surveillance requirement requiring verification of the 
status of ``locked-closed'' manual isolation valves after a refueling 
outage to exclude requiring such verification for valves opened under 
administrative control. ``Administrative'' changes include the deletion 
of containment isolation valve tables and component identifiers from 
the TS in accordance with Generic Letter 91-08 (``Removal of Component 
Lists from Technical Specifications'') and editorial restructuring of 
the affected TS sections to clarify the remaining requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Each proposed change has been classified as ``Administrative,'' 
``More Restrictive,'' or ``Less Restrictive.'' ``Administrative'' 
and ``More Restrictive'' changes are discussed generically; ``Less 
Restrictive'' changes are discussed individually.
    Five of the proposed changes are classified as being ``Less 
Restrictive'':
    (G.1) Allowance in LCO [Limiting Condition for Operation] 3.6.1 
to enter an air lock to perform maintenance.
    (G.2) Allowance in LCO 3.6.1 to open containment isolation 
valves under administrative control.
    (I.2) Revising the applicable conditions of LCO 3.6.2, 
Containment Pressure to exclude Cold Shutdown.
    (J.2) Exception in SR [Surveillance Requirement] 4.5.3d for 
valves opened under administrative control as allowed by LCO 3.6.1.
    (P) Allowance in SR 4.5.2 to enter an air lock to perform 
maintenance.
    Four of the proposed changes are classified as being ``More 
Restrictive'':
    (I.1) Revising LCO 3.6.2 to reduce the allowable containment 
pressure.
    (I.3) Addition of an action statement to LCO 3.6.2, Containment 
Pressure.
    (K) Addition of a new LCO which restricts Containment 
Temperature.
    (M.2) Revising the applicable conditions for LCO 3.6.5, Purge 
Valves.
    The remaining changes are all classified as being 
``Administrative''.
    Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P 
allow limited access through the operable door of an air lock when 
the other door is inoperable; current Technical Specifications [TS] 
do not. Proposed changes G.2 and J.2 allow unisolating containment 
penetration flow paths intermittently under administrative control; 
current TS do provide a similar allowance, but only for one specific 
penetration. These changes cannot significantly increase the 
probability of an accident because opening an air lock door or a 
containment penetration is not, itself, an initiator and does not 
affect the items which are initiators of any analyzed accident.
    The ability to open the operable door or to open a containment 
penetration, even if it means the containment boundary is 
temporarily not intact, does not significantly increase the 
consequences of an accident previously evaluated because of the low 
probability of an event that could pressurize the containment 
occurring during the short time the operable door or containment 
penetration is expected to be open. In a case where containment 
integrity (or containment operability) is lost due to excessive 
leakage, both the Palisades Technical Specifications and the 
Standard Technical Specifications [STS] allow one hour of continued 
operation for its restoration. That time period is allowed without 
regard to the magnitude of the potential leakage, and would be 
allowed even if both personnel air lock doors [were] leaking 
excessively. The additional allowance of permitting the operable 
door to be opened momentarily for entry or egress when the other 
door is inoperable due to excessive leakage would not significantly 
add to the probability of containment leakage and the resultant 
consequences of an accident. Similarly, the allowance to open any 
containment penetration intermittently under administrative control, 
which currently is allowed for one penetration, would not 
significantly add to the probability of containment leakage and the 
resultant consequences of an accident.
    Therefore, operation of the Facility in accordance with proposed 
changes G.1, G.2, J.2, and P would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment 
Pressure so that it no longer applies during Cold Shutdown. LCO 
3.6.2 is intended to limit containment pressure to that value used 
as an initial condition in the safety analysis. Containment pressure 
is an initial condition in analyses which assure that containment 
internal pressure will not exceed the containment design values 
during a LOCA or MSLB. Containment pressure is not an initiator of 
any accident previously evaluated. Neither a LOCA [loss-of-coolant 
accident] nor a MSLB [main steam line break] occurring during Cold 
Shutdown would pressurize the containment. Therefore, a containment 
pressure LCO is not necessary, during Cold Shutdown, to assure that 
containment design pressure and temperature is not exceeded. The STS 
Containment pressure LCO is not applicable in Cold Shutdown.
    Therefore, operation of the Facility in accordance with proposed 
change I.2 would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    3. More Restrictive Changes: ``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.
    4. ``Administrative'' changes make wording changes which clarify 
existing TS requirements, without affecting their

[[Page 66137]]

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.
    Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P 
allow limited access through the operable door of an air lock when 
the other door is inoperable; current Technical Specifications do 
not. Proposed changes G.2 and J.2 allow unisolating containment 
penetration flow paths intermittently under administrative control; 
current TS do provide a similar allowance, but only for one specific 
penetration. Opening an air lock door or a containment penetration 
does not affect the operating conditions or operation of any plant 
systems (other than the containment); it does not create a threat to 
the integrity of any operating system or alter any system operating 
practice or settings.
    Since the opening of an air lock door or a containment 
penetration only affects the potential leakage from the containment, 
and does not affect any of the operating plant systems, operation of 
the Facility in accordance with the proposed Technical 
Specifications change would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment 
Pressure so that it no longer applies during Cold Shutdown. LCO 
3.6.2 is intended to limit containment pressure to that value used 
as an initial condition in the safety analysis. Containment pressure 
is an initial condition in analyses which assure that containment 
internal pressure will not exceed the containment design values 
during a LOCA or MSLB. Neither a LOCA nor a MSLB occurring during 
Cold Shutdown would pressurize the containment. Therefore, a 
containment pressure LCO is not necessary, during Cold Shutdown, to 
avoid creation of a new or different kind of accident. The STS 
Containment pressure LCO is not applicable in Cold Shutdown.
    Therefore, operation of the Facility in accordance with proposed 
change I.2 would not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. More Restrictive Changes: ``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.
    4. ``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.
    Do the proposed changes involve a significant reduction in a 
margin of safety?
    1. Changes G.1, G.2, J.2, and P: Proposed changes G.1 and P 
allow limited access through the operable door of an air lock when 
the other door is inoperable; current Technical Specifications do 
not. Proposed changes G.2 and J.2 allow unisolating containment 
penetration flow paths intermittently under administrative control; 
current TS do provide a similar allowance, but only for one specific 
penetration. The ability to open the operable door or a containment 
penetration, even if it means the containment boundary is 
temporarily not intact, does not involve a significant reduction in 
a margin of safety because of the low probability of an event that 
could pressurize the containment occurring during the short time the 
operable door or penetration is expected to be open.
    Therefore, operation of the Facility in accordance with the 
proposed Technical Specifications change would not involve a 
significant reduction in a margin of safety.
    2. Change I.2: Change I.2 alters existing LCO 3.6.2, Containment 
Pressure so that it no longer applies during Cold Shutdown. LCO 
3.6.2 is intended to limit containment pressure to that value used 
as an initial condition in the safety analysis. Containment pressure 
is an initial condition in analyses which assure that containment 
internal pressure will not exceed the containment design values 
during a LOCA or MSLB. Neither a LOCA nor a MSLB occurring during 
Cold Shutdown would pressurize the containment. Therefore, 
elimination of a Cold Shutdown LCO for containment pressure would 
not affect the post-accident pressure or temperature. Since peak 
post accident [pressure] and temperature would be unaffected by the 
proposed change, operation of the Facility in accordance with 
proposed change I.2 would not involve a significant reduction in a 
margin of safety.
    3. More Restrictive Changes: ``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.
    4. ``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.

    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: John N. Hannon.

