[Federal Register Volume 64, Number 76 (Wednesday, April 21, 1999)]
[Notices]
[Pages 19554-19570]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 99-9839]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 27 through April 9, 1999. The last 
biweekly notice was published on April 7, 1999 (64 FR 17021).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration Services, Office of

[[Page 19555]]

Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the NRC Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By May 21, 1999, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: 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.

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

    Date of amendment request: March 23, 1999.
    Description of amendment request: The proposed amendment would 
modify Technical Specification Surveillance Requirement 4.4.1.1.1 to 
require each recirculation pump discharge valve to be demonstrated 
OPERABLE at least once every 18 months and will delete footnote * that 
applies to Technical Specification 4.4.1.1.1.
    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 change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to the Technical Specifications (TS) would 
modify the frequency of cycling the recirculation pump discharge 
valves from ``each STARTUP*

[[Page 19556]]

prior to THERMAL POWER exceeding 25% of RATED THERMAL POWER'' to 
``at least once per 18 months;'' and replace the footnote applicable 
to TS 4.4.1.1, ``*If not performed in the previous 31 days'' with 
``*Not Used.'' The change in testing frequency does not affect the 
probability of an accident since the valve testing is not related to 
accident initiation sequences. Consequences of accidents are not 
significantly increased because the proposed testing interval 
provides reasonable assurance that the valves will function. Testing 
of the valves will still be performed on a frequency that is allowed 
by TS if no events occur that require entry into Mode 3 or Mode 4. 
Therefore, the change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Testing the valves in accordance with the inservice testing (IST) 
program on the same testing frequency as testing performed for the 
low pressure coolant injection system, provides adequate assurance 
that the valves can perform their safety function and will not 
increase the consequences of an accident previously evaluated. The 
change to the footnote is administrative in nature and will have no 
effect on the probability of an accident and will not increase any 
safety consequences.
    2. The change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes revise performing the testing of the 
recirculation pump discharge valves from ``prior to Startup* not to 
exceed 25% of rated thermal power.'' to ``at least once per 18 
months'' and replace the footnote applicable to TS 4.4.1.1'' *If not 
performed in the previous 31 days'' with ``*Not Used'' does not 
result in a new accident precursor since the test only verifies that 
the valve can close which is its safety function. Deleting the 
information contained in footnote ``*'' that applies to TS 4.4.1.1.1 
and designating it as ``* Not Used.'' is administrative in nature 
with no safety significance. Therefore, no different type of 
accident from any previously evaluated is introduced.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed changes revise the frequency of cycling the 
recirculation pump discharge valves from ``each STARTUP* prior to 
THERMAL POWER exceeding 25% of RATED THERMAL POWER'' to ``at least 
once per 18 months'' and replace the footnote applicable to TS 
4.4.1.1 ``*If not performed in the previous 31 days'' with ``*Not 
Used.'' Altering the test frequency does not change valve stroke 
time or other performance or design characteristics related to the 
safety function of the valves. The potential for failure of the 
valve to close is not changed as a result of the proposed change 
since the same frequency is allowed by the current TS if no events 
occur that require entry into Mode 3 or Mode 4. Performing stroke 
time testing on a refueling outage basis and MOV testing on a 
periodic basis does not decrease the margin of safety associated 
with the valve performing its safety function. Revising footnote * 
is an administrative change and has no safety consequence. 
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: Monroe County Library System, 
Ellis Reference and Information Center, 3700 South Custer Road, Monroe, 
Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Section Chief: George F. Dick, Acting.

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

    Date of amendment request: March 3, 1999.
    Description of amendment request: The proposed amendments would 
change the required qualifications for operations management specified 
in the Technical Specifications (TSs) for the Beaver Valley Power 
Station, Units 1 and 2 (BVPS-1 and BVPS-2). The requirement that the 
operations manager hold a Senior Reactor Operator (SRO) license at the 
time of appointment would be changed in the TSs to require that the 
assistant operations managers, one for each unit, hold an SRO license 
on their assigned unit. The TSs would not then require the operations 
manager hold an SRO license. Additionally, the Updated Final Safety 
Analysis Report (UFSAR) for each unit would be changed to require the 
operations manager to hold, or have held, an SRO license rather than 
presently hold a license. The UFSAR would require the same as the TS; 
that the assistant operations managers hold an SRO license on the unit 
to which they are assigned. Finally, the proposed amendments would 
substitute generic personnel titles for plant-specific personnel titles 
in the BVPS-1 and BVPS-2 TSs. The correlation between generic titles 
and plant-specific titles would be provided in the BVPS-2 UFSAR.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes are administrative in nature. The revised 
requirements for who must hold a current senior reactor operator 
(SRO) License does not involve any change to the configuration or 
method of operation of any plant equipment that is used to mitigate 
the consequences of an accident nor alter the conditions or 
assumptions in any of the Updated Final Safety Analysis Report 
[UFSAR] accident analyses. The requirement that the operations 
manager hold or have held an SRO License is included in the revised 
Position Qualifications in the Unit 2 UFSAR, Table 13.1-2, sheet 30 
of 35. The title changes are being made, consistent with TSTF-65, 
Rev 1 and help avoid the need for future Technical Specification 
changes. Therefore, it can be concluded that the proposed changes do 
not involve any increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    No new failure modes are defined for any plant system or 
component important to safety nor has any new limiting failure been 
identified as a result of the proposed changes. Therefore, it can be 
concluded that the proposed change does not create the possibility 
of a new or different kind of accident from those previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes are administrative in nature. One of the 
proposed changes requires that the manager who directly supervises 
the licensed operators at each unit be the holder of a current SRO 
license. The other change modifies personnel titles. Therefore, it 
can be concluded that the proposed changes do not involve any 
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: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Section Chief: Singh Bajwa.

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit No. 2, Shippingport, Pennsylvania

    Date of amendment request: March 16, 1999.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3/4.7.1.3

[[Page 19557]]

and associated Bases for the Primary Plant Demineralized Water (PPDW) 
System to clarify that the minimum specified volume of water in the 
PPDW Storage Tank is a usable volume. Additionally, the minimum usable 
volume of water in the PPDW Storage tank is increased, and a clarifying 
footnote that the specified value is an analysis value is added. 
Finally, several editorial and administrative changes, such as revision 
of action statement wording, addition of license number to the TS page, 
and addition of clarifying information to the TS Bases regarding 
analysis assumptions are made.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The failure of the primary plant demineralized water (PPDW) 
storage tank to provide a sufficient source of water to the 
Auxiliary Feedwater (AFW) System is not an accident initiating 
event. Therefore, the probability of an accident previously 
evaluated is not increased by this proposed amendment.
    Limiting Condition for Operation (LCO) 3.7.1.3 titled ``Primary 
Plant Demineralized Water (PPDW)'' will be revised to specify the 
required value for PPDW storage tank volume as a usable volume. To 
reflect the value currently assumed in the analysis, the value 
stated in the LCO, for minimum required PPDW storage tank volume, 
would be slightly increased. The addition of proposed Footnote (1) 
to LCO 3.7.1.3 will ensure that plant operators recognize that the 
specified volume is an analysis value and that the value does not 
include measurement uncertainties. This footnote will require plant 
procedures to specify an increased required volume in the PPDW 
storage tank to account for measurement uncertainties. The proposed 
revisions to LCO 3.7.1.3 will assure that the PPDW storage tank 
minimum usable volume is maintained consistent with the design basis 
for the PPDW storage tank. The PPDW storage tank will continue to 
provide a sufficient source of water to the AFW pumps. Maintaining a 
sufficient source of water will ensure that the AFW System is 
capable of mitigating the consequences of Design Basis Accidents 
(DBAs) that could result in overpressurization of the RCS pressure 
boundary. The AFW system will continue to be capable of providing an 
emergency source of feedwater to the steam generators to act as heat 
sinks for sensible and decay heat removal from the reactor core. A 
sufficient volume of water will continue to be maintained in the 
PPDW storage tank to satisfy the Safe Shutdown evaluation.
    The proposed changes to the Action statements will remove the 
required water volume value and add wording pertaining to the water 
volume not being within the limit. The LCO clearly states the value 
for the minimum required volume in the PPDW storage tank. Therefore, 
the proposed modification to the Action statements is administrative 
in nature and does not affect plant safety. The additional Bases 
wording pertaining to reactor coolant pump operation is 
administrative in nature and does not affect plant safety. The 
remaining change, which consists of the addition of plant operating 
license number, is editorial in nature and does not affect plant 
safety.
    Therefore, operation of the facility in accordance with the 
proposed amendment does not involve a significant increase in the 
probability or consequence of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed amendment will not change the physical plant or the 
modes of plant operation defined in the operating license. This 
change does not involve the addition or modification of plant 
equipment nor does it alter the design or operation of plant 
systems. The proposed amendment will require that the minimum volume 
in the PPDW storage tank be maintained consistent with analysis 
assumptions.
    Therefore, operation of the facility in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The minimum required volume in the PPDW storage tank would be 
slightly increased over the currently required value. This increase 
in the required volume will ensure that an adequate volume of water 
is maintained in the PPDW storage tank. The proposed addition of the 
term ``usable,'' along with the addition of Footnote (1), will 
ensure that the water volume specified in LCO 3.7.1.3 is 
appropriately increased in plant procedures to account for unusable 
volume in the tank and for measurement uncertainties. A sufficient 
volume of water will continue to be maintained in the PPDW storage 
tank to satisfy the Safe Shutdown evaluation.
    The PPDW storage tank will continue to provide a sufficient 
source of water to the AFW pumps to ensure that the AFW System is 
capable of mitigating the consequences of DBAs that could result in 
overpressurization of the RCS pressure boundary. The AFW system will 
continue to be capable of providing an emergency source of feedwater 
to the steam generators to act as heat sinks for sensible and decay 
heat removal from the reactor core.
    The proposed changes to the Action statements will remove the 
required water volume value and add wording pertaining to the water 
volume not being within the limit. The LCO clearly states the value 
for the minimum required volume in the PPDW storage tank. Therefore, 
the proposed modification to the Action statements is administrative 
in nature and does not affect plant safety. The additional Bases 
wording pertaining to reactor coolant pump operation is 
administrative in nature and does not affect plant safety. The 
remaining change, which consists of the addition of plant operating 
license number, is editorial in nature and does not affect plant 
safety.
    Therefore, the proposed amendment does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: S. Singh Bajwa.

