[Federal Register Volume 61, Number 224 (Tuesday, November 19, 1996)]
[Notices]
[Pages 58900-58910]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-29584]


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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 October 28, 1996, through November 7, 1996. 
The last biweekly notice was published on November 6, 1996.

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 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By December 20, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first

[[Page 58901]]

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: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
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.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of amendment request: October 2, 1996
    Description of amendment request: The amendment would change 
Figures 3.1.A-1, 3.1.A-2, and 3.1.A-3, Section 3.1.B and its Bases, 
Figures 3.1.B-1 and 3.1.B-2, and the Bases of Section 4.3 and Figure 
4.3-1 of the Technical Specifications by providing new pressure/
temperature limit curves.
    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:
    Neither the probability nor the consequences of an accident 
previously analyzed is increased due to the proposed changes. The 
adjusted reference temperature of the most limiting beltline 
material was used to correct the pressure-temperature (P-T) curves 
to account for irradiation effects. Thus, the operating limits are 
adjusted to incorporate both the initial fracture toughness 
conservatism present when the reactor vessel was new and the effect 
of fluence. The adjusted reference temperature calculations were 
performed utilizing the guidance contained in RG [Regulatory Guide] 
1.99, Revision 2. Overpressure Protection System (OPS) curves and 
tables were regenerated to be consistent with the new P-T curves. 
The updated curves provide assurance that brittle fracture of the 
reactor vessel is prevented.
    2) Does the proposed license amendment create the possibility of 
a new or different kind of accident from any previously evaluated?
    Response:
    The updated P-T and OPS limits will not create the possibility 
of a new or different kind of accident. The revised operating limits 
merely update the existing limits by taking into account the effects 
of radiation embrittlement, utilizing criteria defined in RG 1.99, 
Revision 2. The updated curves are conservatively adjusted to 
account for the effect of irradiation on the limiting reactor vessel 
material.
    No change is being made to the way the pressure-temperature 
limits provide plant protection. No new modes of operation are 
involved. Incorporating this amendment does not necessitate physical 
alteration of the plant.
    3) Does the proposed amendment involve a significant reduction 
in the margin of safety?
    Response:
    The proposed amendment does not involve a significant reduction 
in the margin of safety. The pressure-temperature operating limits 
and OPS setpoints are designed to maintain an appropriate margin of 
safety. The required margin is specified in ASME [American Society 
of Mechanical Engineers] Boiler and Pressure Vessel Code, Section 
III, Appendix G and 10 CFR [Part] 50 Appendix G. The revised curves 
are based on the latest NRC guidelines along with actual neutron 
fluence data for the reactor vessel. The new limits retain a margin 
of safety equivalent to the original margin when the vessel was new 
and the fracture toughness was slightly greater. The new operating 
limits account for irradiation embrittlement effects, thereby 
maintaining a conservative margin of safety.
    The removal of the pressure-temperature limits for criticality 
does not reduce the plant safety margin because these limits are 
conservatively encompassed and bounded by the requirements of the 
proposed Technical Specification 3.1.C.2.
    The NRC staff has reviewed the licensee's analysis and, based on 
this

[[Page 58902]]

