[Federal Register Volume 62, Number 185 (Wednesday, September 24, 1997)]
[Notices]
[Pages 50000-50018]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X97-30924]


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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 August 29, 1997, through September 12, 1997. 
The last biweekly notice was published on September 10, 1997 (62 FR 
47696).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be

[[Page 50001]]

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 October 24, 1997, 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.

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 amendment request: July 15, 1997
    Description of amendment request: The proposed amendments would: 
(1) add Technical Specification (TS) 3.5.7, ``Main Steam Line Break 
Detection and Feedwater Isolation,'' to identify operability 
requirements and Bases for the main steamline break (MSLB) detection 
isolation circuitry, the feedwater isolation circuitry, the main 
feedwater main control valves, and the main feedwater startup control 
valves; (2) revise TS 3.5.1, ``Operation Safety Instrumentation'' to 
add a reference to TS 3.5.7; (3) revise Table 3.5.1-1, ``Instruments 
Operating Conditions,'' to reflect operability requirements for the 
main steam header pressure and MSLB detection channels, the feedwater 
isolation channels, and the feedwater isolation channels manual 
pushbuttons; and (4) revise Table 4.1-1, ``Instrument Surveillance 
Requirements,'' and Table 4.1-2, ``Minimum Equipment Test Frequency,'' 
to include surveillance requirements for the subject circuitry and 
components.
    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 50002]]