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

    Date of amendment request: August 31, 1998, and revised March 18, 
1999.
    Description of amendment request: The proposed amendment would 
revise Improved Technical Specification (ITS) 5.6.2.10, ``Steam 
Generator (OTSG [once-through steam generator]) Tube Surveillance 
Program,'' to include a new repair process, called a ``repair roll'' or 
``re-roll.'' The process would be used to repair steam generator tubes 
with defects within the upper tubesheet. Changes to inservice 
inspection and reporting requirements are proposed for tubes which are 
repaired using this process. In addition, several format and editorial 
changes are proposed to ITS 5.6.2.10 and to ITS 5.7.2, ``Special 
Reports,'' for clarification purposes. The March 18, 1999 revision 
superceded the August 31, 1998 request, and includes the results of 
recent accident analyses conducted to identify the maximum OTSG tube 
tensile loads. As a result of the increased tube tensile loads, some 
tubes will require a double repair roll. The double repair roll 
methodology was not included in the original amendment request. 
Therefore, this notice revises the previous Notice of Consideration of 
Issuance of Amendment (63 FR 56249).
    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.


[[Page 19558]]


    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The repair roll process is a method to create a new primary-to-
secondary pressure boundary joint in the upper tubesheet of Babcock 
& Wilcox (B&W) Once Through Steam Generators (OTSGs) manufactured 
with Inconel Alloy 600 tubes. The repair roll process creates a new 
roll joint in the OTSG tubes at a point closer to the secondary face 
of the tubesheet than the existing roll joint. The new pressure 
boundary is established by the repair roll to remove degradation of 
the existing roll joint from pressure boundary service. The repair 
roll process has been qualified as an acceptable repair methodology 
for use in the upper tubesheet of the Crystal River Unit 3 (CR-3) 
OTSGs. The proposed License Amendment Request (LAR) proposes to 
implement the qualified OTSG tube repair roll process, and also 
addresses several editorial and format changes which do not impact 
the current CR-3 accident analyses.
    The qualification of the OTSG tube repair roll methodology is 
based on establishing a mechanical joint length that will carry all 
structural loads imposed on the OTSG tubes while maintaining the 
required margins during normal and accident conditions. A series of 
tests and analyses were performed to establish the minimum 
acceptable length of the OTSG tube repair roll. Tests performed 
included leak, tensile, fatigue, ultimate load and eddy-current 
measurement uncertainty. The analyses evaluated plant operating and 
faulted load conditions, in addition to OTSG tubesheet bow effects. 
OTSG tube leakage remains bounded by the evaluation presented in the 
CR-3 Final Safety Analysis Report (FSAR) for a main steam line break 
(MSLB). The proposed change also includes a description of the 
required inspection program for the OTSG tube repair rolls. The 
additional inspection requirements do not change any accident 
initiators. The proposed inspections following OTSG tube repair roll 
installation, and during future inservice inspections, assure 
continuous monitoring of these tubes such that inservice degradation 
of tubes repaired by the repair roll process will be detected. Based 
on the qualification testing and analyses performed, as well as the 
industry experience with the use of OTSG tube repair roll processes, 
there are no new safety issues associated with the use of repair 
roll methodology. Therefore, this change does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from previously evaluated accidents?
    The repair roll creates no new failure modes or accident 
scenarios. The new pressure boundary joint created by the repair 
roll process has been demonstrated, by testing and analysis, to 
provide structural and leakage integrity equivalent to the original 
design and construction for all normal operating and accident 
conditions. Furthermore, the testing and analysis demonstrate the 
repair roll process creates no new adverse effects for the repaired 
tube and does not change the design or operating characteristics of 
the OTSGs. In the unlikely event that a tube with a repair roll 
should fail and sever completely at the transition of the repair 
roll region, the tube would remain engaged in the tubesheet bore, 
preventing interaction with other surrounding tubes. In this case, 
leakage is bounded by the steam generator tube rupture (SGTR) 
accident analysis. Therefore, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) Involve a significant reduction in a margin of safety?
    The repair roll process effectively removes the defective/
degraded area of the tube from service. The repair roll interface 
created with the tubesheet satisfies the necessary structural, 
leakage and heat transfer requirements. The mechanical joint is 
constrained within the tubesheet bore; thus, there is no additional 
risk associated with tube rupture. The accident leakage is shown to 
be less than one gallon per minute primary-to-secondary leakage. 
Therefore, the FSAR analyzed accident scenarios remain bounding, and 
the use of the repair roll process does not reduce the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 34428.
    Attorney for licensee: R. Alexander Glenn, General Counsel, Florida 
Power Corporation, MAC-A5A, P. O. Box 14042, St. Petersburg, Florida 
33733-4042.
    NRC Section Chief: Sheri R. Peterson.

Illinois Power Company, Docket No. 50-461, Clinton Power Station, Unit 
1, DeWitt County, Illinois