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 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: S. Singh Bajwa, 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: September 6, 1996
    Description of amendment request: The proposed amendments would 
revise Item 7.c of BVPS-1 Technical Specifications (TSs) Table 3.3-3 
and Item 7.d of BVPS-2 TS Table 3.3-3 to reflect that a safety 
injection (SI) signal starts all auxiliary feedwater (AFW) pumps. The 
notation on BVPS-1 TS Table 3.3-5 would be revised to state that the 
response time is for all AFW pumps on all SI signal starts. Items 7.d 
of BVPS-2 TS Tables 3.3-4 and 4.3-2 would be revised to reflect that an 
SI signal starts all AFW pumps.
    The proposed amendments would also revise and reformat TSs 3/
4.7.1.2 to more closely resemble the wording contained in the NRC's 
``Standard Technical Specifications Westinghouse Plants,'' (NUREG-1431, 
Revision 1). These changes would require three AFW trains to be 
operable and would provide what constitutes an operable train. The mode 
applicability for these TSs would expand to include Mode 4 when the 
steam generator(s) is relied upon for heat removal.
    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 revisions to reflect that a Safety Injection (SI) 
signal starts the turbine driven Auxiliary Feedwater (AFW) pump, in 
addition to both motor driven AFW pumps, will ensure that plant 
operability requirements for the AFW system actuation signals are 
maintained at a level consistent with current safety analyses. The 
proposed revisions to Limiting Condition for Operation (LCO) 3.7.1.2 
will require that the AFW pumps and associated flow paths are 
maintained operable to ensure that the AFW system can mitigate the 
consequences of a Design Basis Accident (DBA) with a loss of normal 
feedwater. The addition of the Mode 4 applicability will ensure that 
a safety related source of cooling water is available to remove 
decay heat.
    The proposed change will ensure that the plant is placed in Mode 
4 when the number of operable feedwater injection headers is 
insufficient to ensure that at least two steam generators are 
supplied during a feedline break accident.
    The proposed addition of footnote (2) to action statement ``c'' 
will limit plant thermal cycles following a refueling outage due to 
turbine driven AFW pump inoperability. During the additional time 
period provided by footnote (2) to reach Hot Shutdown, the two 
remaining motor driven AFW pumps will provide sufficient flow to the 
steam generators to mitigate the consequences of a DBA assuming no 
single failures during this time period. Since there is negligible 
decay heat following a refueling outage prior to entry into Mode 2, 
the performance capabilities of the two remaining motor driven AFW 
pumps to remove decay heat will not be challenged.
    Changing the AFW pump surveillance test frequencies for Beaver 
Valley Power Station (BVPS) Unit No. 2 to quarterly, as specified in 
the Inservice Testing (IST) Program, will continue to assure that 
the AFW system will be capable of performing its intended functions.
    The proposed change to the current Surveillance Requirement 
4.7.1.2, for BVPS Unit No. 2 only, will not lower the pump 
performance operability criteria for the AFW pumps. The required 
values for developed pump head and flow will continue to satisfy 
accident mitigation requirements and will be maintained and 
controlled in the BVPS Unit No. 2 IST Program. Future changes to the 
AFW pump head and flow requirements will be made under the 10 CFR 
50.59 process to ensure that the AFW design requirement to remove 
sufficient decay heat continues to be met.
    Based on the above factors, the probability of an accident 
previously evaluated is not significantly increased.
    The proposed changes do not affect the ability of the AFW system 
to perform as assumed in the safety analyses. The proposed changes 
will not result in any additional challenges to plant equipment. 
Because the plant design limits will continue to be met, the fuel 
and reactor coolant system pressure boundary integrity is not 
challenged for the assumptions employed in the calculation of the 
offsite radiological doses. The additional time to reach Mode 4 from 
Mode 3 provided by footnote (2) does not result in increased 
radiological consequences. The potential for a radioactivity release 
due to the uncontrolled heatup of [the] reactor coolant system[s] 
are enveloped by the releases postulated in the DBA Loss of Coolant 
Accident (LOCA) analysis in the Updated Final Safety Analysis 
Report. The DBA LOCA analysis assumes 102% power operation prior to 
the event and assumes that core melt occurs. Therefore, there is no 
increase in the radiological consequences as a result of allowing 
additional time to repair/test the turbine driven AFW pump. Hence, 
the consequences of a DBA previously evaluated is not significantly 
increased.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change does not alter the method of operating the 
plant. The AFW system is an accident mitigation system and is 
normally in standby. System operation is initiated in response to a 
DBA. The AFW pumps will continue to provide sufficient flow to 
mitigate the consequences of a DBA. AFW operation continues to 
fulfill the safety function for which it was designed and no changes 
to plant equipment will occur. As a result, an accident which is new 
or different than any already evaluated in the Updated Final Safety 
Analysis Report will not be created due to this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes will not affect the heat removal capability 
of the AFW system to a value less than assumed in the safety 
analysis. The proposed changes will not result in any additional 
challenges to the plant equipment including the fuel and reactor 
coolant system pressure boundary. The additional time period to 
reach Hot Shutdown provided by footnote (2) will not significantly 
reduce the decay heat removal capability provided by the AFW system. 
The two remaining motor driven AFW pumps will continue to provide 
sufficient flow to the steam generators as assumed in the safety 
analysis to mitigate the consequences of a DBA assuming no single 
failure during this time period. The plant will continue to operate 
within the bounds of the safety analysis.
    The AFW system will continue to be tested in a manner and at a 
frequency which will ensure acceptable system performance should it 
be relied upon to remove decay heat following a DBA.
    The AFW pumps' performance requirements will continue to be 
controlled in a manner to ensure safety analysis assumptions are 
met.
    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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

[[Page 58903]]

    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: July 25, 1996
    Description of amendment request: The proposed change modifies 
Technical Specification (TS) 3/4.7.4 Ultimate Heat Sink (UHS) by 
incorporating more restrictive fan operability requirements and lower 
basin temperature. Several other administrative changes are 
incorporated to improve the humanfactors associated with this TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?

    Response: No
    The proposed change modifies the UHS TS by revising [Wet Cooling 
Tower] WCT basin water temperature from less than or equal to 95 
Degrees Fahrenheit to less than or equal to 89 Degrees Fahrenheit 
and incorporating more restrictive cooling tower fan operability 
requirements. These changes are necessary to adequately preserve the 
assumptions and limits of the revised UHS design basis calculations. 
These calculations conclude that the UHS is capable of dissipating 
the maximum peak heat load resulting from the limiting design bases 
accident (i.e., large break LOCA) and the most severe natural 
phenomena (i.e., tornado event). Other changes are purely 
administrative in nature. The proposed change does not directly 
affect any material condition of the plant that could directly 
contribute to causing an accident. The proposed change ensures that 
the mitigating effects of the UHS will be consistent with the design 
basis analysis. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?

    Response: No
    The proposed change modifies the UHS TS to be consistent with 
revised design basis calculations. These new calculations adjust 
margin to incorporate an additional allowance for fouling in the 
[Component Cooling Water] CCW heat exchangers and more restrictive 
UHS minimum fan requirements that were not adequately addressed in 
the initial design basis. This change also incorporates 
administrative changes that are intended to improve the application 
and use of this specification. The proposed change will not alter 
the operation of the plant or the manner in which the plant is 
operated. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?