    A. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    NO
    This proposed Technical Specification amendment does not create 
any conditions or events which lead to accidents (events) previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report], other 
than a loss of Main Feedwater (FDW). The new MSLB detection and 
feedwater isolation circuitry addressed by this change is designed 
so that a credible single failure will not cause a loss of FDW to 
the steam generator unless [an] MSLB is detected. Single failures 
are not assumed if entry into a Technical Specification action 
statement occurs.
    During [an] MSLB, the circuitry is intentionally stopping and 
isolating FDW. Operators are currently instructed to isolate FDW on 
indication of [an] MSLB. The new circuitry will automatically stop 
FDW to eliminate the need for this operator action. Thus the 
probability of the stopping (loss) of FDW is not increased. The NRC 
has also stated that the stopping of FDW to mitigate [an] MSLB is an 
acceptable response to address the concerns of Inspection and 
Enforcement Bulletin 80-04.
    The Emergency Feedwater (EFW) System is an accident mitigation 
system. The MSLB modification and associated Technical Specification 
to keep the turbine driven emergency feedwater pump (TDEFW) pump 
from starting following [an] MSLB will not initiate any accidents.
    The potential for containment overpressurization currently 
exists without the installed modification and associated Technical 
Specification. The new MSLB detection and feedwater isolation 
circuitry will assist in reducing the potential for the 
overpressurization of containment. The EFW circuitry is designed so 
that the TDEFWP will still auto start for any event other than [an] 
MSLB. The TDEFWP can still be manually started during [an] MSLB or 
FDW line break accident as needed. This action is similar to other 
manual actions to align EFW for the MSLB scenarios that are already 
described in the ONS [Oconee Nuclear Station] UFSAR. This new 
circuitry and associated Technical Specification creates no new 
credible single failures that could prevent the TDEFWP from auto 
starting (except for the MSLB). The motor driven EFW pumps and EFW 
flow control valves are not adversely affected by this change and 
will provide EFW flow for scenarios other than Station Blackout. 
Both FDW and EFW will still provide their design functions of 
supplying feedwater to the steam generators, as evaluated in the 
UFSAR. The ability to shut down following a 10CFR50 Appendix R fire 
is not adversely affected. This Technical Specification change does 
not adversely affect containment integrity and radiological release 
pathways.
    B. Create the possibility of a new or different kind of accident 
from the accident previously evaluated?
    NO
    No accidents different than already evaluated in the UFSAR are 
postulated. The FDW System will still perform its design function of 
supplying feedwater to the steam generators as evaluated in the 
UFSAR. The EFW System will still provide its function of supplying 
feedwater to the steam generators, as evaluated in the UFSAR for 
events resulting in the loss of the FDW System.
    C. Involve a significant reduction in a margin of safety?
    NO
    The design pressure of containment is specified to be 59 psig in 
the bases to several Technical Specifications. With the potential 
for unrestricted FDW and EFW flow during [an] MSLB inside 
containment, the design pressure of the containment could be 
exceeded. The proposed Technical Specifications address equipment 
which will function to isolate FDW in the unlikely event of [an] 
MSLB accident. Therefore, the proposed Technical Specifications do 
not increase the potential for the containment to be pressurized or 
increase the expected pressure of containment following [an] MSLB. 
No plant safety limits, set points, or design parameters are 
adversely affected. The fuel, fuel cladding, and Reactor Coolant 
System are not impacted.
    Duke [Duke Energy Corporation] has concluded based on the above 
that there are no significant hazards considerations involved in 
this 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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: August 28, 1997 (TSC 96-09)
    Description of amendment request: The proposed changes would add 
new limiting conditions for operation and new surveillance requirements 
for the Emergency Condenser Circulating Water System, the Essential 
Siphon Vacuum System, and the Siphon Seal Water System to reflect 
design changes and modifications to these systems.
    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 the change] involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    NO.
    This Technical Specification change does not create any 
conditions or events which lead to accidents previously evaluated in 
the UFSAR [Updated Final Safety Analysis Report]. The new ECCW 
[Emergency Condenser Circulating Water] System Technical 
Specification 3.19, along with the new ECCW Surveillance 
requirements specified in Technical Specification Table 4.1-2, are 
conservative in nature. No existing Technical Specification 
requirements are being deleted with this revision. Surveillance and 
operability requirements are being added for the upgraded ECCW 
System.
    The ECCW System is only required following the occurrence of 
loss of offsite power (LOOP) events. The most limiting of these LOOP 
events is the loss of coolant accident concurrent with the LOOP 
(LOCA/LOOP). Therefore, the ECCW System is not considered to be an 
accident initiator. As a result, the proposed new ECCW Technical 
Specification requirements will not result in any increase in the 
probability of any design basis accidents or events evaluated in the 
UFSAR.
    The credit for restarting a CCW [Condenser Circulating Water] 
pump within 1.5 hours following a LOOP, to ensure suction to LPSW 
[Low Pressure Service Water] is maintained, is being replaced by 
credit for maintaining the ECCW siphon using the new siphon support 
systems (ESV [Essential Siphon Vacuum] System and SSW [Siphon Seal 
Water] System) in conjunction with the upgraded ECCW System. 
Therefore, obsolete requirements specified in Selected Licensee 
Commitments (SLCs) 16.9.7 and 16.9.8 will be revised or deleted 
accordingly. Replacement of the CCW pump restart during a LOOP with 
the ability to maintain ECCW siphon flow will not create any 
conditions or events which lead to accidents previously evaluated in 
the UFSAR.
    The modifications to upgrade the ECCW System were performed to 
improve the reliability of the ECCW System. The proposed new ECCW 
Technical Specification provides additional surveillance and 
operability requirements to ensure that the upgraded ECCW System 
will function reliably during the design basis events which require 
its operation. Therefore, these proposed new Technical Specification 
requirements will not increase the consequences of any accidents 
previously evaluated in the UFSAR.
    [2. Will the change] create the possibility of a new or 
different kind of accident from the accident previously evaluated?
    NO.
    No accidents different than those already evaluated in the UFSAR 
are postulated. The upgraded ECCW System will more reliably perform 
its design function of supplying water to the suction of the Low 
Pressure Service Water (LPSW) System as evaluated in the UFSAR. The 
new Technical Specification requirements will increase the 
reliability of the upgraded ECCW System. In addition, the ECCW 
System is not an accident initiator since it is used following 
certain design basis events such as a LOCA/LOOP.