    Date of amendment request: March 1, 1999.
    Description of amendment request: The proposed amendment would 
approve changes to the Updated Safety Analysis Report (USAR) concerning 
design requirements for physical protection from tornado missiles for 
safety-related equipment.
    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 activity does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    The associated USAR changes reflect use of the Electric Power 
Research Institute (EPRI) Topical Report, ``Tornado Missile Risk 
Evaluation Methodology, (EPRI NP-2005),'' Volumes I and II. This 
methodology has been reviewed, accepted and documented in an NRC 
Safety Evaluation dated October 26, 1983. The NRC concluded that: 
``the EPRI methodology can be utilized when assessing the need for 
positive tornado missile protection for specific safety-related 
plant features in accordance with the criteria of SRP Section 
3.5.1.4.''
    The EPRI methodology has been previously applied at CPS to 
resolve previously identified missile protection issues during the 
initial licensing of the plant. The NRC documented their acceptance 
of this methodology in Supplement 6 to the CPS Safety Evaluation 
Report (NUREG-0853, July 1986).
    As permitted in the Standard Review Plan (NUREG-0800), the total 
probability of damage to plant systems or components initiated from 
tornado missiles leading to consequences in excess of 10 CFR Part 
100 guidelines will be maintained below an acceptable level. The 
results of the current tornado missile hazards analysis are such 
that the calculated total tornado missile hazard probability is 
approximately 3.4  x  10-7 per year. This is lower than the value 
determined to be acceptable, i.e., 1  x  10-6 per year.
    Although it has been calculated that these targets have a higher 
total probability of being exposed to tornado missiles than that 
described to be acceptable in SER Supplement 6, Section 3.5.1.3, the 
revised tornado missile hazards analysis for CPS has determined that 
this probability is acceptably low.
    With respect to the probability of occurrence or the 
consequences of an accident previously analyzed in the USAR, the 
possibility of a tornado reaching CPS and causing damage to plant 
systems, structures and components is a design basis event 
considered in the USAR. The changes being proposed herein do not 
affect the probability that a tornado will reach the plant, but they 
do, from a licensing basis perspective, reflect a slightly 
increased, calculated probability that missiles generated by the 
winds of a tornado might strike certain plant systems or components. 
The tornado missile analysis determined that there are a limited 
number of safety-related components that theoretically could be 
struck. The probability of tornado-generated missile strikes on 
important systems and components (as discussed in Regulatory Guide 
1.117) was analyzed using the probability methods described above. 
Based on the low, calculated probability, the total (cumulative) 
probability of strikes will be maintained below an adequately low 
acceptance criterion to ensure overall plant safety. On this basis, 
the proposed change is not considered to constitute a significant 
increase in the probability of occurrence or the consequences of an 
accident, due to the low probability of a tornado missile striking 
safety-related systems or components.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of previously evaluated 
accidents.

[[Page 19559]]

    (2) The proposed activity does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes involve evaluation of whether any physical 
protection of safety-related equipment from tornado missiles is 
required relative to the probability of such damage without physical 
protection. A tornado at CPS is a design basis event considered in 
the USAR, however, a tornado is not postulated to act as an 
initiator for any new or different kind of accident, or to occur 
coincident with any of the design basis accidents in the USAR. The 
low probability threshold established for missile damage to plant 
systems is consistent with these assumptions.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident.
    (3) The proposed activity does not involve a significant 
reduction in a margin of safety.
    Under the proposed change, physical protection of safety-related 
equipment from tornado missiles must be considered if it has been 
determined that the calculated total tornado missile hazard 
probability is greater than 1  x  10-6 per year. The proposed change 
to the USAR to specifically identify this threshold may slightly 
increase the probability of a malfunction of equipment important to 
safety previously evaluated in the safety analysis report (i.e., 
changing the requirements from protecting all safety-related systems 
and components to not requiring protection if there is an extremely 
low probability that a tornado missile could strike portions of 
safety related systems and components). However, the changes are 
consistent with the minimum acceptable requirements as documented in 
the NRC's Safety Evaluation Report dated October 23, 1983. 
Therefore, there will be no significant reduction to the margin of 
safety that may be associated with the potential for safety-related 
equipment to be damaged from tornado-generated missiles.
    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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, IL 61727.
    Attorney for licensee: Leah Manning Stetzner, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, IL 62525.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: March 17, 1999.
    Description of amendment request: The licensee is proposing to 
change Technical Specifications 3.5.2, ``Emergency Core Cooling 
Systems--ECCS Subsystems--Tavg greater than or equal to 300  deg.F;'' 
3.7.1.7, ``Plant Systems--Atmospheric Steam Dump Valves;'' and 3.7.6.1, 
``Plant Systems--Control Room Emergency Ventilation System.'' The 
proposed Technical Specification changes will revise (1) surveillance 
requirements for Emergency Core Cooling System valves, (2) the 
atmospheric steam dump valve requirements to focus on the steam release 
path instead of the individual valves, and (3) the allowed outage times 
for the atmospheric steam dump valves and Control Room Emergency 
Ventilation System.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do 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 compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

Technical Specification 3.5.2

    The removal of 2-CH-434, a manual valve, from the list of valves 
to be checked every 31 days by Surveillance Requirement (SR) 
4.5.2.a.10 will not change the requirement for this containment 
isolation valve to be locked closed. The position of valve 2-CH-434, 
and the associated locking device, will be verified by SR 4.6.1.1.a. 
Although this change will result in the position of 2-CH-434 being 
checked less often, there are sufficient Technical Specification and 
administrative requirements to ensure that 2-CH-434 will be 
maintained in the proper position. An additional benefit of this 
proposed change will be a reduction in personnel exposure since 2-
CH-434 is located inside containment. This proposed change will not 
result in any modification to Emergency Core Cooling System (ECCS) 
alignment or operation.
    The addition of the footnote to SR 4.5.2.a.10 will clarify that 
2-SI-306 is pinned and locked open to the required throttle 
position. 2-SI-306, which is the Shutdown Cooling (SDC) System 
throttle valve in the discharge piping of the SDC pumps, is required 
to be left in a throttled position after SDC has been secured to 
ensure sufficient low pressure safety injection (LPSI) flow will be 
available. This proposed change will not result in any modification 
to ECCS alignment or operation.
    The change in the valve nomenclature used in SR 4.5.2.e and 
Table 4.5-1 from throttle valve to injection valve will eliminate 
any confusion between valve description and valve operation. This 
proposed change will not result in any modification to ECCS 
alignment or operation.
    The addition of the License Amendment Number to the bottom of 
Page 3/4 5-6a will not result in a technical change to this 
Technical Specification.

Technical Specification 3.7.1.7

    The proposed changes will expand the scope of Technical 
Specification 3.7.1.7 to include the steam release path, instead of 
just the individual atmospheric dump valves (ADVs). The allowed 
outage times will be modified to address inoperable ADV lines and 
the impact inoperable ADV lines will have on the ability of 
Millstone Unit No. 2 to mitigate a loss of coolant accident (LOCA). 
If one ADV line is inoperable, a plant shutdown will be required if 
the ADV line is not restored to operable status within 48 hours. An 
allowed outage time of 48 hours to restore the ADV line to operable 
status is acceptable based on the low probability of a LOCA 
occurring during this time period, and the subsequent loss of 
offsite power and the failure of one train of high pressure safety 
injection (HPSI). This is also consistent with the allowed outage 
time for one ECCS train (Technical Specification 3.5.2).
    If two ADV lines are inoperable, a plant shutdown will be 
required if at least one ADV line is not restored to operable status 
within one hour. The plant will be required to be in Mode 3 within 
the following 6 hours. These time requirements are based on 
Technical Specification 3.0.3. However, the time to reach Mode 4 
will remain at the ``following 24 hours'' to reflect the impact 
inoperable ADV lines may have on the time to cool down the plant.
    The proposed change to the surveillance requirement will ensure 
operation of the ADV lines, consistent with the accident analysis, 
is verified.
    The proposed change in component nomenclature is consistent with 
current Millstone Unit No. 2 terminology. This is not a technical 
change.
    The proposed changes to the Bases of Technical Specification 
3.7.1.7 are consistent with the changes just described.

Technical Specification 3.7.6.1

    The action requirements for the Control Room Emergency 
Ventilation System will be modified to address the situation when 
both Control Room Emergency Ventilation Trains are inoperable in 
Modes 1, 2, 3, and 4. This situation is expected to occur during 
normal plant operation when the air filters in the common supply 
header to both trains are cleaned/replaced. Since this is a common 
supply header, both trains are affected and would be inoperable. The 
proposed action requirements will address this situation so

[[Page 19560]]

that Technical Specification 3.0.3 will not be entered as a result 
of an expected plant activity. However, since the proposed action 
requirements are the same as the requirements of Technical 
Specification 3.0.3, the time the plant is allowed to operate in 
this situation will not change.
    The proposed changes to the Technical Specifications and 
associated Bases will have no adverse effect on plant operation or 
accident mitigation equipment. The proposed changes will ensure that 
the necessary equipment to mitigate the design basis accidents will 
be available, or a plant shutdown will be required. In addition, the 
proposed changes can not cause an accident, and they will ensure the 
accident mitigation equipment will continue to operate as assumed in 
the analyses to mitigate the design basis accidents. Therefore, 
there will be no significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes to the Technical Specifications and 
associated Bases will have no adverse effect on plant operation or 
accident mitigation equipment. The proposed changes will ensure that 
the necessary equipment to mitigate the design basis accidents will 
be available, or a plant shutdown will be required. Therefore, the 
proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to the Technical Specifications and 
associated Bases will ensure that the necessary equipment to 
mitigate the design basis accidents will be available, or a plant 
shutdown will be required. The proposed changes will not result in 
any plant configuration changes. There will be no adverse effect on 
plant operation or accident mitigation equipment. The plant response 
to the design basis accidents will not change. Therefore, there will 
be no significant reduction in the margin of safety as defined in 
the Bases for the Technical Specifications affected by these 
proposed changes.