    Response: No
    The proposed change modifies the UHS TS by revising WCT basin 
water temperature from less than or equal to 95 Degrees Fahrenheit 
to less than or equal to 89 Degrees Fahrenheit and incorporating 
more restrictive cooling tower fan operability requirements. 
Modifying the UHS meteorological design bases reduced WCT basin 
temperature requirement for operability, thus, providing an 
allowance for fouling in the CCW heat exchangers. The proposed 
change better preserves the margin of safety by ensuring that the 
UHS will maintain the CCW accident analysis temperature limit of 115 
Degrees Fahrenheit. Increased cooling tower fan operability 
requirements will ensure that the expected cooling efficiency is 
actually available and not unknowingly degraded due to fouling. 
Other changes requested herein are purely administrative in nature, 
do no affect safety margins and intended to improve the use and 
application of this specification. Therefore, the proposed change 
will not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of amendment request: October 4, 1996
    Description of amendment request: The proposed amendments would 
incorporate the requirements necessary to change the basis for 
prevention of criticality in the fuel storage pool. This change would 
eliminate credit for Boraflex as a neutron absorbing material in the 
fuel storage pool criticality analysis and would support the storage of 
fuel with enrichments up to and including 5.0 weight percent U-235 
rather than the current value of 4.5 weight percent U-235.
    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.
    There is no increase in the radiological consequences of 
accidents previously evaluated in the Vogtle FSAR [Final Safety 
Analysis Report] with the use of 5.0 weight percent U-235 fuel. 
Increasing the enrichment up to and including 5.0 weight percent U-
235 affects the radiological source terms and subsequently the 
potential releases both normal and accidental. Evaluations performed 
(WCAP-12610-P-A, Reference 6) considered the source term, gap 
fraction, normal operating plant releases and the accident doses for 
a maximum fuel enrichment of 5.0 weight percent U-235. It was 
concluded that operating with and storing fuel with 5.0 weight 
percent U-235 enrichment may result in minor increases in the normal 
annual releases of long half-life fission products that are not 
significant. Also, the radiological consequences of accidents are 
minimally affected due to the very small changes in the core 
inventory and the fact that the currently assumed gap fractions 
remain bounding.
    The use of the slightly higher enrichment for VEGP [Vogtle 
Electric Generating Plant] fuel will not result in burnups in excess 
of those currently allowed for VEGP. The cycle design methods and 
limits will remain the same as are currently licensed. Therefore the 
use of fuel with the higher enrichment is not expected to result in 
operating conditions outside those currently allowed for VEGP.
    There is no increase in the probability of a fuel assembly drop 
accident in the fuel storage pool when considering the presence of 
soluble boron in the pool water for criticality control. The 
handling of the fuel assemblies in the fuel storage pool has always 
been performed in borated water.
    Fuel assembly placement will be controlled pursuant to approved 
fuel handling procedures and will be in accordance with the spent 
fuel rack storage configuration limitations in the COLR [Core 
Operating Limit Report]. The consequences of a misplaced assembly 
have been included in the analysis supporting this revision to the 
Technical Specifications.
    There is no increase in the consequences of the accidental 
misloading of a spent fuel assembly into the fuel storage pool racks 
because criticality analyses demonstrate that

[[Page 58904]]