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

    Date of amendments request: October 29, 1997.
    Description of amendments request: The proposed amendments to the 
Technical Specifications (TS) for the Brunswick Steam Electric Plant 
(BSEP) Units 1 and 2 would revise the description of the control rod 
assemblies (CRAs) in TS 5.3.2. The proposed revision was requested to 
support replacement of a portion of the BSEP Unit 1 CRAs during that 
unit's next refueling outage with assemblies of a different design. 
Carolina Power & Light Company, the licensee, has proposed adopting the 
description of CRAs used in NUREG-1433, Revision 1, ``Standard 
Technical Specifications General Electric Plants, BWR/4,'' which 
includes the number and shape of CRAs and a stipulation that NRC-
approved absorber material be used in CRAs. The more detailed 
description in the current TS of CRAs would be relocated to the Updated 
Final Safety Analysis Report. The licensee has stated that the CRA 
description proposed for TS 5.3.2 will be sufficient to ensure that any 
future changes in CRA design that may affect safety will require prior 
NRC review and approval.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Relocation of the control rod assembly descriptive information 
from the Technical Specifications to the Updated Final Safety 
Analysis Report will ensure that adequate control of the information 
is maintained. Any changes to this design information must conform 
with the requirements of 10 CFR 50.59. Restricting use of control 
rod assembly absorber materials to those listed, or to materials 
that have been approved by the NRC, will ensure any changes which 
may affect safety to require prior NRC review and approval. Since 
the information with a potential to affect safety is sufficiently 
addressed by the Technical Specifications, the criteria of 10 CFR 
50.36(c)(4) for

[[Page 66138]]

including the relocated information as Design Features are not met. 
Because the relocated information is not required to be in the 
Technical Specifications to provide adequate protection of the 
public health and safety, relocation of control rod assembly 
descriptive information will not increase either the probability or 
the consequences of an accident previously evaluated.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Relocation, to the Updated Final Safety Analysis Report, of the 
information pertaining to the control rod assembly designs ensures 
that adequate control of the information will be maintained. Since 
the information with a potential to affect safety is sufficiently 
addressed by the Technical Specifications, the criteria of 10 CFR 
50.36(c)(4) for including the relocated information as Design 
Features are not met. Because the relocated information is not 
required to be in the Technical Specifications to provide adequate 
protection of the public health and safety, the proposed Technical 
Specification changes to relocate the control rod assembly design 
information to the Updated Final Safety Analysis Report does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    As discussed in Items 1 and 2 above, relocation of the control 
rod assembly descriptive information from the Technical 
Specifications to the Updated Final Safety Analysis Report will 
ensure that adequate control of the information is maintained. Any 
changes to this design information must conform with the 
requirements of 10 CFR 50.59. Restricting use of control rod 
assembly absorber materials to those listed, or to materials that 
have been approved by the NRC, will ensure any changes which may 
affect safety to require prior NRC review and approval. The 
information with a potential to affect safety is sufficiently 
addressed by the Technical Specifications, therefore, the proposed 
Technical Specification changes to relocate control rod assembly 
design information to the Updated Final Safety Analysis Report 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297
    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: James E. Lyons.

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

    Date of amendment request: July 16, 1997, as supplemented October 
30, 1997.
    Description of amendment request: The amendment would update 
License condition 2.C(4) to reflect the latest revision levels of the 
Oyster Creek Security Training and Qualification Plan, License 
Amendment Request No. 252.
    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:

    GPU Nuclear has concluded that the proposed changes to the 
Security Plan do not involve a significant hazard consideration. In 
support of this determination, an evaluation of each of the three 
standards set forth in 10 CFR 50.92 is provided below.
    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Security Plan provisions are not associated with design basis 
accident initiators nor do they constitute part of any mitigation 
system. Therefore, the probability and consequences of accidents are 
not increased.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The Security Plan changes do not create new or change existing 
physical interfaces with plant equipment. Therefore, the changes do 
not create the possibility of a new or different kind of accident.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety.
    Margins associated with reactor and fuel storage nuclear safety 
are not affected by the proposed Security Plan changes since neither 
physical nor procedural changes to associated systems, structures 
and components are involved. Vital area security measures, which are 
reduced, are compensated by commitments to hold contingency drills 
at a frequency sufficient to maintain response capability for 
response personnel and to use organic-type X-ray equipment.

    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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Ronald B. Eaton, Acting Director.

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

    Date of amendment request: November 14, 1997.
    Description of amendment request: The proposed change to Technical 
Specification 4.5.2.d.1 will clarify the wording and increase the 
setpoint for the open pressure interlock (OPI).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    NNECO has reviewed the proposed revision in accordance with 10 
CFR 50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10 CFR 50.92(c) are not 
satisfied. The proposed revision does not involve [an] SHC because 
the revision would not:
    1. Involve a significant increase in the probability or 
consequence of an accident previously evaluated.
    Increasing the Technical Specification Open Pressure Interlock 
(OPI) pressure to 412.5 psia [pounds per square inch--atmospheric] 
will still maintain the required function of preventing the MOVs 
[motor operated valves] from opening inadvertently. The increased 
pressure is within the design limits of the RHR [residual heat 
removal] piping system and components. The pressure signal is 
generated from a transmitter and results in an electronic input to 
the bistable. This is a clarification of the conditions under which 
the OPI is tested.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There is no change to the function of the OPI. The protection 
provided by the interlock remains intact. The Technical 
Specification OPI pressure has been raised to take into account 
instrument accuracies and reset deadbands. The RHR system design 
pressure remains protected from being exceeded by inadvertent 
opening of the isolation MOVs. The method for the OPI surveillance 
is clarified by clearly stating that the bistable receives a 
simulated transmitter signal representative of the process pressure.

[[Page 66139]]

    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The design pressure of the RHR system is 600 psig [pounds per 
square inch--gauge]. The most limiting case is to prevent the RHR 
pump developed head pressure from exceeding the design pressure when 
aligned to the RCS [reactor coolant system] as suction pressure. RHR 
pump testing has determined that a maximum pump differential 
pressure of 195 psi [pounds per square inch] exists for deadhead/no 
flow conditions. Therefore, to maintain the 600 psig design pressure 
limit, RCS/suction pressure must be limited to 405 psig (420 psia, 
assuming a 15 psi conversion from psig to psia). The proposed 
maximum pressure, including setpoint tolerances and reset deadbands, 
is less than this value; i.e. 412.5 psia. Head corrections due to 
elevation differences are considered to be insignificant.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is 
determined that the proposed revision does not involve an SHC.

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

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

    Date of amendment request: November 3, 1997.
    Description of amendment request: The proposed amendment would 
revise Sections 1, 3.1, 3.3, 4.3, and 6 of Appendix A of the Indian 
Point 3 Technical Specifications. These revisions extend the Heatup-
Cooldown limits from 11 to 13 effective full power years (EFPYs), 
provide the corresponding Overpressure Protection System (OPS) limits, 
relocate the new pressure temperature limit curves and low-temperature 
overpressurization protection (LTOP) system limits to the pressure 
temperature limit report (PTLR) and include some minor revisions which 
ensure specification clarity and conservatism.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously analyzed?
    Response: The proposed license amendment does not involve a 
significant increase in the probability or consequences of a 
previously analyzed accident. The pressure-temperature limit changes 
proposed by this amendment are based on supporting data and 
evaluation methodologies previously submitted to the NRC in 
Reference 3 [see application dated November 3, 1997] and approved as 
Amendments 109 and 121 (References 4 and 5) [see application dated 
November 3, 1997]. These limits are based upon the irradiation 
damage prediction methods of Regulatory Guide 1.99, Revision 2. The 
LTOPs changes contained in this submittal have been conservatively 
adjusted in accordance with the new pressure-temperature limits, in 
accordance with the methodology contained in Reference 3 and ASME 
Code Case N-514.
    The relocation of the pressure-temperature and LTOPs limits from 
the Technical Specifications to the PTLR does not eliminate the 
requirement to operate in accordance with the limits specified in 10 
CFR [Part] 50, Appendix G. The requirement to operate within the 
limits in the PTLR is specified in and controlled by the Technical 
Specifications.
    The revised version of Section 3.1.A.8 clarifies existing 
requirements related to the OPS system and adds an eight hour 
completion time for compensating actions, consistent with the STS 
[standard technical specifications]. The changes to Section 
3.1.A.1.h, i, and j revise the requirements associated with the 
start of an RCP [reactor coolant pump]. These changes improve 
specification clarity and do not increase the probability or 
consequences of an accident.
    The Technical Specification changes associated with the 
restriction on SI [safety injection] pumps provides added 
conservatism to the Technical Specifications and limits the 
likelihood of an RHR [residual heat removal] overpressurization 
event. Current plant procedures prohibit actuation of any SI pumps 
when RHR is in service, except during testing, loss of RHR cooling, 
or reduced inventory operations. Therefore, the change to the 
Technical Specifications will not alter current plant operation.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response: The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously analyzed. The pressure-temperature limits are updating 
the existing limits by taking into account the effects of radiation 
embrittlement, utilizing criteria defined in Regulatory Guide 1.99, 
Revision 2, and extending the effective period to 13 EFPYs. The 
updated OPS limits have been adjusted to account for the effect of 
irradiation on the limiting reactor vessel material. These changes 
do not affect the way the pressure-temperature or OPS limits provide 
plant protection and no physical plant alterations are necessary. 
The relocation of the pressure-temperature and OPS limits from the 
Technical Specifications to the PTLR does not alter the requirements 
associated with these limits.
    The revisions to Section 3.1.A.8 concerning the OPS system 
improve on the clarity of existing specifications and add a 
completion time for compensating actions that is consistent with the 
STS. These changes do not involve any hardware modifications and do 
not affect the function of the OPS system.
    The revisions concerning the operation of SI pumps bring the 
Technical Specifications into line with current operating 
procedures. The changes to Specification 3.1.A.1.h, i, and j provide 
specification clarity and are more conservative than existing 
Technical Specifications. Therefore, the changes cannot create the 
possibility of a new or different kind of accident.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: The proposed amendment does not involve a significant 
reduction in a margin of safety. The margins of safety against 
fracture provided by the pressure-temperature limits are those 
limits specified in 10 CFR Part 50, Appendix G and ASME Boiler and 
Pressure Vessel Code Section XI, Appendix G. The guidance in these 
documents has been utilized to develop the pressure-temperature 
limits with the requisite margins of safety for the heatup and 
cooldown conditions. The new LTOP limits are based upon Reference 3 
and ASME Code Case N-514. The relocation of the pressure-temperature 
and OPS limits to the PTLR does not alter the requirements 
associated with these limits.
    The revisions to Section 3.1.A.8 clarify the requirements 
associated with the OPS system. The revisions associated with the 
operation of SI pumps with RHR in service (Sections 3.3.A.8, 9 and 
10) and the changes regarding RCP starts (Section 3.1.A.1.h, i, and 
j) are more conservative than the current Technical Specifications, 
and are consistent with plant operating procedures. Therefore, they 
do not reduce a margin of safety.