[[Page 50003]]

    [3. Will the change] involve a significant reduction in a margin 
of safety?
    NO.
    The proposed Technical Specifications address equipment which 
will function in certain design basis events, such as a LOCA/LOOP, 
to ensure a reliable water supply to the LPSW System. The LPSW 
System must function to remove decay heat from primary systems and 
the reactor building during a LOCA/LOOP. The proposed Technical 
Specifications addressing the upgraded ECCW System will further 
enhance the reliability of the ECCW System and will result in 
greater assurance that the LPSW System can perform its safety 
functions. No plant safety limits, setpoints, or design parameters 
are adversely affected. The fuel, fuel cladding, and Reactor Coolant 
System are not impacted. The proposed Technical Specifications 
provide additional, conservative, operational requirements beyond 
the current Technical Specifications which address the ECCW System.
    Duke [Duke Energy Corporation] has concluded based on this 
information that there are no significant hazards considerations 
involved in this amendment request.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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 amendment request: September 4, 1997
    Description of amendment request: The proposed changes would 
incorporate changes to the Oconee Final Safety Analysis Report and 
Technical Specification Bases to address a potential unreviewed safety 
question associated with implementation of revised small break loss-of-
coolant accident analysis. The proposed changes would address operation 
of the facility and single failure criteria related to the high 
pressure injection system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. None of the proposed changes [have] any impact upon the 
probability of any accident which has been evaluated in the UFSAR 
[Updated Final Safety Analysis Report].
    None of these changes have any impact upon the ability of the 
HPI [high-pressure injection] System to mitigate the consequences of 
a small break LOCA [loss-of-coolant accident], which is addressed 
below. The small break LOCA is the limiting design basis accident 
with respect to the HPI System operability requirements.
    The proposed changes to the Bases of Specification 3.3.1 and 
Chapter 15 of the Oconee UFSAR include operator actions that have 
not previously been reviewed and approved by the [NRC] staff for 
licensing basis small break LOCA analyses. However, these operator 
actions have been included in the Emergency Operating Procedure for 
over 10 years and crediting these actions in the safety analyses 
does not result in any change to the operator's response to a small 
break LOCA. These actions are simply changes to the assumptions 
contained in the licensing basis small break LOCA analyses. The 
operability requirements for the HPI System contained in 
Specification 3.3.1 are supported by a spectrum of small break LOCA 
analyses based on the approved Evaluation Model described in FTI 
[Framatome Technologies, Inc.] topical report BAW-10192P. These 
small break LOCA analyses demonstrate that the acceptance criteria 
of 10CFR 50.46 are satisfied.
    The operability requirements in Technical Specification 3.3.1.c 
assure that the HPI System can withstand the worst single failure 
and still result in two HPI pumps injecting through two trains. The 
full power small break LOCA analyses supporting this proposed 
license amendment have been performed in accordance with the 
approved Evaluation Model described in FTI topical report BAW-
10192P.
    When at or below 75% FP [full power], one HPI train provides 
sufficient flow to mitigate a small break LOCA. The 60% power level 
currently in Specification 3.3.1 is justified by analyses using the 
Evaluation Model described in FTI topical report BAW-10192P, 
considering the worst case break location and size described in LER 
[Licensee Event Report] 269/90-15 and Attachment 2 to this 
submittal. The proposed changes to the Bases of Technical 
Specification 3.3.1 describe the operator actions credited to 
justify the adequacy of the current specification and eliminate the 
need for the administrative restrictions imposed by LER 269/90-15. 
These requirements ensure that, following the worst single failure, 
one train of HPI would remain available to mitigate a small break 
LOCA.
    In summary, the technical analyses described in this license 
amendment justify the adequacy of this specification and assure that 
operability of the HPI System is maintained in a manner consistent 
with the requirements of the design basis accidents. Therefore, it 
is concluded that this amendment request will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    No. The proposed changes to the Bases of Technical Specification 
3.3.1 and Chapter 15 of the Oconee UFSAR do not result in any new 
operator actions or changes in plant operation. The proposed changes 
involve crediting operator actions in the licensing basis small 
break LOCA analyses that have been included in the Emergency 
Operating Procedure for years. No new initiating events or 
potentially unanalyzed conditions have been created. Therefore, this 
proposed amendment will not create the possibility of any new or 
different kind of accident.
    (3) Involve a significant reduction in a margin of safety.
    No. The HPI System requirements associated with the proposed 
UFSAR and Technical Specification Bases changes are supported by 
analyses which demonstrate that the acceptance criteria of 10 CFR 
50.46 are not violated for any small break LOCA. These analyses were 
performed in accordance with the Evaluation Model described in FTI 
topical report BAW-10192P. Therefore, it is concluded that the 
proposed amendment request will not result in a significant decrease 
in the margin of safety.
    Duke [Duke Energy Corporation] has concluded, based on the 
above, that there are no significant hazards considerations involved 
in this amendment request.
    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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: August 6, 1997
    Description of amendment request: The proposed amendment would 
eliminate the provisions in Technical Specification 3.8.1, ``AC Sources 
- Operating,'' for accelerated testing of the emergency diesel 
generators (DG). The proposed changes are the following: (1) the 
frequency of verifying DG starts and operation in Surveillance 
Requirements 3.8.1.2 and 3.8.1.3, respectively, would be changed to 31 
days, from the present reference to Table 3.8.1-1, and (2) Table 3.8.1-
1, ``Diesel Generator Test