    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 Section Chief: James W. Clifford.

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

    Date of amendment request: March 19, 1999.
    Description of amendment request: The proposed changes will 
relocate Technical Specifications (TSs) 3.3.3.2, ``Instrumentation, 
Incore Detectors,'' 3.3.3.3, ``Instrumentation, Meteorological 
Instrumentation,'' to the Millstone, Unit No. 2 Technical Review Manual 
(TRM). Index Page V will be revised by eliminating the sections 
corresponding to incore detectors (Page \3/4\ 3-0), seismic 
instrumentation (Page \3/4\ 3-32), and meteorological instrumentation 
(Page \3/4\ 3-36). These sections, as well as changes to the associated 
Bases, will be relocated to the TRM.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do 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 compromised. The 
proposed changes do not involve an SHC because the changes would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Technical Specification 3.3.3.2, Instrumentation, ``Incore 
Detectors,'' is proposed to be relocated to the TRM where future 
changes will be controlled in accordance with 10 CFR 50.59. 
Relocation of this Technical Specification to the TRM does not imply 
any reduction in its importance in confirming that core power 
distribution are bounded by safety analysis limits. These 
instruments are neither used for, nor capable of, detecting a 
significant abnormal degradation of the reactor coolant pressure 
boundary before a design basis accident, nor do they function as a 
primary success path to mitigate events which assume a failure of, 
or a challenge to, the integrity of fission product barriers. 
Although the core power distribution (measured by the incore 
detectors) constitutes an important initial condition to design 
basis accidents and therefore needs to be addressed by Technical 
Specifications, the detectors themselves are not an active design 
feature needed to preclude analyzed accidents or transients. The 
proposed change will not alter the way core power distribution is 
measured by the incore detectors, nor will it alter any of the power 
distribution assumptions used in the accident analysis. Therefore, 
this change will not significantly increase the probability or 
consequences of an accident previously evaluated.
    Technical Specification 3.3.3.3, Instrumentation, ``Seismic 
Instrumentation,'' is proposed to be relocated to the TRM where 
future changes will be controlled in accordance with 10 CFR 50.59. 
Relocation of Technical Specification 3.3.3.3 to the TRM does not 
imply any reduction in its importance in determining the response of 
those nuclear power plant features important to safety in the event 
of an earthquake. Seismic instrumentation does not actuate any 
protective equipment or serve any direct role in the mitigation of 
an accident. The capability of the plant to withstand a seismic 
event or other design basis accident is determined by the initial 
design and construction of systems, structures, and components. The 
instrumentation is used to alert operators to the seismic event and 
evaluate the plant response. The seismic instrumentation does not 
serve as a protective design feature or part of a primary success 
path for events which challenge fission product barriers. The 
proposed change will not alter the way these instruments are used in 
determining the response of those nuclear power plant features 
important to safety in the event of an earthquake, nor will it alter 
the capability of the plant to withstand a seismic event. Therefore, 
this change will not significantly increase the probability or 
consequences of an accident previously evaluated.
    Technical Specification 3.3.3.4, Instrumentation, 
``Meteorological Instrumentation,'' is proposed to be relocated to 
the TRM where future changes will be controlled in accordance with 
10 CFR 50.59. Relocation of Technical Specification 3.3.3.4 to the 
TRM does not imply any reduction in its importance in providing a 
basis for estimating annual radiation doses resulting from 
radioactive materials released in airborne effluents. The 
instrumentation does not serve to ensure that the plant is operated 
within the bounds of initial conditions assumed in design basis 
accident and transient analyses or that the plant will be operated 
to preclude transients or accidents. Likewise, the meteorological 
instrumentation does not serve as part of the primary success path 
of a safety sequence analysis used to demonstrate that the 
consequences of these events are within the appropriate acceptance 
criteria. The proposed change will not alter the way these 
instruments are used in providing a basis for estimating annual 
radiation doses resulting from radioactive materials released in 
airborne effluents. Therefore, this change will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    Revision of Index page V and the proposed changes to the 
associated Bases sections are administrative changes. Therefore, 
these changes will not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed changes do not alter how any structure, system, or 
component functions. There will be no effect on

[[Page 19561]]

equipment important to safety. The proposed changes have no effect 
on any of the design basis accidents previously evaluated. 
Therefore, this License Amendment Request does not impact the 
probability of an accident previously evaluated, nor does it involve 
a significant increase in the consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. They do not alter the way any 
structure, system, or component functions and do not alter the 
manner in which the plant is operated. The proposed changes do not 
introduce any new failure modes. Therefore, the proposed changes 
will 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 relocation of incore detector instrumentation 
requirements to the TRM does not imply any reduction in their 
importance in confirming that core power distribution is bounded by 
safety analysis limits. The incore detectors will still be used to 
measure core power distribution and the assumptions used in the 
accident analysis will be verified. The proposed relocation of 
seismic instrumentation requirements to the TRM does not imply any 
reduction in their importance in determining the response of those 
nuclear power plant features important to safety in the event of an 
earthquake. The seismic instrumentation will still be used to 
determine the response of those nuclear power plant features 
important to safety in the event of an earthquake. The capability of 
the plant to withstand a seismic or other design basis accident, 
which is determined by the initial design and construction of 
systems, structures, and components will not be altered. The 
relocation of meteorological instrumentation requirements to the TRM 
does not change the way these instruments are used in providing a 
basis for estimating annual radiation doses resulting from 
radioactive materials released in airborne effluents. The 
meteorological instrumentation will continue to perform their 
function in exactly the same way.
    The proposed changes do not affect any of the assumptions used 
in the accident analysis, nor do they affect any operability 
requirements for equipment important to plant safety. Therefore, the 
proposed changes will not result in a significant reduction in the 
margin of safety as defined in the Bases for Technical 
Specifications covered in this License Amendment Request.

    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 Section Chief: James W. Clifford.

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

    Date of amendment request: September 11, 1998, as supplemented by 
letter dated January 14, 1999.
    Description of amendment request: The proposed amendments would 
change the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to revise TS 6.8.4f., ``Containment 
Polar and Turbine Building Cranes,'' to control the operation of the 
containment polar cranes in jet impingement zones.
    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 Technical Specification (TS) 6.8.4f requirement to have a 
program that will ensure the position of the polar cranes precludes 
jet impingement from a postulated pipe rupture was previously 
evaluated in the NRC staff's safety evaluation for License 
Amendments (LA) 20 and 21. The proposed change is to control the 
operation of the containment polar cranes in jet impingement zones.
    PG&E evaluated a high energy line break (HELB) scenario for core 
damage frequency (CDF) considering operation of a polar crane. A 
postulated HELB would have to damage the crane or cause its load to 
drop in a manner that damages a component that exacerbates the HELB 
event and leads to core damage. The PRA evaluation for this scenario 
concluded the CDF is 1.6E-9 per year. It is not a significant 
increase in CDF compared to never operating the polar crane in jet 
impingement zones. The CDF for this scenario is nonrisk significant 
when compared to the industry standard threshold for risk 
significance for an operational evolution, which is 1E-6 per year. 
Several factors that further lower the risk of CDF include: 1) the 
movement of heavy loads is done in accordance with the DCPP Heavy 
Loads Program, which provides assurance that a dropped load would 
not lead to core damage, 2) the polar crane had been evaluated to 
withstand jet impingement loads without the seismic loads, and 3) 
the probability of simultaneous seismic and HELB events is low.
    Therefore, based on probabilistic considerations, the risk 
associated with this 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.
    Deterministic engineering methods required combining both the 
seismic and jet impingement loads to qualify Design Class I 
structures. The polar cranes were not originally qualified for these 
combined loads. This resulted in administrative controls that 
prohibited parking the polar cranes in jet impingement zones to 
preclude jet impingement loads from a postulated pipe rupture. The 
proposed change does not involve a physical change to the plant, but 
it does involve a change to the TS required program for containment 
polar crane operation.
    The proposed change is to control the operation of the 
containment polar cranes in jet impingement zones. It recognizes 
that there are jet (HELB) and target (polar crane) interactions. 
They were previously not considered for postulated jet impingement 
analyses because administrative controls prohibited parking the 
polar cranes in jet impingement zones. PG&E has evaluated jet 
impingement loads on the polar crane and determined it is able to 
withstand these loads without seismic loads. Based on this 
evaluation, the polar crane would not fail due to a HELB event. The 
movement of a heavy load would be done in accordance with the DCPP 
Heavy Loads Program. Thus, there would be no consequential failures 
that would lead to core damage.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The current TS 6.8.4f. requirement to have a program that will 
ensure the position of the polar cranes precludes jet impingement 
from a postulated pipe rupture was previously evaluated in the NRC 
staff's safety evaluation for LAs 20 and 21.
    The credible HELB sources that could impinge on the polar crane 
were identified and evaluated. The feedwater and main steam line 
steam generator nozzles are the only credible HELBs that could 
impinge upon the polar crane. The structural integrity of these 
lines was evaluated and determined to be of robust design.
    The margin of safety affected by the proposed change involves a 
comparison between the margin of safety afforded by no operation of 
the polar crane and operation that is controlled by procedures. The 
margin of safety in this case is the increase in risk for CDF caused 
by a scenario that postulates that operation of the polar crane 
would lead