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

[[Page 66140]]

100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa.

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

    Date of amendment request: October 24, 1997.
    Description of amendment request: The amendments would increase the 
containment hydrogen analyzer surveillance frequency in Technical 
Specification 4.6.4.1 from once per refueling outage to quarterly.
    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 containment hydrogen analyzers provide control room 
indication of hydrogen concentration in the containment atmosphere. 
They do not affect the probability of any previously evaluated 
accident. The proposed change would increase the calibration 
frequency specified in TS 4.6.4.1 to make it consistent with 
manufacturer's recommendations and the current calibration frequency 
at [Salem Generating Station] SGS as imposed by administrative 
controls. The change in TS-required calibration frequency is in the 
conservative (more frequent) direction, to ensure that potential 
degradation of the sensor electrolyte over time would not result in 
unacceptable performance of the hydrogen analyzers. The change in 
specified frequency would not adversely affect the consequences of 
any previously evaluated accident.
    2. Proposed change does not create the possibility of a new or 
different kind of accident from any accident previously analyzed 
[evaluated].
    The proposed change affects only the specified calibration 
frequency of the containment hydrogen analyzers. The proposed change 
does not affect the design of any SGS structure, system or 
component, nor would it result in any new plant configuration. 
Therefore, it does not create the possibility of a new or different 
kind of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to the containment hydrogen analyzer 
calibration frequency does not affect the design or operating limits 
of any SGS structure, system or component. The change would make the 
specified calibration frequency more conservative, to ensure the 
hydrogen analyzers perform as designed over time. 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: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: October 24, 1997.
    Description of amendment request: The amendments would revise 
Technical Specification (TS) 3/4.7.7, ``Auxiliary Building Exhaust Air 
Filtration System.'' The revisions would: (1) Require both Auxiliary 
Building Ventilation (ABVS) supply fans to be operable, (2) require all 
three ABVS exhaust fans to be operable, (3) align ABVS TSs to be 
consistent with current TS bases and recently revised system 
descriptions in the Salem Updated Final Safety Analysis Report (UFSAR), 
(4) assure that negative pressure is maintained in the Auxiliary 
Building under all postulated single active failures, (5) clarify 
required Engineered Safety Feature filter testing, (6) provide 
consistency between Unit 1 and Unit 2 TSs, and (7) for Unit 2 only, 
remove the requirement to verify safety injection auto-start 
capabilities.
    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 alters the number of fans which must be 
OPERABLE to ensure that a sufficient number of supply and exhaust 
fans will be operable, following a most limiting single failure, to 
mitigate the consequences of design basis accidents. The changes to 
the ABVS surveillance requirements still provide an appropriate 
means for demonstrating the operability of the ABVS.
    The ABVS cannot initiate or otherwise cause any accident or 
operational transient evaluated in the UFSAR. Consequently, the 
probability of such events is not increased. The ABVS cannot 
increase the consequences of a design basis LOCA unless: (1) 
Auxiliary Building negative pressure is lost, resulting in 
uncontrolled, ground level release of radioactive material; (2) ABVS 
carbon adsorbers are bypassed, resulting in uncontrolled release of 
radioactive iodine from the plant vent; or (3) Auxiliary Building 
temperatures are not controlled, resulting in failure of accident 
mitigating equipment.
    By requiring OPERABILITY of all ABVS supply and exhaust fans, 
the proposed changes contained in this submittal assures Auxiliary 
Building negative pressure is maintained under all postulated post-
accident, single-failure scenarios. The proposed changes to ABVS 
will not affect the elemental iodine adsorption capability of the 
system. Finally, engineering analyses conclude that these fan 
combinations, with single-active failures of the fans or their 
support systems considered, provide sufficient Auxiliary Building 
ventilation. Under the most limiting temperature conditions, the 
fans will maintain room temperatures within design limits. 
Accordingly, the consequences of a design basis LOCA, hence 
applicable design basis accidents or operational transients, are not 
increased.
    Therefore, the proposed amendment 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.
    ABVS supply fans are not considered essential to the primary 
safety-function of preventing or mitigating radioactive releases, 
nor are they currently required to be OPERABLE. Similarly, accident 
analyses take no credit for operation of supply fans. Accordingly, 
malfunctions of vital buses and ABVS exhaust fans are the only 
malfunctions of active ABVS related equipment important to safety 
that are previously evaluated.
    The probability of failure of a vital bus is not increased by 
this proposal since the proposal has no direct effect on electrical 
power. Neither is the probability of exhaust fan failure increased 
by the proposal, since exhaust fans are not affected by this 
proposal, except that the number that must be OPERABLE is increased 
from two to three.
    By requiring additional supply fans and exhaust fans to be 
OPERABLE, no single failure of either a vital bus or ABVS fan 
prevents (1) maintenance of negative Auxiliary Building pressure or 
(2) maintenance of temperatures within design limits. Since ABVS 
supply and exhaust fans cannot initiate accidents, increasing the 
number of fans required to be OPERABLE cannot create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. In addition, the proposed changes to the ABVS 
surveillance testing concern ABVS leakage, HEPA filter and carbon 
adsorber capabilities, and laboratory test methods. Therefore, the 
proposed surveillance requirement changes would have no impact on 
the initiation of accidents.
    Thus, the proposed changes do not create the possibility of a 
new or different kind of

[[Page 66141]]

accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety is dependent upon the maintenance of 
specific operating parameters within designated design limits. Since 
iodine removal capability is not affected by the proposed changes, 
and negative Auxiliary Building pressure and temperatures will 
continue to be maintained within existing design limits under post-
accident conditions, including consideration of the most limiting 
single active failure, the margin of safety is not reduced. By 
imposing new restrictions on the allowed outage times of ABVS 
components, the margin of safety is increased with the proposed 
changes to the ABVS Technical Specification Limiting Condition for 
Operation (LCO).