the pool will remain subcritical following an accidental misloading 
of an assembly even considering a dilution event. The proposed 
Technical Specifications and COLR limitations will ensure that an 
adequate fuel storage pool boron concentration will be maintained.
    There is no increase in the probability of the loss of normal 
cooling to the fuel storage pool water due to the presence of 
soluble boron in the pool water for subcriticality control, because 
a high concentration of soluble boron has been maintained in the 
fuel storage pool water.
    The loss of normal cooling to the fuel storage pool will cause 
an increase in the temperature of the fuel storage pool water. This 
will cause a decrease in water density which would normally result 
in an addition of negative reactivity. However, since Boraflex is 
not considered to be present, and the fuel storage pool water has a 
high concentration of boron, a density decrease causes a positive 
reactivity addition. The amount of soluble boron required to offset 
this postulated accident was evaluated for the allowed storage 
configurations. The amount of soluble boron necessary to mitigate 
these accidents and ensure that the Keff will be maintained 
less than or equal to 0.95 has been included in the fuel storage 
pool boron concentration. Because adequate soluble boron will be 
maintained in the pool water, the consequences of a loss of normal 
cooling to the fuel storage pool will not be increased.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will 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 analyzed.
    The potential for criticality accidents in the fuel storage pool 
are not new or different types of accidents. It has been reanalyzed 
in the Criticality Analysis report (Enclosure 5 [of the proposed 
amendment request]).
    Because soluble boron has been maintained in the fuel storage 
pool water, the possibility of a fuel storage pool dilution has 
previously existed. Therefore, the implementation of Technical 
Specification controls for the soluble boron will not create the 
possibility of a new or different kind of accidental pool dilution.
    With credit for soluble boron now a major factor in controlling 
criticality, an evaluation of fuel storage pool dilution events was 
completed. A generic methodology was applied... to identify 
potential events which would dilute the soluble boron contained in 
PWR [pressurized water reactor] fuel storage pools, and to quantify 
the frequency of those events. This methodology utilized a 
probabilistic assessment of a composite plant model to calculate the 
event frequency of a dilution event. The results of the assessment 
concluded that the event frequency remained less than the NRC Safety 
Goal Policy Statement target risk objective of IE-6/reactor year.
    Differences between the composite plant described in WCAP-14181 
and Vogtle relative to the potential sources of pool dilution were 
addressed in an individual analysis of the Vogtle pool. This 
analysis was conducted with methodology which closely paralleled 
that employed in WCAP-14181. That analysis, found in Enclosure 6 [of 
the licensee's proposed amendment request], concluded that the 
frequency of pool dilution to the 0.95 Keff boron concentration 
(1250 ppm) is on the same order of magnitude as reported in WCAP-
14181 and less than the NRC Safety Goal Policy Statement criterion 
of 1.0E-6/reactor year.
    Proposed Technical Specifications 3.7.17 and 3.7.18 which ensure 
the maintenance of the fuel storage pool boron concentration and 
storage configuration, do not represent new concepts. The actual 
boron concentration in the fuel storage pool has been maintained at 
a higher value than the proposed limits for the Unit 1 and 2 fuel 
storage pools for refueling purposes. The criticality analysis 
(Enclosure 5 [of the licensee's proposed amendment request]) 
determined that a boron concentration of 1,100 ppm (Unit 1) and, 
1,250 ppm (Unit 2) results in a Keff<0.95 including all the 
calculational uncertainties and additional margin to compensate for 
the possibility of loss of cooling, or a misplaced assembly.
    There is no significant change in plant configuration, equipment 
design, or usage of plant equipment. The safety analysis for 
dilution accidents has been expanded; however, the criticality 
analyses assure that the pool will remain subcritical with no credit 
for soluble boron. Therefore, the proposed changes will not create 
the possibility of a new or different kind of accident.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    Proposed Technical Specifications 3.7.17 and 3.7.18 and the 
associated spent fuel boron concentration and storage limits in the 
COLR will provide adequate safety margin to assure that the stored 
fuel assembly array will always remain subcritical. Those limits are 
based on a plant specific criticality analysis (Enclosure 5 [of the 
licensee's proposed amendment request]) performed in accordance with 
the Westinghouse criticality analysis methodology...
    While the cricality analysis utilized credit for soluble boron, 
a storage configuration has been defined using maximum feasible 
Keff calculations to ensure that the spent fuel rack Keff 
will be less than 1.0 with no soluble boron under normal storage 
conditions and assuming nominal fuel assembly parameters and fuel 
rack dimensions. Soluble boron credit is used to offset 
uncertainties, tolerances and off-normal conditions (such as a 
misplaced assembly) and to provide subcritical margin such that the 
fuel storage pool Keff is maintained less than or equal to 
0.95.
    The loss of a considerable amount of soluble boron in the fuel 
storage pool which could lead to exceeding a Keff of 0.95 
during accidents and under adverse conditions has been evaluated and 
shown to be very improbable.
    The combination of the probabilistic evaluation which shows that 
the dilution of the fuel storage pool is a low probability 
occurrence, the maximum feasible Keff calculation which shows 
that the Keff will remain less than 1.0 when flooded with 
unborated water and assuming nominal fuel assembly parameters and 
fuel rack dimensions, and the unavailability of the large volumes of 
water which are necessary to dilute the fuel storage pool, provide a 
level of safety comparable to the conservative criticality analysis 
methodology...
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: September 25, 1996
    Description of amendment request: The proposed amendment would (1) 
revise the required number of operable gaseous radioactivity monitoring 
system channels and particulate radioactivity monitoring system 
channels from one in each of the monitoring systems to one in either of 
the monitoring systems, (2) allow both the gaseous radioactivity 
monitoring system and the particulate monitoring system to be 
inoperable for up to 30 days provided that grab samples are obtained 
and analyzed at least once per 12 hours, and (3) add an action for the 
loss of all reactor coolant system leakage detection systems (drywell 
floor sump level monitoring system, gaseous radioactivity monitoring 
system and particulate radioactivity monitoring 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:

[[Page 58905]]