[[Page 50004]]

Schedule,'' would be deleted. The emergency DG provide emergency AC 
power to the site with the loss of offsite AC power.
    Basis for proposed no significant hazardsconsideration 
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. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    [These] change[s] will provide flexibility to structure the 
standby diesel generator maintenance program based on the risk 
significance of the structures, systems, and components [(SSCs)] 
that are within the scope of the Maintenance Rule [(10 CFR 50.65, 
``Requirements for Monitoring the Effectiveness of Maintenance at 
Nuclear Power Plants)]. The removal of the diesel generator 
accelerated testing is acceptable as the maintenance rule applies 
site and system specific performance criteria to monitor diesel 
generator performance. This criteria includes a running availability 
and reliability goal as well as specific goals to monitor 
maintenance preventable functional failures. The performance 
criteria for the diesel generator reliability and availability 
established by the maintenance rule and the causal determinations 
and corrective actions required for maintenance preventable 
functional failures are considered to be an acceptable method for 
monitoring diesel generator performance.
    The proposed change[s] [have] no effect on the probability of 
the initiation of an accident, because the emergency diesel 
generators do not serve as the initiator of any event. Additionally, 
as diesel generator performance will continue to be [ensured] by the 
maintenance rule, the proposed changes do not affect the ability to 
mitigate the consequences of an accident previously evaluated. The 
changes do not impact the diesel [generator]'s design sources, 
operating characteristics, system functions, or system 
interrelationships. The failure mechanisms for the accident 
previously evaluated are not affected and no additional failure 
modes are created that could cause an accident that has been 
previously evaluated. Since the diesel generator performance and 
reliability will continue to be [ensured] by the maintenance rule, 
the proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. This request does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    [These] proposed change[s] [do] not involve a change to the 
plant design or operation. As a result, the proposed change[s] [do] 
not affect any of the parameters or conditions that could contribute 
to the initiation of any accidents. The proposed changes only affect 
the methods used to monitor and [ensure] diesel generator 
performance. The performance criteria for both the diesel generator 
reliability and unavailability established by the maintenance rule, 
and the causal determinations and corrective actions required for 
maintenance preventable functional failures, [are] considered by 
[the Nuclear Regulatory Commission (NRC) in] GL [(Generic Letter)] 
94-01[, ``Removal of Accelerated Testing and Special Reporting 
Requirements for Emergency Diesel Generators,'' issued May 31, 
1994,] to be an acceptable method for monitoring diesel generator 
performance.
    No SSC, method of operation, or system interface is altered by 
[these] change[s]. The changes do not impact the diesel 
[generator]'s design sources, operating characteristics, system 
functions, or system interrelationships. The failure mechanisms for 
the accidents are not affected, and no additional failure modes are 
created. Because the diesel generator performance and reliability 
will continue to be [ensured] by the maintenance rule, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. This request does not involve a significant reduction in a 
margin [of] safety.
    The proposed changes only affect the methods used to monitor and 
[ensure] diesel generator performance and reliability. The 
performance criteria for the diesel generator reliability and 
availability established by the maintenance rule, and the causal 
determinations and corrective actions required for maintenance 
preventable functional failures, [are] considered by [NRC in] GL 94-
01 to be an acceptable method for monitoring diesel generator 
performance. No margin [of] safety as defined in the bases for any 
technical specification is impacted by these changes. [These] 
change[s] [do] not impact any uncertainty in the design, 
construction, or operation of any SSC. Diesel generator response to 
accident initiators is unchanged. No SSC, method of operating, or 
system interface is altered by [these] change[s]. The changes do not 
impact the diesel [generator]'s design sources, operating 
characteristics, system functions, or system interrelationships. 
Because the diesel generator performance and reliability will 
continue to be [ensured] by the maintenance rule, the proposed 
changes do not involve a significant reduction in the margin [of] 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: James W. Clifford, Acting