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: November 4, 1997.
    Description of amendment request: The amendments would change 
Technical Specification (TS) 3/4.6.2, ``Containment Spray System,'' to 
verify on recirculation flow that the containment spray pumps develop a 
differential pressure of at least 204 psi.
    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 revises the CS [containment spray] pump 
technical specification surveillance test acceptance from pump 
discharge pressure to pump differential pressure. This will account 
for the effect of RWST [refueling water storage tank] level on test 
results and provide acceptance criteria that verifies each CS pump 
performs as assumed in the accident analyses. This surveillance test 
is also being added to the Salem Unit 1 TS. The proposed change does 
not alter the physical plant arrangement or the method of CS pump 
inservice testing. Therefore it does not increase the probability of 
an accident. There is no change to pump performance requirements as 
assumed in the accident analyses. There is no change to CS system 
performance in response to an accident. Therefore, the proposed 
change does not involve an increase in the consequences of an 
accident previously evaluated.
    The proposed change also corrects a typographical error by 
removing a repeated word. This change does not involve an increase 
in the consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the Salem Unit 2 CS pump 
surveillance test acceptance criteria from pump discharge pressure 
to pump differential pressure. This will account for the effect of 
RWST level on test results and provide acceptance criteria that 
verify the CS pumps perform as assumed in the accident analyses. 
This surveillance test is also being added to the Salem Unit 1 TS. 
The proposed change does not alter the plant configuration. The 
change does not alter the method of performing inservice testing on 
the CS pumps. The change does not alter the CS pump performance 
assumed in the accident analyses. Therefore, the change will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    The proposed change also corrects a typographical error by 
removing a repeated word. This change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change ensures the CS pump Salem Unit 2 TS 
surveillance test acceptance criteria verify CS pump performance as 
assumed in the accident analyses accounting for RWST level effects. 
This surveillance test is also being added to the Salem Unit 1 TS. 
The proposal does not change the CS pump performance requirements 
assumed in the accident analyses and thus does not reduce the margin 
of safety.
    The proposed change also corrects a typographical error by 
removing a repeated word. This does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: November 14, 1997.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) include administrative and editorial 
changes to correct errors in the TSs that have either existed since 
initial issuance or were introduced during subsequent changes. In 
addition, surveillance requirements are added that are considered 
administrative changes since the surveillances should have been 
incorporated with the TS when the applicable amendment to the TSs was 
approved by the NRC.
    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 to the TS are administrative or editorial 
changes to the TS and do not involve any physical changes to the 
plant. The administrative changes and editorial changes do not 
delete any existing surveillance requirements or delete any 
requirements from the Limiting Condition for Operations (LCOs) or 
Action Statements and therefore do not reduce the actions that are 
currently taken in the TS to demonstrate operability of plant 
structures, systems, or components (SSCs). The additional 
surveillance requirements that are being added to the TS including 
the new surveillances correct past administrative errors and should 
have been incorporated within the TS as part of the approved 
Amendments to the TS. These changes will provide additional 
assurance that SSCs perform their intended safety functions. 
Surveillance testing has been and is currently being performed for 
the surveillance requirements that should have been incorporated and 
are now administratively being added to the TS. Since these changes 
do not modify any SSCs or reduce the current requirements for 
demonstrating operability of these SSCs or reduce the current 
requirements for demonstrating operability of these SSCs, the 
proposed changes to the TS 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.

[[Page 66142]]

    The proposed changes to the TS are administrative and editorial 
corrections to the TS that do not affect the ability of the plant 
systems to meet their current TS requirements or design basis 
functions. There is no reduction in the current surveillance 
requirements required to demonstrate the operability of plant SSCs. 
These changes also do not involve any physical changes to plant 
SSCs. Therefore the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes are administrative and editorial 
corrections to the TS that do not affect the ability of plant SSCs 
to perform their design basis accident functions. There is no 
reduction in the current surveillance requirements required to 
demonstrate the operability of plant SSCs. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: November 14, 1997.
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) include administrative and editorial 
changes to correct errors in the TSs that have either existed since 
initial issuance or were introduced during subsequent changes. In 
addition, surveillance requirements are added that are considered 
administrative changes since the surveillances should have been 
incorporated with the TS when the applicable amendment to the TSs was 
approved by the NRC.
    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 to the TS are administrative or editorial 
changes to the TS and do not involve any physical changes to the 
plant. The administrative changes and editorial changes do not 
delete any existing surveillance requirements or delete any 
requirements from the Limiting Condition for Operations (LCOs) or 
Action Statements and therefore do not reduce the actions that are 
currently taken in the TS to demonstrate operability of plant 
structures, systems, or components (SSCs). The additional 
surveillance requirements that are being added to the TS including 
the new surveillances correct past administrative errors and should 
have been incorporated within the TS as part of the approved 
Amendments to the TS. These changes will provide additional 
assurance that SSCs perform their intended safety functions. 
Surveillance testing has been and is currently being performed for 
the surveillance requirements that should have been incorporated and 
are now administratively being added to the TS. Since these changes 
do not modify any SSCs or reduce the current requirements for 
demonstrating operability of these SSCs or reduce the current 
requirements for demonstrating operability of these SSCs, the 
proposed changes to the TS do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the TS are administrative and editorial 
corrections to the TS that do not affect the ability of the plant 
systems to meet their current TS requirements or design basis 
functions. There is no reduction in the current surveillance 
requirements required to demonstrate the operability of plant SSCs. 
These changes also do not involve any physical changes to plant 
SSCs. Therefore the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes are administrative and editorial 
corrections to the TS that do not affect the ability of plant SSCs 
to perform their design basis accident functions. There is no 
reduction in the current surveillance requirements required to 
demonstrate the operability of plant SSCs. Therefore, the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit-N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: October 29, 1997.
    Description of amendment request: The proposed amendment would make 
a one-time change to Technical Specification 3/4.4.6, ``Steam 
Generators,'' to require that the next inspection be performed within 
24 months of criticality for fuel cycle 10, rather than within 24 
months from the previous inspection. The previous inspection was 
performed in May 1996; thus, adhering to the current Technical 
Specification would require inspection by May 1998 and would require a 
forced outage. It would also eliminate description of an alternate 
sampling plan that was applicable only to Unit 2's fourth refueling 
outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Design Basis Accident (DBA) analyzed in UFSAR Chapter 
15.4.4, is Steam Generator Tube Rupture. The Technical Specification 
steam generator tube inspection attempts to avoid this DBA by 
maintenance of the integrity of the primary to secondary coolant 
boundary represented by steam generator tubes. The process by which 
this integrity is maintained is inspection of steam generator tubes 
at prescribed intervals, and the removal of defective tubes from 
service. Inspection intervals are based on preventing corrosion 
growth from exceeding tube structural strength, thereby preventing 
tube failure. An extensive steam generator inspection in May of 1996 
characterized existing steam generator tube degradation, and 
degraded tubes were removed from service at that time. Degradation 
growth rates were evaluated for the next operating interval and it 
was determined that full cycle operation would not challenge tube 
structural integrity. Because degraded tubes were plugged, the 
integrity of the steam generators has been restored, and, because 
further degradation was prevented by a strictly controlled wet lay-
up program in place since the inspection, steam generator integrity 
has since been maintained at the May 1996 level. This is the level 
normally expected for commencement of full power operations at the 
beginning of a fuel cycle. Thus, it can be reasonably