[[Page 19562]]

to core damage. The risk for CDF has been evaluated and determined 
to be nonrisk significant. The CDF value is well below the industry 
standard threshold for acceptable risk for an operational evolution, 
which is 1E-6 per year.
    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 requests involve no significant hazards consideration.
    Local Public Document Room Location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: December 12, 1998.
    Description of amendment request: The proposed amendments would 
change the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to revise TS 6.9.1.8, ``Core Operating 
Limits Report,'' to allow use of NRC approved addenda to WCAP-10054-P-
A, ``Westinghouse Small Break ECCS Evaluation Model Using NOTRUMP 
Code,'' August 1985, to determine core operating limits.
    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.
    This change is administrative in nature in that it revises the 
Technical Specification (TS) Administrative Controls for the Core 
Operating Limits Report to include reference to NRC approved addenda 
to WCAP-10054-P-A, ``Westinghouse Small Break ECCS Evaluation Model 
Using the NOTRUMP Code,'' August 1985. The proposed change would 
allow the use of the analytical methods in WCAP-10054-P-A, Addendum 
2, Revision 1, Addendum to the Westinghouse Small Break ECCS.
    Evaluation Model Using the NOTRUMP Code: Safety Injection Into 
the Broken Loop and COSI Condensation Model,'' July 1997, and other 
NRC approved addenda to WCAP-10054-P-A to determine core operating 
limits for Diablo Canyon Power Plant (DCPP). Because plant operation 
will continue to be limited in accordance with cycle specific core 
operating limits that are established using an NRC approved 
methodology, NRC approved addenda to WCAP-10054-P-A are acceptable 
for use in determining DCPP Unit 1 and 2 cycle specific core 
operating limits.
    The change does not affect plant operation, or physically alter 
or change the function of structures, systems, or components 
required to mitigate the consequences of a design basis accident. In 
addition, it cannot initiate a transient or affect the probability 
of occurrence of any previously analyzed accident.
    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.
    The proposed change revises the TS to allow the use of NRC 
approved analytical methods in WCAP-10054-P-A, Addendum 2, Revision 
1, and other NRC approved addenda to WCAP-10054-P-A, to determine 
core operation limits. The change is consistent with the 
requirements of the TS, and does not affect plant operation, or 
physically alter or change the function of structures, systems, or 
components required to mitigate the consequences of a design basis 
accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change revises the TS to allow the use of the NRC 
approved analytical methods in WCAP-10054-P-A, Addendum 2, Revision 
1 and other NRC approved addenda to WCAP-10054-P-A, to determine 
core operating limits. The change is consistent with the 
requirements of the TS, and does not affect plant operation, or 
physically alter or change the function of structures, systems, or 
components required to mitigate the consequences of a design basis 
accident. The acceptance limits for the small break loss-of-coolant 
accident are not affected by this change and will continue to be 
met.
    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 requests involve no significant hazards consideration.
    Local Public Document Room Location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for Licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: Stuart A. Richards.

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: January 25, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to 
relocate the time restriction for movement of irradiated fuel and its 
related basis page from the TSs to the IP3 Final Safety Analysis Report 
(FSAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

Response

    Relocation (i.e., removal from TS) of TS 3.8.A.9 and its basis 
for the minimum time prior to movement of more than 76 irradiated 
fuel assemblies (267 hour limit) will not involve a significant 
increase in the probability or consequences of an accident since the 
relocation of the TS to administrative controls governed by 10 CFR 
50.59 (FSAR) does not affect the availability or function of fuel 
storage and handling equipment or the SFP [spent fuel pool] cooling 
system. The waiting time of 267 hours following plant shutdown 
before unloading more than 76 assemblies from the reactor is to 
ensure that the maximum SFP water temperature will be within design 
objectives as stated in the FSAR.
    The waiting time of 267 hours is not an initiator of an accident 
and the proposed change does not alter overall system operation, 
physical design, system configuration, or operational setpoints. 
There will be no significant increase in the consequences of an 
accident because the restricted movement time for irradiated fuel 
will continue to be administratively controlled under 10 CFR 50.59.
    The other TS of section 3.8.A (such as the remaining portion of 
3.8.A.9, and 3.8.A. 10) and the other controls ensure that doses 
from a postulated FHA are within 10 CFR 100 limits.

[[Page 19563]]

    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?

Response

    The basis for the waiting time of 267 hours following plant 
shutdown before unloading more than 76 assemblies from the reactor 
is to ensure that the maximum pool water temperature will be within 
design objectives as stated in the FSAR. Relocation of this waiting 
time of 267 hours for irradiated fuel will not create the 
possibility of a new or different kind of accident from any 
previously evaluated. The TS change will not create the possibility 
of a new or different kind of accident from any previously evaluated 
since it does not alter the administrative controls for fuel 
handling or the operation, physical design, system configuration, or 
operational setpoints for fuel handling and SFP cooling. The plant 
systems for fuel storage and handling, and SFP cooling are operated 
in the same manner as before and, consequently, the relocation does 
not introduce any new accident initiators or failure mechanisms and 
does not invalidate the existing FHA response. The minimum waiting 
time for movement of more than 76 irradiated fuel assemblies is not 
an accident initiator. The minimum waiting time will continue to be 
controlled under 10 CFR 50.59.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?

Response

    Relocation (i.e., removal from TS) of TS 3.8.A.9 and its basis 
for the waiting time of 267 hours following plant shutdown for 
irradiated fuel will not involve a significant reduction in margin 
of safety. The waiting time of 267 hours following plant shutdown 
before unloading more than 76 assemblies from the reactor is to 
ensure that the maximum SFP water temperature will be within design 
objectives as stated in the FSAR. The relocation is a change to the 
administrative controls that are used to limit the heat load on the 
SFP cooling system, and those administrative controls will be 
governed by 10 CFR 50.59. The manner in which fuel storage and 
handling is performed, and how the SFP cooling system is operated 
does not change and there is no change to physical design, system 
configuration, or operational setpoints. The other controls and the 
existing TS assure that dose from a postulated FHA are within 10 CFR 
100 limits. Previous analyses remain unchanged. The current TS does 
not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the 
Technical Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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: January 28, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to 
change the setpoint of the automatic reactor trip on turbine trip to at 
or below the P-8 setpoint.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    The addition of reactor trip on turbine trip at [greater than or 
equal to] 50% to the P-8 Permissive function versus its current 
setting of [greater than or equal to] 10%, as revised in TS section 
2.3.1.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, item 12, Table 4.1-1, 
item 21 and associated bases, does not significantly increase the 
probability or consequences of an accident previously evaluated. 
This additional function, change in reactor trip on turbine trip 
setpoint, does not cause the initiation of any accident, nor create 
any new credible limiting single failure, nor result in any event 
previously deemed incredible being made credible. The existing 
separation of the reactor and protection functions are not adversely 
impacted. In addition, the safety functions of safety related 
systems and component, which are related to accident mitigation, 
have not been altered. The change in the P-7 or P-8 circuitry does 
not directly initiate an accident. The consequences of accidents 
previously [evaluated] in the IP3 FSAR [final safety analysis 
report] are unaffected by this change because no change to any 
equipment response or accident mitigation scenario has resulted. 
There are no additional challenges to fission product barrier 
integrity. Therefore, the probability or consequences of an accident 
previously evaluated will not be increased.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    By adding the reactor trip on turbine trip at [greater than or 
equal to] 50% to the P-8 Permissive function and setpoint, versus 
its current setting of [greater than or equal to] 10% and revising 
TS sections 2.3. l.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, item 12, 
Table 4.1-1, item 21 and associated bases, does not create the 
possibility of a new or different kind of accident than any accident 
already evaluated. The additional function added to the P-8 
Permissive does not result in any event previously deemed incredible 
being made credible. No new accident scenarios, failure mechanisms, 
or limiting single failures are introduced as a result of this 
change. In addition, the safety functions of safety related systems 
and components, which are related to accident mitigation, have not 
been altered. Therefore, the possibility of a new or different kind 
of accident is not created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    The addition of the reactor trip on turbine trip at [greater 
than or equal to] 50% to the P-8 Permissive function, versus its 
current setting of [greater than or equal to] 10% and associated 
changes to TS Sections 2.3. l.C.(3), 2.3.2.A, 2.3.2.B, Table 3.5-2, 
item 12, Table 4.1-1, item 21 and the associated bases, will have no 
effect on the availability, operability or performance of the 
safety-related systems and components and does not affect the plant 
TS requirements. The current licensing basis safety analyses for IP3 
remain bounding with the modification to the P-8 Permissive 
function; therefore, the margin of safety as defined in the TS is 
not reduced. The change to the IP3 TS 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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: January 28, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 Technical Specifications (TSs) proposes to reduce 
the number of Emergency Diesel Generators (EDGs) required to be 
operable during cold shutdown from 2 to 1 under certain conditions.
    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