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 26, 1997
    Description of amendment request: The proposed amendment would 
revise the Crystal River Unit 3 (CR3) Technical Specifications Bases 
(TSB) to change the design basis of the Emergency Diesel Generator 
(EDG) Air Handling System. Specifically, TSB Sections B 3.8.1 and B 
3.8.2 would be revised to indicate that a single or dual fan operation 
depending upon fan supply air temperature, would maintain the 
temperature of the EDG engine and control rooms within the EDG 
manufacturer's limits.
    Basis for proposed no significant hazardsconsideration 
determination:
    The EDG Air Handling System provides continuous ventilation, and 
dissipates internal heat gains in the EDG engine and control rooms when 
the diesel is operating. Presently, the CR3 plant documentation 
requires operation of only one cooling fan per room to maintain the EDG 
room temperature within the manufacturer's limit and is inconsistent 
with the Final Safety Analysis Report (FSAR) which requires operation 
of two fans.
    As part of its EDG upgrade to increase their service ratings and 
associated cooling analysis, the licensee has determined that operation 
of either a single or dual cooling fans depending upon fan supply air 
temperature, would achieve the required room cooling limits. The 
licensee has determined that reliance on the operation of two cooling 
fans instead of one involves an unreviewed safety question and requires 
a license amendment.
    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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change does not involve a significant increase in 
the probability of an accident previously evaluated. The EDG room 
cooling fans support operation of the EDGs which are used to 
mitigate design basis accidents. Although EDG availability is a 
contributor to the risk of station blackout, the CR-3 licensing 
basis assumes a station blackout without regard to EDG reliability. 
Therefore, the probability of previously evaluated accidents is not 
significantly increased.
    For design basis accidents, the proposed change does not involve 
a significant increase in the consequences of an accident previously 
evaluated. The proposed change to operate both cooling fans for each 
EDG to

[[Page 50005]]

provide adequate ventilation potentially increases the probability 
of malfunction of equipment important to safety. However, the 
proposed changes do not affect the independence of the EDGs or the 
independence of the EDG Air Handling System and, based on single 
failure criteria, one EDG will be fully operable and capable of 
meeting its mission at all times as required by the CR-3 Technical 
Specifications. Therefore, no significant increase in the 
consequences of an accident previously evaluated, including the 
offsite radiological dose exists.
    Based on the above, the probability of an accident previously 
evaluated has not been significantly increased, and this change does 
not involve a significant increase in the consequences of an 
accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
Neither the fans nor the EDGs are initiators of any new accidents. 
The EDG room cooling fans support operation of the EDGs, which are 
used to mitigate design basis accidents. Reliance on two fans rather 
than one has reduced the redundancy of the EDG Air Handling System 
and increased the probability of a malfunction of an EDG. However, 
the proposed changes do not affect the independence of the EDGs or 
the independence of the EDG Air Handling System and, based on single 
failure criteria, one EDG will be fully operable and capable of 
meeting its mission at all times as required by the CR-3 Technical 
Specifications. Results of analyses to evaluate the failure of an 
EDG to operate following a design basis accident are documented in 
the FSAR. Therefore, this change does not create the possibility of 
a new or different kind of accident.
    3. Does not involve a significant reduction in the margin of 
safety
    The proposed change does not involve a significant reduction in 
the margin of safety. The EDG room cooling fans support operation of 
the EDGs. Following this change, two fans will be required to 
maintain the EDG engine room and EDG control room temperatures 
within the design basis limit when the fan supply air temperature is 
greater than or equal to 85 deg.F. Reliance on two fans rather than 
one has reduced the redundancy of the EDG Air Handling System and 
slightly increased the probability of malfunction of an EDG, but 
only after it has run for some period of time. However, the proposed 
changes do not affect the independence of the EDGs or the 
independence of the EDG Air Handling System and, based on single 
failure criteria, one EDG will be fully operable and capable of 
meeting its mission at all times as required by the CR-3 Technical 
Specifications. Therefore, this change does not result in a 
significant reduction to the margin 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 Project Director: Frederick J. Hebdon