[[Page 66143]]

concluded that this request to extend the inspection interval to 
conclude 24 months after the start of Unit 2 fuel cycle 10 does not 
involve an increase in the probability of an accident previously 
analyzed.
    Salem UFSAR Chapter 15, Section 15.4.4., discusses the Design 
Basis Accident involving steam generator tube rupture. Since the 
Salem Unit 2 steam generators were extensively inspected and all 
degraded tubes were removed from service by plugging, integrity of 
the generators was restored to fully serviceable condition at that 
time. Degradation of steam generator tubes has been prevented since 
the inspection by a carefully controlled, EPRI Guidelines based, 
corrosion prevention program. It follows, then, that the Unit 2 
steam generators were in the same condition immediately prior to 
fill and vent as if the inspection had just been concluded. This is 
the condition assumed for commencement of normal operation. Thus, it 
is reasonable to conclude that this proposal to extend the current 
steam generator inspection interval to end 24 months after start of 
Unit 2 fuel cycle 10 represents no significant increase in the 
consequences of an accident previously analyzed.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Steam generator tube inspections determine tube integrity and 
provide reasonable assurance that a tube rupture or primary to 
secondary leak will not occur. Accidents involving steam generator 
tube rupture are analyzed in Salem UFSAR Section 15.4.4, Steam 
Generator Tube Rupture. The only type of accident that can be 
postulated from extending the steam generator inspection interval 
would be a tube leak or rupture. Thus, it can be concluded that 
extending the steam generator inspection interval on a one-time 
basis cannot create the possibility of a different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety, as with any TS, depends upon maintenance 
of specific operating parameters within design limits. In the case 
of steam generators, that margin is maintained through assurance of 
tube integrity as the primary to secondary boundary. Assurance of 
tube integrity is provided through periodic inservice testing of 
tube integrity and removal from service of defective tubes. 
Additional margin is provided through protection from possible 
consequences of steam generator tube failure by detection and 
mitigation systems. As discussed in 1., above, there was an 
extensive steam generator inspection, and the steam generators have 
been maintained since the inspection, using a lay-up program that 
complies with EPRI Guidelines, to prevent further tube degradation. 
Also, N-16 monitors were added, enhancing detection capabilities. 
The margin as established by the latest inspection has been 
maintained by the corrosion control program of EPRI Primary and 
Secondary Guidelines based on wet lay-up conditions. Thus, it can be 
reasonably concluded that this proposal to amend the Salem Unit 2 
Technical Specifications, on a one-time basis, to extend the steam 
generator inspection interval to end 24 months after start of Unit 2 
fuel cycle 10 does not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

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

    Date of application request: August 8, 1997.
    Description of amendment request: The proposed amendment would 
revise the surveillance requirements (SR) of Technical Specification 
(TS) 3/4.7.4 ``Essential Service Water System'' by removing the 
requirement to perform SR 4.7.4.b.1, 4.7.4.b.2 and 4.7.4.c during 
shutdown.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to TS has no adverse impact on the 
probability of occurrence or the consequences of an accident. The 
proposed amendment does not change or alter the design assumptions 
for the systems or components used to mitigate the consequences of 
an accident and the methodologies used in the accident analysis 
remain unchanged. The operating limits and the radiological 
consequences will not be changed. No design basis accidents will be 
affected by this change since the required TS surveillances will 
continue to be performed on an 18 month frequency.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    All design and performance criteria continue to be met and no 
new failure mechanisms have been identified. The proposed change 
does not affect the design or operation of any system or component 
in the plant since the required TS surveillances will continue to be 
performed on an 18 month frequency. The safety functions of the 
related structures, systems or components are not changed in any 
manner, nor is the reliability of any structure, system or component 
reduced. Conducting these surveillances online will not increase the 
possibility of plant transients. Since the safety functions and 
reliability are not adversely affected, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change will not affect or change a safety limit or 
affect plant operations since the required TS surveillances will 
continue to be performed on an 18 month frequency. This change will 
not reduce the margin of safety assumed in the accident analysis nor 
reduce any margin of safety as defined in the basis for any TS.

    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of application request: August 8, 1997.
    Description of amendment request: The proposed amendment would 
revise Table 3.3-3, Functional Units 4.b.2 and 5.a.2 of the Callaway 
Technical Specifications (TS) by (1) changing the main steam and 
feedwater isolation system (MSFIS) channels to be consistent with the 
requirements for the solid state protection system (SSPS), (2) adding a 
clarifying note, and (3) deleting and replacing Action Statements 27a 
and 34a with Action Statements 27 and 34. In addition, Table 4.3-2, 
Functional Units 4.b and 5.a are proposed to be revised by changing the 
slave relay quarterly surveillance to a quarterly actuation logic test 
for the MSFIS actuation and relays.
    Basis for proposed no significant hazards consideration 
determination:As required by 10 CFR 50.91(a), the licensee has provided 
its analysis of the

[[Page 66144]]

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 to Technical Specifications (TS) have no 
adverse impact on the probability of occurrence or the consequences 
of an accident. The proposed amendment does not change or alter the 
design assumptions for the systems or components used to mitigate 
the consequences of an accident and the methodologies used in the 
accident analysis remain unchanged. The operating limits and the 
radiological consequences will not be changed. No design basis 
accidents will be affected by these changes. The proposed changes do 
not result in any hardware changes.
    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.
    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 parameters governing normal plant operation. All design 
and performance criteria continue to be met and no new failure 
mechanisms have been identified. The proposed changes do not affect 
the design or operation of any system or component in the plant. The 
safety functions of the related structures, systems or components 
are not changed in any manner, nor is the reliability of any 
structure, system or component reduced. However, these changes are 
consistent with the requirements for the SSPS. Since the safety 
functions and reliability are not adversely affected, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes will not affect or change a safety limit or 
affect plant operations. These changes will not reduce the margin of 
safety assumed in the accident analysis nor reduce any margin of 
safety as defined in the basis for any TS. The proposed changes do 
not affect the acceptance criteria for any analyzed event. No 
setpoints are revised and the system response time will not be 
affected.

    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of application request: August 8, 1997.
    Description of amendment request: The proposed amendment would 
revise Table 3.7-2 of the Technical Specifications to specify that the 
lift setting tolerance for the main steam line safety valves be +3/-1% 
as-found and plus or minus 1% as-left. Table 2.2-1 would be revised by 
reducing the sensor error for the pressurizer pressure-high trip.
    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 main steam line safety valves are designed to mitigate 
transients by preventing overpressurization of the main steam 
system. The proposed change does not alter this design basis. The 
revised analysis shows that the probability or consequences of all 
previously analyzed accidents are not changed by increasing the 
setpoint tolerance of the safety valves. Therefore, there is no 
increase in the probability of occurrence or the consequences of any 
accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There is no new type of accident or malfunction created, the 
method and manner of plant operation will not change nor is there a 
change in the method in which any safety related system performs its 
function. Any main steam safety valve lifting at the extremes of the 
proposed tolerance will not result in a low lift setpoint that is 
less than the normal no load system pressure or a high lift setpoint 
that allows main steam system overpressurization.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This is based on the fact that no plant design changes are 
involved and the method and manner of plant operation remains the 
same. With the increased setpoint tolerance, the main steam safety 
valves will still prevent pressure from exceeding 110 percent of 
design pressure in accordance with the ASME code. All FSAR accident 
analysis conclusions remain valid and unaffected by this change.

    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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman.

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

    Date of application request: August 8, 1997.
    Description of amendment request: The proposed amendment 
application would revise feedwater isolation engineered safety feature 
actuation system (ESFAS) functions in Technical Specification Tables 
3.3-3, 3.3-4 and 4.3-2 as follows:
    (1) The Applicable MODES for Functional Units 5.a.1), Automatic 
Actuation Logic and Actuation Relays, and 5.a.2), Automatic Actuation 
Logic and Actuation Relays, in Tables 3.3-3 and 4.3-2 would be revised 
to add MODE 3.
    (2) A new Functional Unit 5.d, Steam Generator (SG) Water Level 
Low-Low (for feedwater isolation only), would be added to Tables 3.3-3, 
3.3-4, and 4.3-2.
    (3) In conjunction with the changes under item (2), the Applicable 
MODES in Table 3.3-3 for AFW SG Water Level Low-Low Functional Units 
6.d.1).c), Start Motor-Driven Pumps Vessel delta T (Power-1, Power-2), 
would be revised to delete MODE 3. Functional Unit 6.d.3) in Table 4.3-
2 would also be revised to delete MODE 3.
    (4) The Bases for Functional Unit 11.b, Reactor Trip P-4, in Table 
3.3-3 would be revised to add a note allowing the feedwater isolation 
function on P-4 coincident with low Tavg to be blocked.
    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:

Actuation Logic Applicability and New SG Water Level Low-Low Functional 
Unit

    1. The proposed change does not involve a significant increase 
in the probability or

[[Page 66145]]

consequences of an accident previously evaluated.
    The proposed changes impose more stringent requirements and have 
been reviewed to ensure no previously evaluated accident has been 
adversely affected. The more stringent requirements are imposed to 
ensure the plant's operation and testing are consistent with the 
safety analysis and licensing basis. 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.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed other 
than the bypass switch addressed in a separate 50.92 evaluation 
below) or changes in controlling parameters. The proposed changes do 
impose different requirements; however, these changes are consistent 
with assumptions made in the safety analysis and licensing basis. 
Actuation logic applicability is extended to MODE 3 and the SSPS 
slave relays that implement feedwater isolation on SG water level 
low-low will continue to be surveilled quarterly as they have always 
been tested. Thus, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The imposition of more stringent requirements does not reduce 
the margin of safety. The margin of safety would be increased since 
the scope of the Technical Specifications has been increased to 
include additional plant equipment and add additional Applicability 
requirements. The changes are consistent with the safety analysis 
and licensing basis. Therefore, the proposed changes do not involve 
a reduction in a margin of safety.