[[Page 19564]]

probability or consequences of an accident previously [evaluated]?

Response

    No. The equipment, which is affected by the proposed Technical 
Specification change, is not an initiator to those accidents 
postulated to occur during Cold Shutdown or Refueling operating 
conditions. A comprehensive systems review and EDG loading 
electrical analysis has demonstrated the ability of those shutdown 
support systems, necessary to provide safe shutdown needs, to 
perform their accident mitigation functions for the postulated 
accidents during Cold Shutdown and Refueling conditions. One EDG can 
support the necessary electrical loads required in Cold Shutdown and 
Refueling in the event of postulated accidents along with a LOOP 
[loss of offsite power] in the time frame required to prevent 
reactor core/cavity/SFP [spent fuel pool] heatup concerns. This EDG 
support relies upon existing plant designed manual closure of 480VAC 
EDS [electrical distribution system] bus tie breakers to allow a 
single EDG to pick up other 480VAC EDS bus loads, such as supplying 
an RHR [residual heat removal] pump and SFP cooling pump, located on 
480VAC EDS buses 3A, 5A, or 6A. Together, operability of the 
required offsite circuit(s) and one EDG ensures the availability of 
sufficient AC sources to operate the unit in a safe manner and to 
mitigate the consequences of postulated accidents during shutdown 
(e.g., Fuel Handling Accidents). Action statements provide prompt, 
specific guidance to ensure sufficiently conservative plant response 
should the expected EDG power supply not be available. These Action 
Statements are similar to those in the STS [Standard Technical 
Specifications]. Therefore, the proposed license amendment (i.e., 
changes to 3.7.F.4 and the added sections of 3.7.F.5 & 3.7.F.6) does 
not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?

Response

    No. The proposed license amendment does not involve any physical 
changes to plant systems or component set points. The use of 480VAC 
EDS bus tie breakers to power loads from an energized 480VAC bus is 
part of present plant design and included within the present LOOP 
Off-Normal operating procedures when the reactor is in Cold Shutdown 
operating conditions. As discussed in the Standard Technical 
Specifications, NUREG 1431, during plant shutdown with one EDG, it 
is not required to assume a single failure and concurrent loss of 
all offsite or all onsite power. Worst case bounding events are 
deemed not credible in Cold Shutdown and Refueling conditions 
because the energy contained within the reactor pressure boundary, 
reactor coolant temperature and pressure, and the corresponding 
stresses result in the probabilities of occurrence being 
significantly reduced or eliminated, and ultimately result in 
minimal consequences. The lone EDG is capable of accepting and 
starting required loads within the assumed loading sequence 
intervals and continue to operate until offsite power can be 
provided to the 480VAC EDS buses. Action statements provide prompt, 
specific guidance to ensure sufficiently conservative plant response 
should the expected EDG power supply not be available. These action 
statements are similar to those in the STS. Therefore, the proposed 
license amendment (i.e., changes to 3.7.F.4 and added sections 
3.71.5 & 3.7.F.6) does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?

Response

    No. The electrical power system specifications support the 
equipment required to be operable, commensurate with the current 
level of safety, including the equipment requiring an EDG backed 
power source. The design review results demonstrate that operation 
in the conditions of Cold Shutdown and Refueling, in accordance with 
the proposed Technical Specification change, is acceptable from an 
accident mitigation standpoint. The basic system functions in Cold 
Shutdown and Refueling operating conditions are not changed. One EDG 
can supply the necessary electrical power needs during these plant 
operating conditions, and in the time frame required to prevent 
reactor core/cavity/SFP heatup concerns, with sufficient ``kw 
loading'' to spare. The analysis conducted shows that the systems 
are capable of performing their design basis functions. Applicable 
safety analysis in the Standard Technical Specifications, NUREG 
1431, discusses these system requirements as well (i.e., it is not 
required to assume a single failure and concurrent loss of all 
offsite or all onsite power). Action statements, similar to those in 
the Standard Technical Specifications, provide prompt, specific 
guidance to ensure sufficiently conservative plant response should 
the expected EDG power supply not be available. On this basis, the 
proposed license amendment (i.e., changes to 3.7.F.4 and added 
sections 3.7.F.5 & 3.7.F.6) does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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: January 29, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to 
change the allowable indicated control rod misalignment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

Response

    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS Bases is not reduced in any of the IP3 licensing 
basis accident analysis. Increasing the magnitude of allowed control 
rod indicated misalignment (in section 3.10.5) is not a contributor 
to the mechanistic cause of an accident evaluated in the FSAR [Final 
Safety Analysis Report]. Neither the rod control system nor the rod 
position indicator function is being altered. Therefore, the 
probability of an accident previously evaluated has not 
significantly increased. Because design limitations continue to be 
met, and the integrity of the reactor coolant system pressure 
boundary is not challenged, the assumptions employed in the 
calculation of the offsite radiological doses remain valid.
    Therefore, the consequences of an accident previously evaluated 
will not be significantly increased.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?

Response

    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS Bases is not reduced in any of the IP3 licensing 
basis accident analysis. Increasing the magnitude of allowed control 
rod indicated misalignment is not a contributor to the mechanistic 
cause of any accident. Neither the rod control system nor the rod 
position indicator function is being altered. Therefore, an accident 
which is new or different than any previously evaluated will not be 
created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?

Response

    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has

[[Page 19565]]

determined that all pertinent licensing basis acceptance criteria 
have been met, and the margin of safety as defined in the TS Bases 
is not reduced in any of the IP3 licensing basis accident analysis 
based on the changes to safety analyses input parameter values as 
discussed in WCAP-14668. Since the evaluations in Section 3.0 of 
WCAP-14668 demonstrate that all applicable acceptance criteria 
continue to be met, the proposed change will not involve a 
significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

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: January 29, 1999.
    Description of amendment request: This application for amendment to 
the Indian Point 3 (IP3) Technical Specifications (TSs) proposes to 
change the allowable indicated control rod misalignment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

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

Response

    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS Bases is not reduced in any of the IP3 licensing 
basis accident analysis. Increasing the magnitude of allowed control 
rod indicated misalignment (in Section 3.10.5) is not a contributor 
to the mechanistic cause of an accident evaluated in the FSAR [Final 
Safety Analysis Report]. Neither the rod control system nor the rod 
position indicator function is being altered. Therefore, the 
probability of an accident previously evaluated has not 
significantly increased. Because design limitations continue to be 
met, and the integrity of the reactor coolant system pressure 
boundary is not challenged, the assumptions employed in the 
calculation of the offsite radiological doses remain valid.
    Therefore, the consequences of an accident previously evaluated 
will not be significantly increased.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?