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

    Date of amendment request: September 9, 1997
    Description of amendment request: The proposed amendment would 
revise the Crystal River 3 (CR3) Final Safety Analysis Report (FSAR) to 
reflect the revised analysis for the hypothetical Makeup System Letdown 
Line Failure Accident. In the original analysis, the event was modeled 
as being terminated by an automatic isolation of the failed letdown 
line on low reactor coolant system pressure. The revised analysis has 
modeled the event as being terminated by manual operator action to 
isolate the line. The licensee has determined that reliance on a manual 
operator action in place of the automatic action involves an unreviewed 
safety question (USQ) and requires prior Nuclear Regulatory Commission 
(NRC) approval. Other FSAR changes are being proposed to clarify that 
this accident is a hypothetical event that is presented only to 
demonstrate that the dose consequences are below 10 CFR Part 100 
limits. The licensee submitted its proposed FSAR changes which, upon 
NRC approval, will be incorporated in the next revision to the FSAR.
    Basis for proposed no significant hazardsconsideration 
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 not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This change involves a revision to the analysis for the Makeup 
System Letdown Line Failure Accident. The revised analysis assesses 
the resultant change in consequences of this event based on the 
actions specified in EOP-3 [Emergency Operating Procedure - 3] to 
manually isolate the letdown line failure. No changes have been made 
to any precursors to this event. Therefore, the probability of an 
accident previously evaluated has not been increased.
    This change has resulted in an increase in the calculated doses 
due to the greater release of reactor coolant prior to termination 
of the leak. Although the doses have increased, they remain 
significantly less than the limits of 10 CFR 100. These doses also 
remain lower than the resultant doses for the design basis LOCA 
[loss-of-coolant-accident].
    The revised analysis evaluates the consequences of this accident 
based on the replacement of the automatic isolation of the letdown 
line with a manual operator action to isolate the letdown line. This 
action was added to EOP-3 when it was identified that the manual 
initiation of the HPI [high pressure injection] system directed by 
the EOP would interfere with the automatic isolation signal assumed 
to terminate this event. Manual initiation of the HPI system for a 
LSCM [loss of subcooling margin] event is consistent with the 
symptomatic philosophy of the EOPs. This philosophy is utilized in 
order to manage a wide range of event/leaks that would be indicated 
by a LSCM. Early initiation of the HPI system is intended to ensure 
adequate core cooling as the primary concern during a LSCM event.
    Prior to the addition of the EOP step to manually isolate the 
letdown line, the EOP directed actions towards locating and 
isolating the source of the leak resulting in the LSCM. However, due 
to the potential significance of the letdown line failure which can 
result in RCS [reactor coolant system] leakage outside the reactor 
building, the manual action was added early in EOP-3 to isolate the 
letdown line. This action is proactive in ensuring early isolation 
of the potential leakage path and is consistent with the concept of 
a ``simple'' operator action (Reference 9) [NRC to Florida Power 
Corporation letter, Long-term modifications regarding emergency core 
cooling system Small Break Analysis problem, dated September 26, 
1978].
    Crediting a manual operator action instead of the automatic 
isolation introduces the possibility of a malfunction of a different 
type (i.e., operator error). The revised analysis assumes that 
operator action to isolate the letdown line occurs 10 minutes 
following a LSCM. Although the probability of operator error during 
this action may be greater than the probability of the failure of 
the automatic function, the consequences of this error would be 
small. Several indications would be available to the operator to 
identify the continued loss of coolant through this line. As 
discussed above, the radiological dose calculated by this event 
remains a small fraction of the limits of 10 CFR Part 100. 
Therefore, adequate time would exist for the identification of an 
operator error and correction of this error before any significant 
increase in the consequences of this event would occur.
    Additionally, the probability for operator error in this event 
is considered to be small due to the extensive training plant 
operators receive regarding the EOPs and the simple nature of the 
action. Validation of the required actions in the EOPs, including 
isolation of the letdown line, is performed on the plant simulator 
to ensure the validity of the EOPs as well as to ensure that these 
actions can be performed as required.