TTD Applicability

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The proposed change adds a relaxation 
to the Applicability for the SG Water Level Low-Low Vessel delta T 
channels. The proposed change in the Applicability will not affect 
any of the analysis assumptions for any of the accidents previously 
evaluated. The proposed change will not affect the probability of 
any event initiators nor will the proposed change affect the ability 
of any safety-related equipment to perform its intended function. A 
Vessel delta T channel should only be tripped if it is inoperable 
and the reactor is operating, when the need to restrict trip time 
delays is applicable. There will be no degradation in the 
performance of nor an increase in the number of challenges imposed 
on safety-related equipment assumed to function during an accident 
situation. Accident analyses have been performed with the maximum 
trip time delays enabled at power levels up to 19% RTP (10% RTP plus 
uncertainty). Therefore, operation in MODE 3 with the maximum trip 
time delays is enveloped. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no hardware changes nor are there any changes in the 
method by which any safety-related plant system performs its safety 
function. The change in Applicability will not impact the normal 
method of plant operation. The maximum trip time delay should be 
enabled in MODE 3 to preclude an unnecessary feedwater isolation or 
auxiliary feedwater actuation from occurring prior to the expiration 
of the trip time delay previously analyzed for MODE 1 operation. No 
new accident scenarios, transient precursors, failure mechanisms, or 
limiting single failures are introduced as a result of this change. 
Therefore, the proposed change does not create the possibility of a 
new of 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 does not affect the acceptance criteria for 
any analyzed event. There will be no effect on the manner in which 
safety limits or limiting safety system settings are determined nor 
will there be any effect on those plant systems necessary to assure 
the accomplishment of protection functions. There will be no impact 
on any margin of safety.

Feedwater Isolation on P-4/Low Tavg Bypass Switch

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses. The P-4/Low 
Tavg Bypass Switch design change will not impact any 
accidents previously evaluated in the FSAR since feedwater isolation 
upon reaching this function was never credited.
    The ESFAS will continue to function in a manner consistent with 
the accident analysis assumptions and the plant design basis. As 
such, there will be no degradation in the performance of nor an 
increase in the number of challenges to equipment assumed to 
function during an accident situation.
    This Technical Specification change does not affect the 
probability of any event initiators. There will be no change to 
normal plant operating parameters or accident mitigation 
capabilities. Therefore, there will be no increase in the 
probability or consequences of any accident occurring due to this 
change.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    There are no changes in the method by which any safety-related 
plant system performs its safety function and the normal manner of 
plant operation is unaffected, other than the proposed allowance to 
bypass feedwater isolation on P-4 coincident with low 
Tavg. This bypass switch modification will be performed 
under the design standards applicable to all safety system bypasses 
at Callaway, except for Section 4.12 of IEEE 279-1971. Section 4.12 
of IEEE 279-1971 requires that an operating bypass of a protective 
function be automatically removed whenever permissive conditions are 
not met. However, the subject circuitry does not provide a 
protective function. It is not assumed or credited in any safety 
analysis. In addition, plant conditions that would call for the 
restoration of the feedwater isolation function cannot occur without 
operator action to close the reactor trip breakers. Administrative 
controls will govern the proper use of and restoration from the 
proposed bypass. Although the addition of the bypass switch 
introduces the potential for an equipment malfunction of a different 
type from any previously evaluated in the FSAR, the possibility of a 
new or different type of accident is not created. The switch 
functions only to allow a manual bypass of feedwater isolation. The 
failure of the switch or its improper use will not be an event 
initiator for the previously analyzed Loss of Normal Feedwater event 
in FSAR Section 15.2.7 since it cannot fail in such a manner as to 
cause feedwater isolation.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of this change. There will be no adverse effect or challenges 
imposed on any safety-related system as a result of this change. 
Therefore, the possibility of a new or different type of accident is 
not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. There will be no impact on DNBR limits, 
FQ, F-delta-H, LOCA PCT, peak local power density, or any 
other 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

[[Page 66146]]

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

    Date of amendment request: November 5, 1997.
    Description of amendment request: The current Technical 
Specifications requirements prohibit loads in excess of 2500 pounds 
from traveling over irradiated fuel assemblies in the spent fuel pit. 
Due to the number of irradiated fuel assemblies currently stored in the 
spent fuel pit over years of operation, additional flexibility is 
needed to accomplish the movement of the spent fuel pit gates during 
refueling activities and to reduce fuel handling activities in 
preparation for refueling outages. In order to perform gate seal 
maintenance prior to each outage, a gate is moved across the irradiated 
fuel storage area to the cask handling area where it can be lifted out 
of the spent fuel pit. When a clear path of empty fuel storage cells 
cannot be established, seal maintenance cannot be performed unless 
relief from the current Limiting Condition of Operation is granted. The 
proposed changes will exempt these requirements for the movements of 
the spent fuel gates provided specific administrative controls are 
satisfied.
    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:

    Specifically, operation of the North Anna Power Station in 
accordance with the proposed changes will not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    The accident in question is a fuel handling accident in the 
spent fuel pit. The proposed changes will actually reduce the 
probability of a fuel handling accident by eliminating unnecessary 
fuel assembly movements. After this change is implemented, only 
those assemblies containing control rod assemblies will be subjected 
to such moves prior to movement of the gates instead of the current 
practice of moving all the fuel necessary to establish a load path 
of empty cells. A redundant rigging system will be provided which 
eliminates the possibility of a load drop due to a hoist failure. 
Furthermore, even though the double rigging system makes a load drop 
due to a hoist failure an incredible event, a calculation was 
performed to determine the effects of a direct impact load on a 
single fuel storage cell or the SFP [spent fuel pit] structure. The 
calculation concludes that there will be no adverse consequences to 
either irradiated fuel or the SFP structure. The plant design basis 
fuel handling accident will not be violated. Therefore, with the 
administrative controls in place to eliminate the possibility of a 
gate drop the probability of occurrence or the consequences of a 
fuel handling accident are not increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes establish adequate administrative controls 
over the spent fuel pit gate movements to prevent damage to stored 
irradiated fuel and fuel racks thereby ensuring the design basis 
fuel handling accident remains bounding and that fuel spacing is 
maintained in the racks precluding criticality.
    3. Involve a significant reduction in any margin of safety.
    The new administrative controls ensure that a postulated gate 
drop will not occur due to compliance with our licensing commitments 
to NUREG-0612 and the requirement to install a redundant rigging 
system to eliminate the possibility of a load drop initiated by 
hoist failure. Analysis has determined that in the event the gate 
was to be dropped from its controlled lift height: (1) There will be 
no damage to irradiated fuel caused by the direct impact loading on 
a single storage cell and (2) the fuel storage rack will maintain 
fuel in a non-critical array. A new criteria, demonstrating the 
ability of the pool floor to remain intact after a gate drop has 
been shown by analysis. New controls prevent the degradation of the 
existing margin of safety and ensure an adequate safety margin for 
the new criteria. The administrative controls added for the gate 
lift preclude the possibility of a load drop induced by a hoist 
failure and, therefore ensure the potential for radioactivity 
release and inadvertent criticality remain bounded by the present 
design basis. Therefore, the margin of safety is not reduced by the 
proposed change.

    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: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: James E. Lyons.