Response

    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS Bases is not reduced in any of the IP3 licensing 
basis accident analysis. Increasing the magnitude of allowed control 
rod indicated misalignment is not a contributor to the mechanistic 
cause of any accident. Neither the rod control system nor the rod 
position indicator function is being altered. Therefore, an accident 
which is new or different than any previously evaluated will not be 
created.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?

Response

    No. Based on the Westinghouse evaluation in WCAP-14668, the 
Authority has determined that all pertinent licensing basis 
acceptance criteria have been met, and the margin of safety as 
defined in the TS Bases is not reduced in any of the IP3 licensing 
basis accident analysis based on the changes to safety analyses 
input parameter values as discussed in WCAP-14668. Since the 
evaluations in Section 3.0 of WCAP-14668 demonstrate that all 
applicable acceptance criteria continue to be met, the proposed 
change will not involve a significant reduction in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. David E. Blabey, 10 Columbus Circle, New 
York, New York 10019.
    NRC Section Chief: S. Singh Bajwa.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 22, 1999.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.7.1.6, ``Atmospheric Steam Relief 
Valves,'' and add a new TS for atmospheric steam relief valve 
instrumentation, to ensure that the automatic feature of the steam 
generator power-operated relief valve (i.e., atmospheric steam relief 
valves) remains operable during Modes 1 and 2. In addition, the 
proposed change would add an associated surveillance requiring that a 
channel calibration on the steam generator power-operated relief valve 
be performed every 18 months.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The methodologies used in the accident analyses remain 
unchanged. The automatic actuation of the Steam Generator Power 
Operated Relief Valves is not a new design feature. The effects of 
the inadvertent opening of a Steam Generator Power Operated Relief 
Valve are currently analyzed as described in Section 15.1.4 of the 
Updated Final Safety Analysis Report. The radiological consequences 
for the Small Break Loss of Coolant Accident (SBLOCA) event 
presented in the Updated Final Safety Analysis Report remain 
unchanged. The calculated Peak Clad Temperature is 1849 deg.F 
remaining substantially below the 2200 deg.F acceptance limit of 10 
CFR 50.46. Although the manual control specification is relocated 
from Specification 3.7.1.6 to the new instrumentation specification, 
the limiting condition for operation, applicability and action 
statements for manual controls remain unchanged. Therefore no 
increase in the probability or consequences of any accident 
previously evaluated will occur.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The automatic actuation of the Steam Generator Power Operated 
Relief Valves is not an accident initiator for the SBLOCA event. The 
automatic actuation of the Steam Generator Power Operated Relief 
Valves currently exists at the South Texas Project and is not a new 
design feature. The description of the Steam Generator Power 
Operated Relief Valves currently exists in the Updated Final Safety 
Analysis Report. This change does not represent a change to the 
facility and does not affect the safety functions and reliability of 
systems, structures, or components in any new manner. Operating 
procedures have a temporary administrative control to ensure the 
automatic actuation of the Steam Generator Power Operated Relief 
Valves remains operable in Modes 1 and 2. This condition will become 
permanent with the approval of this Technical Specification 
Amendment proposal. Although the manual control specification is 
relocated from Specification 3.7.1.6 to the new instrumentation 
specification, the limiting condition for operation, applicability 
and

[[Page 19566]]

action statements for manual controls remain unchanged. Since the 
automatic actuation of the Steam Generator Power Operated Relief 
Valves is not an accident initiator and is not a new design feature 
to the facility, no possibility exists for a new or different kind 
of accident from those previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change results in the calculated Peak Clad 
Temperature of 1849 deg.F remaining well below the acceptance limit 
of 10 CFR 50.46 and comparable to the results currently described in 
the Updated Final Safety Analysis Report. Therefore, the change does 
not involve a significant reduction in a margin of safety.
    Based on the above, the South Texas Project has evaluated the 
proposed Technical Specification change and determined it does not 
represent a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Section Chief: Robert A. Gramm.

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

    Date of application for amendments: March 19, 1999 (TS 99-01).
    Brief description of amendments: The proposed amendments would 
change the SQN Technical Specifications (TS) for Operating Licenses 
DPR-77 (Unit 1) and DPR-79 (Unit 2) by relocating TS Sections 3.8.3.1, 
3.8.3.2, and 3.8.3.3 to the SQN Technical Requirements Manual. These 
sections provide requirements for electrical overcurrent isolation 
devices.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed revision to the TS relocates the requirements for 
SQN's electrical equipment protective devices without changing the 
current requirements. TVA does not consider these devices to be the 
source of any accident; therefore, this administrative relocation of 
the requirements will not increase the possibility of an accident. 
SQN's electrical equipment protective devices will continue to 
provide fault protection for circuits and equipment. Changes to the 
relocated requirements will be processed, in accordance with 10 CFR 
50.59, to ensure changes are not implemented that would reduce the 
functionality or introduce an unreviewed safety question to SQN's 
electrical equipment devices. Therefore, the proposed relocation of 
the TS requirements for electrical equipment protective devices will 
not increase the consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    SQN's electrical equipment protective devices ensure proper 
operation of plant equipment. These devices are not associated with 
accident mitigation or previously evaluated accidents and would not 
be the initiator of any new or different kind of accident. The 
proposed change does not alter the current functions of these 
devices, therefore, this proposed change will not create the 
possibility of a new or different kind of accident.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The requirements for SQN's electrical equipment protective 
devices are unchanged by the proposed relocation of the requirements 
to the SQN Technical Requirements Manual. The function of these 
devices and the surveillance testing to ensure operability of these 
devices remains unchanged. Any future changes to these requirements 
will be evaluated, in accordance with 10 CFR 50.59, to ensure 
acceptability and NRC review as required. Accordingly, the proposed 
change will not result in a reduction in a margin of safety.

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

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendment: December 16, 1997, as 
supplemented August 31, and December 7, 1998.
    Brief description of amendment: This amendment changes Technical 
Specification 4.7.1.2.1.a.2.a, Auxiliary Feedwater (AFW) System 
Surveillance Requirements, by changing the differential pressure and 
flow requirements of the steam turbine-driven AFW pump to allow testing 
of the pump at a lower speed.
    Date of issuance: April 1, 1999.
    Effective date: April 1, 1999.
    Amendment No.: 87.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6981).

[[Page 19567]]

    The August 31, and December 7, 1998, submittals contained 
clarifying information only, and did not change the initial no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendment: September 1, 1998, as 
supplemented on March 19, 1999.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) \3/4\.9.11, ``Water Level--New and Spent Fuel 
Pools,'' and its associated Bases by requiring 23 feet of water above 
the top of fuel rods within irradiated fuel assemblies seated in the 
storage racks.
    Date of issuance: April 8, 1999.
    Effective date: April 8, 1999.
    Amendment No.: 88.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 23, 1998 (63 
FR 50935).
    The March 19, 1999, submittal contained clarifying information 
only, and did not change the initial no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 1999.
    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 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of application for amendments: November 25, 1998.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TS) to support on-line replacement of the 
Braidwood, Unit 2, vital batteries.
    Date of issuance: March 26, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 99 and 99.
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9185).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481.

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

    Date of application for amendments: January 21, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) by relocating TS Section \3/4\.6.I, 
``Primary System Boundary-Chemistry'' and associated bases to the 
Updated Final Safety Analysis Report (UFSAR) and to applicable plant 
procedures.
    Date of issuance: March 31, 1999.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 187 and 184.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9186).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 31, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

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

    Date of application for amendments: January 28, 1999.
    Brief description of amendments: The amendments revised Technical 
Specifications Section 3.7.13, ``Fuel Handling Ventilation Exhaust 
System,'' and associated Bases to correct discrepancies between the 
current design and this section.
    Date of issuance: March 26, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-176; Unit 2-168.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1999 (64 
FR 9187).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South 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 of amendments: October 15, 1998, as 
supplemented December 15, 1998, and January 11 and 21, 1999.
    Brief description of amendments: The amendments revised the 
Technical Specifications (TSs) to change the heatup, cooldown, and 
inservice test limitations for the reactor coolant system of each unit 
to a maximum of 26 effective full-power years. The amendments also 
revise the TSs for low temperature overpressure protection to reflect 
the revised pressure-temperature limits of the reactor vessels.
    Date of Issuance: March 30, 1999.
    Effective date: As of the date of issuance to be implemented within 
90 days from the date of issuance.
    Amendment Nos.: Unit 1-302; Unit 2-302; Unit 3-302.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66592).
    The December 15, 1998, and January 11 and 21, 1999, letters 
provided clarifying information that did not change the scope of the 
original Federal Register notice and the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 30, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina.