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The function of the reactor coolant system leakage detection 
systems is to detect leakage from the reactor coolant pressure 
boundary so that appropriate actions can be taken before the 
integrity of the reactor coolant pressure boundary is impaired. In 
the plant accident analysis, no credit for mitigation of an accident 
is taken for the reactor coolant system leakage detection systems. 
These proposed changes do not alter this function, therefore, these 
changes do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    The function of the reactor coolant system leakage detection 
systems is to detect leakage from the reactor coolant pressure 
boundary so that appropriate actions can be taken before the 
integrity of the reactor coolant pressure boundary is impaired. 
These proposed changes do not alter this function; therefore, these 
changes do not create the possibility of a new or different kind of 
accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The change to allow both the gaseous and particulate 
radioactivity monitoring systems to be inoperable at the same time 
provided a grab sample is obtained and analyzed at least once per 12 
hours is predicated on the availability of the primary leak 
detection system (drywell floor sump level monitor system). Since 
the gaseous and particulate radioactivity monitoring systems are 
backups to the drywell floor sump level monitoring system, allowing 
grab samples every 12 hours provides periodic information that is 
adequate to detect leakage. The addition of the action to require an 
orderly shutdown of the unit for the loss of all reactor coolant 
system leakage detection systems does not affect the margin of 
safety. Therefore, these proposed changes do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: October 24, 1996
    Description of amendment request: The proposed amendments would 
change Technical Specification 3/4.7.1.2, ``Auxiliary Feedwater 
System.'' The changes would revise the 18-month surveillances performed 
on the system's pumps and valves because testing of the turbine driven 
Auxiliary Feedwater pump (TDAFWP) can only be performed in higher modes 
when there is sufficient secondary steam pressure.
    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 changes proposed on the testing of components in the AFW 
[Auxiliary Feedwater] System do not affect the operation of the 
equipment during conditions when they are required to perform their 
safety function. No physical changes to the plant result from the 
proposed changes made to the surveillance requirements. The AFW 
System is used as a backup system upon loss of main feedwater which 
is analyzed as a Condition II event in the UFSAR [Updated Final 
Safety Analysis Report] and as such, does not impact the probability 
of an accident.
    Testing is being performed with the plant in the condition in 
which the automatic initiation signals would result, that is, with 
the plant in Hot Standby. The changes do not impact the availability 
of the AFW System in providing feedwater to the steam generators. 
The 24 hour duration to perform testing is sufficiently short that 
it is considered unlikely that a condition requiring AFW initiation 
would occur with the TDAFWP unable to feed the generators. For such 
an occurrence, however, the motor driven AFW pumps would be 
available to mitigate the consequences of the event. This time is 
less than the 72 hour allowed outage time for an inoperable TDAFWP 
in Modes 1-3.
    Therefore, the consequences of an accident previously evaluated 
are not significantly increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve any modifications to 
existing plant equipment, do not alter the function of any plant 
systems, do not introduce any new operating configurations or new 
modes of plant operation, nor change the safety analyses. Testing of 
the TDAFWP in Mode 3, Hot Standby, will not impact auxiliary 
feedwater capability or impact the ability to maintain Reactor 
Coolant temperature. The proposed changes will, therefore, 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 changes to the valve surveillance does not decrease the 
scope of the existing testing, but will clarify the automatic valves 
to be included.
    The time in which testing is performed, within 24 hours of 
reaching 680 psig steam generator pressure, ensures that testing is 
performed in a timely manner after attaining the required steam 
pressure. This does not impose a significant safety impact since the 
testing is performed within the plant at the zero load conditions 
prior to increasing reactor power.
    Elimination of the wording ``during shutdown,'' in reference to 
the time in which the surveillance is performed, is considered 
editorial and is proposed for consistency with the change made to 
the pump surveillance requirement.
    All changes are consistent with the intent of Salem's current TS 
and with the 18 month surveillances specified in NUREG-1431, 
Revision 1.
    The proposed change, therefore, does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: September 30, 1996 (TSCR 192)
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 15.3.3, ``Emergency Core 
Cooling System, Auxiliary Cooling Systems, Air Recirculation Fan 
Coolers, and Containment Spray,'' TS 15.3.7, ``Auxiliary Electrical 
Systems,'' and the TS Bases to reflect proposed changes to the limiting 
conditions for operation, action statements, allowable outage times, 
and design specifications for the Point Beach Nuclear Plant (PBNP) TS 
associated with the containment

[[Page 58906]]