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

    Date of amendment request: November 18, 1997.
    Description of amendment request: The Technical Specifications 
surveillance requirements currently require testing and inspection of 
the Turbine Overspeed Protection System control valves, at least once 
per 31 days, to ensure their ability to prevent overspeeding of the 
turbine. Based on an analysis of Westinghouse BB-296 turbines with 
steam chests, the proposed change would increase the surveillance test 
interval from at least once per 31 days to at least once per 92 days.
    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:

    Specifically, operation of the North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No new or unique accident precursors are introduced by these 
changes in surveillance requirements. The probability of turbine 
missile ejection with an extended test interval to 92 days for the 
turbine governor and throttle valves has been determined to remain 
within the applicable NRC acceptance criteria. The heavy hub design 
of the turbine rotors provides further assurance that the 
probability of ejection of turbine missiles due to destructive 
overspeed remains within the acceptance criteria. Therefore, these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The demonstrated high reliability of the turbine governor and 
throttle valves and the verification of the operability of the other 
turbine control valves provide adequate assurance that the turbine 
overspeed protection system will operate as designed, if needed. 
Turbine governor and throttle valve testing performed to date has 
demonstrated the reliability of these valves. In addition, the 
operability of the other turbine valves (i.e., reheat and intercept 
stop valves) will continue to be verified every 18 months as 
required by the Technical Specifications.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Since the implementation of the proposed change to the 
surveillance requirements will not require hardware modifications 
(i.e., alterations to plant configuration), operation of the 
facilities with these proposed Technical Specifications does not 
create the possibility for any new or different kind of accident 
which has not been already been evaluated in the Updated Final 
Safety Analysis Report (UFSAR). In addition, the results of the 
probabilistic evaluation indicate that no additional transients have 
been introduced.
    The proposed revision to the Technical Specifications will not 
result in any physical alteration to any plant system, nor would 
there be a change in the method by which any safety-related system 
performs its function. The design and operation of the

[[Page 66147]]

turbine overspeed protection and turbine control systems are not 
being changed.
    The proposed Technical Specifications changes do not affect the 
design, operation, or failure modes of the valves and other 
components of the turbine overspeed protection system. Therefore, 
the proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not reduce the margin of safety as 
defined in the basis for any Technical Specifications. Furthermore, 
the total turbine missile ejection probability continues to be 
enveloped by the applicable acceptance criteria of 1E-5. The design 
and operation of the turbine overspeed protection and turbine 
control systems are not being changed and the operability of the 
turbine governor and throttle valves will be demonstrated on a 
refuelling outage basis. In addition, the results of the accident 
analyses, which are documented in the UFSAR, continue to bound 
operation with the proposed change in surveillance interval for the 
turbine throttle and governor valves, so that there is no safety 
margin reduction. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: James E. Lyons.

Previously Published Notices of Consideration of Issuance of Amendments 
to Facility Operating Licenses, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing

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

Tennessee Valley Authority, Docket Nos. 50-327, Sequoyah Nuclear Plant, 
Unit 1, Hamilton County, Tennessee

    Date of application for amendments: November 21, 1997.
    Description of amendments request: Amend Technical Specifications 
to add a one-time allowance through Operating Cycle 9 to Surveillance 
Requirement 4.4.3.2.1.b to perform stroke testing of the power-operated 
relief valve in Mode 5 rather than in Mode 4.
    Date of publication of individual notice in the Federal Register: 
December 1, 1997 (62 FR 63565).
    Expiration date of individual notice: December 31, 1997.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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: November 30, 1995, as 
supplemented March 15, 1996, March 6, 1997, and June 27, 1997.
    Brief description of amendments: The amendments incorporate 
references to a new Combustion Engineering, Inc. topical report 
describing steam generator tube sleeves, delete references to the 
previous CE topical report, incorporate sleeve/tube inspection scope 
and expansion criterion, revise the plugging limit for a CE sleeve to 
28% of the nominal sleeve wall thickness, and incorporate a post weld 
heat treatment for free span welds.
    Date of issuance: November 18, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 223 and 199.
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
176). The March 15, 1996, March 6, 1997, and June 27, 1997, letters 
provided clarifying information that did not change the initial 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated November 18, 1997.
    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: August 6, 1997.
    Brief description of amendments: The amendments address an 
unreviewed safety question associated with the handling of the spent 
fuel shipping cask at the Brunswick Steam Electric Plant, Units 1 and 
2.
    Date of issuance: December 2, 1997.
    Effective date: December 2, 1997.

[[Page 66148]]

    Amendment Nos.: 190 and 221.
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
authorize changes to the facility's Updated Final Safety Analysis 
Report.
    Date of initial notice in Federal Register: September 17, 1997 (62 
FR 48897) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 2, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendment: February 18, 1997.
    Brief description of amendment: This amendment revises the maximum 
allowable power range neutron flux high setpoints (percent of rated 
thermal power) shown in Technical Specification Table 3.7-1.
    Date of issuance: November 25, 1997.
    Effective date: November 25, 1997.
    Amendment No.: 75.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: April 9, 1997 (62 FR 
17225) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 25, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: April 7, 1997, as supplemented 
on August 7, 1997.
    Brief description of amendments: The amendments revise the 
technical specifications to permit installation and use of C&D Charter 
Power Systems, Inc., batteries.
    Date of issuance: November 25, 1997.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 93 and 93.
    Facility Operating License Nos. NPF-37 and NPF-66: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54868). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 25, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.

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

    Date of application for amendments: October 13, 1997, as 
supplemented by letters dated October 28 and November 5, 1997.
    Brief description of amendments: The amendments revise TS Table 
3.3-4, ``Engineered Safety Features [ESF] Actuation System Instrument 
Trip Setpoints.'' Specifically, the amendments support the replacement 
of three safety-related narrow range Refueling Water Storage Tank level 
instruments with three safety-related wide range level instruments. The 
ESF trip setpoint for the refueling water automatic switchover to 
recirculation is revised to account for the difference in instrument 
uncertainty associated with wide range level instruments and provides 
additional operator response time margin.
    Date of issuance: November 25, 1997.
    Effective date: Unit 1--As of the date of issuance to be 
implemented consistent with the refueling outage scheduled for June 
1998; Unit 2--As of the date of issuance to be implemented within 30 
days from the date of issuance.
    Amendment Nos.: 177 (Unit 1); 159 (Unit 2).
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54859). The October 28 and November 5, 1997, letters provided 
additional and clarifying information that did not change the scope of 
the October 13, 1997, application and the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 25, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, Charlotte, North Carolina.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina

    Date of application for amendments: October 10, 1997, as 
supplemented by letters dated November 3, 6, and 10, 1997.
    Brief description of amendments: The amendments revise Technical 
Specifications to implement alternate repair criteria for steam 
generator tubes that have degraded roll joints inside of the upper 
tubesheet. The alternate repair criteria would allow new roll joints to 
be installed below the degraded roll joints in the upper tubesheet.
    Date of issuance: November 21, 1997.
    Effective date: November 21, 1997.
    Amendment Nos.: Unit 1--227; Unit 2--227; Unit 3--224.
    Facility Operating License Nos. DPR-38, DPR-47, AND DPR-55: The 
amendments revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes. (62 FR 55835 dated October 28, 1997). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by November 28, 1997, but indicated that if the Commission 
makes a final no significant hazards consideration determination, any 
such hearing would take place after issuance of the amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and a final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated 
November 21, 1997.
    Attorney for licensee: M. J. Michael McGarry, III, Winston and 
Strawn, 1200 17th Street, NW., Washington, DC.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley Power 
Station, Unit No. 1, Shippingport, Pennsylvania

    Date of application for amendment: March 10, 1997, as supplemented 
July 28 and September 17, 1997.
    Brief description of amendment: The amendment modifies Technical 
Specification 3/4.4.5, ``Steam Generators,'' and its associated Bases

[[Page 66149]]

and adds a new license condition to Appendix C for Beaver Valley Power 
Station, Unit No. 1 (BVPS-1) to allow repair of steam generator tubes 
by installation of sleeves developed by ABB Combustion Engineering. In 
addition, the amendment deletes the option for using the kinetic 
sleeving methodology previously approved for use at BVPS-1.
    Date of issuance: November 25, 1997.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 208.
    Facility Operating License No. DPR-66: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19829). The July 28 and September 17, 1997, letters provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination or expand the amendment request 
beyond the scope of the April 23, 1997, Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 25, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: September 12, 1997, as 
supplemeneted November 7, 1997.
    Brief description of amendment: The proposed amendment involves a 
revision to the Emergency Diesel Generator protective relaying scheme 
at Crystal River Unit 3, to be reflected in the next revision to the 
Final Safety Analysis Report (FSAR).
    Date of issuance: December 1, 1997.
    Effective date: Effective upon issuance.
    Amendment No.: 159.
    Facility Operating License No. DPR-72:. Amendment revises the FSAR.
    Date of initial notice in Federal Register: September 30, 1997 (62 
FR 51165). By letter dated November 7, 1997, the licensee provided 
additional information which did not affect the original no significant 
hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 1, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