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

    Date of application for amendments: October 15, 1998, as 
supplemented

[[Page 19568]]

December 14, 1998, February 18, 1999, and February 23, 1999.
    Brief description of amendments: These amendments made several 
changes that are administrative in nature. The changes (1) made 
editorial changes that delete obsolete material or material adequately 
described elsewhere, changed action statement numbers, updated 
technical specification (TSs) index pages, and made changes to be 
consistent with the guidance provided in the improved standard 
technical specifications for Westinghouse reactors (NUREG-1431, 
Revision 1); (2) deleted reporting requirements that are duplicated in 
various sections of Title 10 of the Code of Federal Regulations; and 
(3) relocated the requirement for meteorological monitoring 
instrumentation from the TSs to the Licensing Requirements Manual.
    The February 18, 1999, and February 23, 1999, letters withdrew a 
portion of the amendment request that would have deleted the 
description of the site exclusion boundary from the TSs. The 
description of the site exclusion boundary will remain in the TS.
    Date of issuance: March 26, 1999.
    Effective date: Units 1 and 2, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 220 and 97.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications and licenses.
    Date of initial notice in Federal Register: November 18, 1998 (63 
FR 64111).
    The December 14, 1998, February 18, 1999, and February 23, 1999, 
letters did not change the initial proposed no significant hazards 
consideration determination or expand the amendment request beyond the 
scope of the initial notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley Power 
Station, Unit 2, Shippingport, Pennsylvania

    Date of application for amendment: March 10, 1997, as supplemented 
July 28, 1997, September 17, 1997, April 30, 1998, January 29, 1999, 
and February 26, 1999.
    Brief description of amendment: The amendment modifies Technical 
Specification 3/4.4.5, ``Steam Generators,'' and its associated Bases 
and adds a new license condition to Appendix D 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 Beaver 
Valley Power Station, Unit 2.
    Date of issuance: March 26, 1999.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No: 98.
    Facility Operating License No. NPF-73. Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: April 23, 1997 (62 FR 
19829).
    The July 28, 1997, September 17, 1997, April 30, 1998, January 29, 
1999, and February 26, 1999, 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 March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

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

    Date of amendment request: January 12, 1999, supersedes application 
dated May 31, 1996.
    Brief description of amendment: The amendment adds an additional 
required action to the Limiting Condition for Operation (LCO) 3.9.1, 
``Refueling Equipment Interlocks,'' of the RBS Technical 
Specifications. The additional action will allow an alternative to the 
current action for one or more inoperable refueling equipment 
interlocks. The current action is to ``suspend in-vessel fuel movement 
with equipment associated with the inoperable interlock(s).'' The 
alternative action will be to (1) insert a control rod withdrawal 
block, and (2) verify all control rods are fully inserted in core cells 
containing one or more fuel assemblies. The amendment also revised the 
Bases for LCO 3.9.1 actions to describe the alternative action.
    Date of issuance: March 26, 1999.
    Effective date: March 26, 1999.
    Amendment No.: 104.
    Facility Operating License No. NPF-47: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 10, 1999 (64 
FR 6695).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

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

    Date of application for amendment: November 30, 1998.
    Brief description of amendment: The amendment changes Technical 
Specification 3.1.2, ``Liquid Poison System,'' and its associated Bases 
to correct the required concentration and volume of boron solution.
    Date of issuance: April 2, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 166.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998 (63 
FR 71970).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: April 1, 1998, as supplemented 
May 29, June 26, and August 4, 1998.
    Brief description of amendment: The amendment revises the Millstone 
Unit 3 final safety analysis report (FSAR) by adding a new sump pump 
subsystem to address groundwater inleakage through the containment 
basemat.
    Date of issuance: March 17, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days from the date of issuance.

[[Page 19569]]

    Amendment No.: 168.
    Facility Operating License No. NPF-49: Amendment authorized changes 
to the FSAR.
    Date of initial notice in Federal Register: April 22, 1998 (63 FR 
19974).
    The May 29, June 26, and August 4, 1998, letters provided 
clarifying information that did not change the scope of the April 1, 
1998, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment, state 
consultation, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated March 17, 
1999.
    No significant hazards consideration comments received: No public 
comments received. A petition to intervene was received from the 
Citizens Regulatory Commission that was dismissed and terminated by the 
NRC Atomic Safety Licensing Board (LBP-98-22).
    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.

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

    Date of application for amendments: February 4, 1998, as revised 
September 29, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications surveillance requirements concerning secondary 
containment doors.
    Date of issuance: April 7, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendments Nos.: 227 and 230.
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 14, 1998 (63 FR 
38202).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 7, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (Regional Depository) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
PA 17105.

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

    Date of application for amendment: October 22, 1998.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) 4.8.2.1.b.3 to increase the minimum battery 
electrolyte temperature limit from 60 deg.F to 72 deg.F. This change 
resolves a discrepancy in the electrolyte temperature assumed in the 
Class 1E battery sizing calculations versus the limit specified in the 
TSs.
    Date of issuance: March 25, 1999.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 118.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66602).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 25, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of application for amendments: December 4, 1998.
    Brief description of amendments: The amendments make two changes to 
the TS. The first change revises the Unit 1 TS Section 2.1.1.2 to 
delete the footnote that specifies that the Safety Limit Minimum 
Critical Power Ratios are for Cycle 18 only. The second change revises 
the TS for both units by deleting Section 5.6.5.b.2) and incorporating 
Section 5.6.5.b.1) into Section 5.6.5.b.
    Date of issuance: April 1, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-215; Unit 2-156.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1999 (64 FR 
4161).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of application for amendments: October 15, 1998, as 
supplemented by letter dated November 11, 1998.
    Brief description of amendments: The amendments change the Vogtle 
Electric Generating Plant Unit 1 and 2 Facility Operating Licenses to 
delete or modify certain license conditions that have become obsolete 
or inappropriate. In addition, the Technical Specifications and Bases 
are reissued to reflect new word processing software.
    Date of issuance: March 26, 1999.
    Effective date: As of the date of issuance to be implemented within 
30 days from the date of issuance.
    Amendment Nos.: Unit 1-107; Unit 2-85.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised Facility Operating Licenses and the Technical Specifications.
    Date of initial notice in Federal Register: December 2, 1998 (63 FR 
66602).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 26, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia

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

    Date of applications for amendment: October 31, 1997, as 
supplemented by letter dated September 29, 1998, and application dated 
July 30, 1998.
    Brief description of amendment: The amendment revised Tables 3.3-3, 
3.3-4, and 4.3-2 of the technical specifications regarding the 
engineered safety feature actuation system (ESFAS) Functional Unit 6.f, 
``Loss of Offisite Power--Start Turbine-Driven Pump,'' by establishing 
separate requirements for the analog and digital portions of the 
associated circuit. The amendment also adds a note to TS Table 4.3-2 to 
clarify that the verification of time delays associated

[[Page 19570]]

with ESFAS Functional Units 8.a and 8.b, ``Loss of Power,'' is only 
performed as part of the channel calibration.
    Date of issuance: April 2, 1999.
    Effective date: April 2, 1999, to be implemented within 30 days of 
the date of issuance.
    Amendment No.: 130.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 16, 1998 (63 
FR 69348).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Elmer Ellis Library, 
University of Missouri, Columbia Missouri 65201.

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

    Date of application for amendment: November 18, 1998, as 
supplemented with additional information by letters dated March 1, 
1999, and March 9, 1999.
    Brief description of amendment: The amendment revises the pressure/
temperature limits and the low-temperature overpressure protection 
requirements in the facility technical specifications.
    Date of issuance: April 1, 1999.
    Effective date: April 1, 1999.
    Amendment No.: 144.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 30, 1998. 
(63FR71978)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 1999.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, WI 54311-7001.

    Dated at Rockville, Maryland, this 14th day of April 1999.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 99-9839 Filed 4-20-99; 8:45 am]
BILLING CODE 7590-01-P