accident fan coolers, service water equipment, and normal and emergency 
power supplies.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Operation of this facility under the proposed Technical 
Specifications will not create a significant increase in the 
probability or consequences of an accident previously evaluated.
    The probabilities of accidents previously evaluated are based on 
the probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated for Point Beach include: 
Control rod withdrawal and drop, CVCS [chemical volume and control 
system] malfunction (Boron Dilution), startup of an inactive reactor 
coolant loop, reduction in feedwater enthalpy, excessive load 
increase, losses of reactor coolant flow, loss of external 
electrical load, loss of normal feedwater, loss of all AC power to 
the auxiliaries, turbine overspeed, fuel handling accidents, 
accidental releases of waste liquid or gas, steam generator tube 
rupture, steam pipe rupture, control rod ejection, and primary 
coolant system ruptures.
    This license amendment request proposes to change the limiting 
conditions for operation, action statements, allowable outage times, 
and design specifications for the Point Beach Nuclear Plant 
Technical Specifications associated with the containment accident 
fan coolers, service water equipment, and normal and emergency power 
supplies.
    These proposed changes do not cause an increase in the 
probabilities of any accidents previously evaluated because these 
changes will not cause an increase in the probability of any 
initiating events for accidents previously evaluated. In particular, 
these changes affect accident mitigation systems and equipment which 
do not cause accidents.
    The consequences of the accidents previously evaluated in the 
PBNP FSAR [final safety analysis report] are determined by the 
results of analyses that are based on initial conditions of the 
plant, the type of accident, transient response of the plant, and 
the operation and failure of equipment and systems. The changes 
proposed in this license amendment request provide appropriate 
limiting conditions for operation, action statements, and allowable 
outage times for service water, containment cooling and normal and 
emergency power supplies.
    The proposed changes affect components that are required to 
ensure the proper operation of engineered safety features equipment. 
The proposed changes do not increase the probability of failure of 
this equipment or its ability to operate as required for the 
accidents previously evaluated in the PBNP FSAR. The proposed 
changes that increase the allowed outage times for engineered safety 
features equipment continue to provide appropriate limitations for 
these conditions because sufficient equipment is still required to 
be operable for accident mitigation and the proposed allowed outage 
times are consistent with currently accepted time periods for these 
situations.
    Therefore, this proposed license amendment does not affect the 
consequences of any accident previously evaluated in the Point Beach 
Nuclear Plant FSAR, because the factors that are used to determine 
the consequences of accidents are not being changed.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    New or different kinds of accidents can only be created by new 
or different accident initiators or sequences. New and different 
types of accidents (different from those that were originally 
analyzed for Point Beach) have been evaluated and incorporated into 
the licensing basis for Point Beach Nuclear Plant. Examples of 
different accidents that have been incorporated into the Point Beach 
licensing basis include anticipated transients without scram and 
station blackout.
    The changes proposed by this license amendment request do not 
create any new or different accident initiators or sequences because 
these changes to limiting conditions for operation, action 
statements, allowable outage times, and design specifications for 
service water, containment cooling and normal and emergency power 
supplies will not cause failures of equipment or accident sequences 
different than the accidents previously evaluated. Therefore, these 
proposed Technical Specification changes do not create the 
possibility of an accident of a different type than any previously 
evaluated in the Point Beach FSAR.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design 
and operation of the reactor and containment and the safety systems 
that provide their protection.
    The changes proposed by this license amendment request provide 
the appropriate limiting conditions for operation, action 
statements, allowable outage times, and design specifications for 
service water, containment cooling and normal and emergency power 
supplies. This ensure that the safety systems that protect the 
reactor and containment will operate as required. The design and 
operation of the reactor and containment are not affected by these 
proposed changes. Therefore, the margins of safety for Point Beach 
are not being reduced because the design and operation of the 
reactor and containment are not being changed and the safety systems 
and limiting conditions of operation for these safety systems that 
provide their protection that are being changed will continue to 
meet the requirements for accident mitigation for Point Beach 
Nuclear Plant.
    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: John N. Hannon

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.

[[Page 58907]]

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

    Date of amendment request: April 2, 1996 (BSEP 96-0123), as 
supplemented by an earlier submittal dated November 20, 1995 (BSEP 95-
0535), and by subsequent submittals dated July 1, 1996 (BSEP 96-0242), 
July 30, 1996 (BSEP 96-0287), August 7, 1996 (BSEP 96-0300), September 
13, 1996 (BSEP 96-0340), September 20, 1996 (BSEP 96-0348), October 1, 
1996 (BSEP 96-0362), October 22, 1996 (BSEP 96-0392), October 22, 1996 
(BSEP 96-0403), and October 29, 1996 (BSEP 96-0412).
    Brief description of amendment: The proposed amendment would modify 
Facility Operating Licenses Nos. DPR-71 and DPR-62 and the Technical 
Specifications (TS) for the Brunswick Steam Electric Plant, Units 1 and 
2, respectively, to authorize an increase in the maximum power level 
from 2436 megawatts thermal (MWt) to 2558 MWt.
    Date of issuance: November 1, 1996
    Effective date: November 1, 1996
    Amendment No.: 183 (Unit 1); 214 (Unit 2)
    Facility Operating License Nos. DPR-71 and DPR-62: Amendment 
revises
    Facility Operating License Nos. DPR-71 and DPR-62 and the Technical 
Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25698) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 1, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: August 5, 1994, as supplemented 
by letters dated November 17, 1994, December 2, 1994, and August 1, 
1996.
    Brief description of amendment: The amendment revises surveillance 
intervals for various systems, components and instruments to 
accommodate a 24-month refueling cycle. These revisions are being made 
in accordance with the guidance provided by Generic Letter 91-04, 
``Changes in Technical Specification Surveillance Intervals to 
Accommodate a 24-Month Fuel Cycle.''
    Date of issuance: October 30, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 187
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63117) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 30, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: December 11, 1995, as 
supplemented by letters dated January 15, September 3, October 2, 
October 18, and October 25, 1996.
    Brief description of amendment: The amendment revises the 
Administrative Controls section of the TS by deleting or relocating 
requirements that are adequately controlled by existing regulatory 
requirements, adding requirements, and editorially restructuring the TS 
to be consistent with NUREG-1432, ``Standard Technical Specifications, 
Combustion Engineering Plants.'' In addition, containment leak rate 
testing requirements are revised to allow the Type A integrated leak 
rate test to be scheduled in accordance with Option B of 10 CFR Part 
50, Appendix J. Review of several changes proposed by the licensee have 
not yet been completed by the staff. The NRC will issue an evaluation 
of these changes upon completion of staff review.
    Date of issuance: October 31, 1996
    Effective date: October 31, 1996
    Amendment No.: 174
    Facility Operating License No. DPR-20 Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 20, 1996 (61 
FR 49493). The October 2, October 18, and October 25, 1996, letters 
provided clarifying information and updated TS pages that were within 
the scope of the initial application and did not affect the staff's 
initial proposed no significant hazards consideration determination. 
Therefore, renoticing was not warranted.The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
October 31,1996. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, IllinoisDate of 
application for amendments: April 8, 1996, as supplemented on 
October 14, 1996.