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

    Date of application for amendment: October 10, 1996, as 
supplemented March 25, June 6, and August 29, 1997.
    Brief description of amendment: The amendment extends the 
instrumentation surveillances for the condenser low vacuum, high 
temperature main steamline tunnel, recirculation flow, and reactor 
coolant leakage. Additionally, the change extends the equipment test/
operability checks for containment vent and purge isolation, 
electromagnetic relief valve operability, and drywell to torus leakage 
test.
    Date of Issuance: November 26, 1997.
    Effective date: November 26, 1997, with full implementation within 
60 days.
    Amendment No.: 193.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57485). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated November 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendment: May 20, 1997, as supplemented on 
September 23, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by relocating the containment isolation valve 
(CIV) list from the TSs to the Technical Requirements Manual in 
accordance with Generic Letter 91-08, ``Removal of Component Lists from 
the Technical Specifications.'' The amendment also changes the 
surveillance requirement for valves, blind flanges, and deactivated 
automatic valves located inside containment that are locked, sealed, or 
otherwise secured in the closed position from once every 31 days to 
during each cold shutdown, but no more than once per 92 days. The TS 
Bases is changed to reflect the relocation of the containment isolation 
valve list from the TSs to the Technical Requirements Manual and 
dicusses administrative controls for CIV operation in Modes 1 through 
4. Also, a license condition has been added to paragraph 2.C. of the 
Operating License to ensure enforceability and to provide a method of 
tracking the license condition back to the license amendment.
    Date of issuance: November 19, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 90 days.
    Amendment No: 210
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications and License Conditions.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33128). The September 23, 1997, letter provided clarification of the 
initial submittal and did not affect the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated November 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: September 16, 1997.
    Brief description of amendment: The amendment changes the main 
steam line American Society of Mechanical Engineers Code (Code) safety 
valves Technical Specifications (TSs) by: (1) Deleting TS Table 3.7.1, 
``Maximum Allowable Power Level-High Trip Setpoint with Inoperable 
Steam Line Safety Valves During Operation with Both Steam Generators,'' 
by not allowing operation in Mode 1 or 2 with inoperable Code safety 
valves while allowing operation in Mode 3 with up to three Code safety 
valves inoperable per steam generator, (2) modifing the associated 
action statement in TS 3.7.1.1 to reflect the operational changes, and 
(3) updating the TS Bases to reflect the proposed changes and include 
the correct amendment history numbers to

[[Page 66150]]

reflect previously approved amendments.
    Date of issuance: November 19, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 211.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52582). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendments: September 26, 1997.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3.4.B, ``Auxiliary Feedwater System,'' to provide 
specific guidance for conducting post-maintenance operational testing 
of the turbine-driven auxiliary feedwater pump and associated system 
valves to meet operability requirements and limiting conditions for 
operation during unit startup. Additionally, the amendments revise 
Table TS.3.5.2B to allow the auxiliary feedwater pump auto-start 
actuation instrumentation to be bypassed during startup and shutdown 
operations when the main feedwater pumps are not required to supply 
feedwater to the steam generators.
    Date of issuance: November 25, 1997.
    Effective date: November 25, 1997, with full implementation within 
30 days.
    Amendment Nos.: 134 and 126.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54874). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 25, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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

    Date of amendment request: May 20, 1996.
    Brief description of amendment: The amendment revises the technical 
specifications to correct and clarify surveillance test requirements 
for the reactor protective system and other plant instrumentation and 
control systems.
    Date of issuance: November 24, 1997.
    Effective date: November 24, 1997, to be implemented within 120 
days of the date of issuance.
    Amendment No.: 182.
    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44361). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 24, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear 
Generating Station, Unit No. 1, Salem County, New Jersey

    Date of application for amendment: May 10, 1996, as supplemented 
March 19 and August 29, 1997.
    Brief description of amendment: The amendment incorporates into the 
Technical Specifications the Margin Recovery portion of the Fuel 
Upgrade Margin Recovery Program and support increased steam generator 
plugging, improved fuel reliability, reduced fuel costs, longer fuel 
cycles, reduced spent fuel pool storage, and enhanced reactor safety.
    Date of issuance: November 26, 1997.
    Effective date: As of date of issuance. To be implemented on Unit 1 
prior to entry into Mode 2 from the current outage.
    Amendment No.: 201.
    Facility Operating License No. DPR-70: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34898). The March 19 and August 29, 1997, letters 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 
November 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.

Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San 
Diego County, California

    Date of application for amendments: January 4, 1996.
    Brief description of amendments: These amendments delete License 
Condition 2.C(26) for SONGS Unit 2 and License Condition 2.C(27) for 
SONGS 3. These license conditions require that Southern California 
Edison implement and maintain a plan for scheduling all capital 
modifications based on an NRC approved Integrated Implementation 
Schedule Program Plan.
    Date of issuance: December 3, 1997.
    Effective date: December 3, 1997.
    Amendment Nos.: Unit 2--137; Unit 3--129.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: April 10, 1996 (61 FR 
15997). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 3, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713.

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

    Date of application for amendment: January 20, 1997.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Section 3/4.5.2, ``Emergency Core Cooling Systems, 
ECCS Subsystems-Tavg greater than or equal to 280 deg.F,'' 
TS Section 3/4.5.3, ``Emergency Core Cooling Systems, ECCS 
Subsystems-Tavg less than 280  deg.F,'' and TS Section 3/
4.7, ``Plant Systems.'' Several surveillance intervals were changed 
from 18 months to once each refueling interval.
    Date of issuance: December 2, 1997.

[[Page 66151]]

    Effective date: December 2, 1997.
    Amendment No.: 216
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11498). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 2, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: September 17, 1996, as 
supplemented by letters dated November 27, 1996, and October 14, 1997.
    Brief description of amendment: This amendment revises the 
surveillance interval from 18 months to less than or equal to 730 days, 
nominally 24 months, for Technical Specification (TS) 3/4.5.2, 
``Emergency Core Cooling Systems--ECCS Subsystems--Tavg 
greater than or equal to 280 degrees F''; TS 3/4.6.5.1, ``Containment 
Systems--Shield Building--Emergency Ventilation System''; TS 3/4.7.6.1, 
``Plant Systems--Control Room Emergency Ventilation System''; TS 3/
4.7.7, ``Plant Systems--Snubbers''; TS 3/4.9.12, ``Refueling 
Operations--Storage Pool Ventilation''; and TS Bases 3/4.7.7--
``Snubbers.''
    Date of issuance: December 2, 1997.
    Effective date: Immediately, and shall be implemented no later than 
120 days after issuance.
    Amendment No.: 217.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 9, 1996 (61 FR 
52972). The supplemental information submitted by the licensees did not 
impact the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 2, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

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

    Date of application for amendment: December 11, 1996 (as 
supplemented by letter dated January 6, 1997), January 30, 1997 (as 
supplemented by letter dated September 15, 1997), and April 18, 1997.
    Brief description of amendment: This amendment extends surveillance 
requirement intervals from 18 to 24 months, revises setpoints, and 
revises TS 2.2, ``Limiting Safety System Settings.'' Administrative 
changes have also been made.
    Date of issuance: December 2, 1997.
    Effective date: December 2, 1997.
    Amendment No.: 218.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Dates of initial notice in Federal Register: January 15, 1997 (62 
FR 2194), March 12, 1997 (62 FR 11498) and June 4, 1997 (62 FR 30654). 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated December 2, 1997.
    No significant hazards consideration comments received: No. The 
supplemental information provided by the licensees did not affect the 
proposed no significant hazards consideration.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, OH 43606.

    Dated at Rockville, Maryland, this 10th day of December 1997.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 97-32763 Filed 12-16-97; 8:45 am]
BILLING CODE 7590-01-P