    Brief description of amendments: The amendments revise various 
sections of the Technical Specifications (TS) to reflect the transition 
of fuel supplier from General Electric (GE) to Siemens Power 
Corporation (SPC). The amendments revise the definitions, limiting 
conditions for operation, required actions, or surveillance 
requirements related to the following fuel thermal limits: Linear Heat 
Generation Rate, Critical Power Ratio, Minimum Critical Power Ratio, 
and Average Planar Linear Heat Generation Rate. The previous GE 
terminology is replaced with vendor independent terms and new, NRC-
approved methodologies are incorporated. The amendments also include 
changes to Section 6.0 of the TS to include SPC references, relocate 
the requirements for the traversing in-core probe system from the TS to 
the Core Operating Limits Report, and revise the fuel description in TS 
Section 5.0.
    Date of issuance: October 29, 1996
    Effective date: Immediately, to be implemented prior to startup of 
Cycle 9 for Unit 1 and Cycle 8 for Unit 2.
    Amendment Nos.: 116, 101
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 22, 1996 (61 FR 
25699) The October 14, 1996, submittal provided additional clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
October 29, 1996.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station, Units 1 and 2, Lake County, IllinoisDate of 
application for amendments: August 16, 1996, as supplemented on 
October 4, 1996.

    Brief description of amendments: The amendments revise the 
definition of the F* distance by removing the uncertainty

[[Page 58908]]

term from the specified distance and removing the footnote which 
specifies the time frame for which it is applicable.
    Date of issuance: November 6, 1996
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 174, 161
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47968) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 6, 1996No significant 
hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments: September 3, 1996
    Brief description of amendments: The amendments incorporate revised 
installation procedures for steam generator tube sleeves designed by 
ABB Combustion Engineering (ABB/CE).
    Date of issuance: October 29, 1996
    Effective date: October 29, 1996
    Amendment Nos.: 173 and 160
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 11, 1996 (61 
FR 47966) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated October 29, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan Date of application for amendment: September 5, 1996 (NRC-
96-0075), as supplemented by letters dated October 14, October 23, 
October 29, and October 31, 1996

    Brief description of amendment: The amendment revises Technical 
Specification (TS) 2.1.2 to incorporate cycle-specific safety limit 
minimum critical power ratios (SLMCPRs) for the core that will be 
loaded for Cycle 6. In addition, TS 3.4.1.1 is revised to delete the 
specific SLMCPR number and replace it with a reference to TS 2.1.2.
    Date of issuance: November 5, 1996
    Effective date: November 5, 1996, with full implementation within 
45 days
    Amendment No.: 109
    Facility Operating License No. NPF-43 Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 25, 1996 (61 
FR 50342) The letters of October 14, 23, 29, and 31, 1996, provided 
clarifying information and were not outside the scope of the initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 5, 1996.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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

    Date of application for amendments: December 14, 1995, as 
supplemented by letters dated May 16 and August 29, 1996
    Brief description of amendments: The amendments modify the 
Technical Specifications for diesel generators to incorporate guidance 
and recommendations contained in NRC Generic Letter (GL) 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation,'' GL 94-01, ``Removal 
of Accelerated Testing and Special Reporting Requirements for Emergency 
Diesel Generators,'' NUREG-1431, ``Revised Standard Technical 
Specifications for Westinghouse PWRs,'' and NUREG-1366, ``Improvements 
to Technical Specifications Surveillance Requirements.''
    Date of issuance: October 30, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 155 and 147
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 19, 1996 (61 FR 
31175) The August 29, 1996, submittal provided additional information 
that did not change the scope of the December 14, 1995, application and 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 1996. No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

Duke Power Company, Docket Nos. 50-269, 50-270 and 50-287, Oconee 
Nuclear Station, Units 1, 2 and 3, Oconee County, South 
CarolinaDate of application for amendments: August 12, 1996, as 
supplemented by letter dated September 10, 1996

    Brief description of amendments: The amendments revise the 
Technical Specifications associated with the containment leak-rate 
tests by implementing 10 CFR Part 50, Appendix J, Option B, for Type A 
leak-rate testing.
    Date of issuance: October 30, 1996
    Effective date: As of the date of issuance to be implemented 30 
days from the date of issuance.
    Amendment Nos.: 218, 218, 215
    Facility Operating License Nos. DPR-38, DPR-47 and DPR-55: 
Amendments revise the Technical Specifications.
    Date of initial notice in Federal Register: August 28, 1996 (61 FR 
44356) The September 10, 1996, letter provided additional information 
that did not change the scope of the August 12, 1996, application and 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated October 30, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina
    GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey
    Date of application for amendment: April 15, 1996 (TSCR No. 244)
    Brief description of amendment: The amendment revises Specification 
5.3.1.B to allow the shield plug and the associated lifting hardware to 
be moved over irradiated fuel assemblies that are in a dry shielded 
canister within the transfer cask in the cask drop protection system.
    Date of Issuance: November 7, 1996, to be implemented within 30 
days of issuance
    Effective date: November 7, 1996
    Amendment No.: 187
    Facility Operating License No. DPR-16. Amendment revises the 
Technical Specifications

[[Page 58909]]

    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20849) The Commission's related evaluation of this amendment and final 
determination of no significant hazards consideration addressing 
comments received on the proposed no significant hazards consideration 
determination are contained in a Safety Evaluation dated November 7, 
1996.No significant hazards consideration comments received: Yes.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753

GPU Nuclear, Inc., Docket No. 50-320, Three Mile Island Nuclear 
Station, Unit No. 2, (TMI-2), Dauphin County, Pennsylvania

    Date of application for amendment: February 6, 1995
    Brief description of amendment: This amendment revised the 
Technical Specifications by extending the surveillance interval to 
demonstrate operability of the containment airlocks from quarterly to 
annually and to decrease the personnel exposure with implementing the 
surveillance.
    Date of issuance: October 24, 1996
    Effective date: October 24, 1996
    Amendment No.: 51Possession-Only License No. DPR-73: The amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28616) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 24, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105

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

    Date of amendment request: May 1, 1995, as supplemented by letters 
dated June 22, August 28, November 22, and December 19, 1995, and 
January 4, 8 (two letters), and 23, June 27, July 9, August 8, and 
September 23, 1996.
    Brief description of amendments: The amendments allowed extension 
of the standby diesel generator allowed outage time to 14 days, and 
extension of the essential cooling water loop and the essential chilled 
water loop allowed outage times to 7 days. The amendments also added to 
Administrative Controls a description of the Configuration Risk 
Management Program (CRMP) used to assess changes in core damage 
probability resulting from applicable plant configurations.
    Date of issuance: October 31, 1996
    Effective date: October 31, 1996, to be implemented within 30 days
    Amendment Nos.: 85 and 72
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 31, 1996 (61 FR 
40019) The additional information contained in the supplemental letters 
dated August 8 and September 23, 1996, were clarifying in nature and 
thus, within the scope of the initial notice and did not affect the 
staff's proposed no significant hazards consideration determination.The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated October 31, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: February 22, 1996, and 
supplemented July 22, 1996
    Brief description of amendments: The amendments revise the 
administrative controls section of the technical specifications to 
change the operator license requirements for operations management.
    Date of issuance: October 29, 1996
    Effective date: October 29, 1996, with full implementation within 
45 days
    Amendment Nos.: 212 and 197
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13527) The July 22, 1996, submittal was more restrictive than the 
original submittal and did not change the staff's original no 
significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated October 29, 1996.No significant hazards consideration 
comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

Northeast Nuclear Energy Company, et al., Docket No. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London County, 
ConnecticutDate of application for amendment: August 27, 1996

    Brief description of amendment: The Technical Specification (TS) 
amendment clarifies the limiting condition for operation and 
surveillance requirements to ensure that the appropriate number of 
charging pumps and high pressure safety injection pumps are operable 
for reactivity control and reactor coolant system (RCS) makeup 
requirements, while also limiting the number of operable pumps to 
ensure that the low temperature overpressure limits will not be 
exceeded in the event of a mass addition to the RCS during shutdown 
conditions. The TS Bases remain unchanged as the result of this 
amendment.
    Date of issuance: October 25, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 205
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 20, 1996 (61 
FR 49498) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 25, 1996No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New YorkDate of 
application for amendment: March 22, 1996, as supplemented October 
11, 1996

    Brief description of amendment: The amendment proposed changes to 
the Technical Specifications to establish operability requirements for 
avoidance and protection from thermal hydraulic instabilities to be 
consistent with Boiling Water Reactor Owners Group long-term solution 
Option I-D. Editorial changes are also made to support the revised 
specifications, improve readability of Bases sections, and enhance the 
presentation of requirements for single loop operation.
    Date of issuance: October 30, 1996

[[Page 58910]]

    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 236
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 8, 1996 (61 FR 
20854) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 30, 1996No 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.

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: May 30, 1996, as supplemented by 
letter dated October 11, 1996
    Brief description of amendment: The amendment proposes to eliminate 
selected response time testing requirements for certain sensors and 
specified loop instrumentation.
    Date of issuance: October 28, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 235
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 3, 1996 (61 FR 
34896) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated October 28, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Public Service Electric & Gas Company, Docket Nos. 50-272 and 50-
311, Salem Nuclear Generating Station, Unit Nos. 1 and 2, Salem 
County, New Jersey Date of application for amendments: July 12, 
1996, as supplemented September 12, 1996

    Brief description of amendments: The amendments revise Technical 
Specification Table 3.3-3, ``Engineered Safety Feature Actuation System 
Instrumentation,'' to clarify the setpoint for the interlock designated 
P-12.
    Date of issuance: November 4, 1996
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendment Nos. 185 and 167
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 23, 1996 (61 FR 
38229) The supplemental letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination nor the Federal Register notice.The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
November 4, 1996.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079
    Dated at Rockville, Maryland, this 13th day of November 1996.
    FOR THE NUCLEAR REGULATORY COMMISSION
Steven A. Varga,
Director, Division of Reactor Projects - I/II,Office of Nuclear Reactor 
Regulation
[FR Doc. 96-29584 Filed 11-18-96; 8:45 am]
BILLING CODE 7590-01-F