[[Page 50006]]

    The clarification added to FSAR Section 5.4.4.2 and 14.2.2.6.1 
reflects the previously approved evaluation for pipe rupture 
criteria outside the reactor building for CR-3. A break in the high 
energy portion of the letdown line outside containment is not 
considered a credible event. This accident is presented only to 
demonstrate that the dose consequences from a postulated break in 
the letdown line outside containment remain below the 10 CFR Part 
100 limits.
    Based on the above, this change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    This change does not involve any modification to the plant nor a 
change in the operation of the plant prior to the postulated failure 
of the letdown line. This change only evaluates the radiological 
dose consequences of the actions taken following the line failure. 
The addition of the action to manually isolate the letdown line for 
a LSCM event is consistent with the need to isolate potential RCS 
leakage paths and replaces the automatic isolation that was 
previously assumed to occur. Therefore, this change does not create 
the possibility of a new or different kind of accident.
    3. Does not involve a significant reduction in the margin of 
safety.
    This change does not result in a reduction to the margin of 
safety as defined in the Bases for any Technical Specifications. As 
discussed above, the radiological doses for the revised analysis 
have increased but remain a small fraction of the 10 CFR Part 100 
limits.
    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 Project Director: Frederick J. Hebdon

Florida Power and Light Company, et al., Docket Nos. 50-335 and 50-
389, St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: August 22, 1997
    Description of amendment request: The proposed amendment revises 
Technical Specification (TS) 4.0.5, Surveillance Requirements for 
Inservice Inspection and Testing of ASME Code Class 1, 2, and 3 
components, to relocate the Inservice Testing Program requirements from 
TS 4.0.5 to the Administrative Controls Section 6.8, Procedures and 
Programs. The proposed amendment also provides conforming changes to 
several Surveillance Requirements to change the reference from TS 4.0.5 
to the Inservice Testing Program.
    Basis for proposed no significant hazardsconsideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
There are no changes to the testing and evaluation related to pumps 
and valves in the Inservice Testing Program. The only substantive 
change allows the implementation of alternate testing provisions 
where Code-requirements are impractical and the NRC has not formally 
provided written approval. Since impractical testing would not be 
performed in any event, the actual testing program is unaffected.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The use of the modified specifications cannot create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
implementation of this administrative change since the proposed 
changes do not involve the addition or modification of equipment, 
nor do they alter the design or operation of affected plant systems, 
structures, or components.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
systems, structures, and components remain unchanged by the proposed 
amendments, therefore, these changes do 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: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420
    NRC Project Director: Frederick J. Hebdon

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

    Dates of amendment request: August 27, 1997
    Description of amendment request: The licensee proposed modifying 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to delete 
a sentence from section 6.2.2.f and add clarification to section 
6.2.2.f of the Administrative section of TS to allow the use of up to 
12 hour shifts without routine heavy use of overtime.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change does not involve a physical or procedural 
change to any structure, system or component that significantly 
effects the probability or consequences of any accident or 
malfunction of equipment important to safety. The proposed changes 
will allow the use of 12 hour shifts for a nominal 40 hours per 
week.
    This change is only administrative in nature and has no 
significant impact on the probabilities or consequences of any 
evaluated accident or malfunction of equipment important to safety.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed amendment will not change the physical plant or 
modes of plant operation defined in the Turkey Point Units 3 and 4 
operating license. The proposed amendment will not involve addition 
or modification of permanent equipment for any systems structures or 
components at Turkey Point.
    The change does modify the controls on working shift hours for 
operating personnel without significantly changing the hours worked 
per week and retains the current limitations on excessive overtime. 
The changes are administrative in nature.
    Consequently, operation of either unit in accordance with the 
proposed amendment would not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

[[Page 50007]]

    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The proposed amendment will allow the use of 12 hour shifts by 
virtue of the administrative change. This will result in fewer 
turnovers per day and will allow more contiguous days off between 
work shifts. The sum of these 12 hour work shift features will be 
more rested crews with better communications between shifts. The 
proposed change will not alter the basis for any Technical 
Specification that is related to the establishment of, or 
maintenance of, a nuclear safety margin.
    Consequently, operation of Turkey Point Units 3 and 4 in 
accordance with this proposed amendment would 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: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: Frederick J. Hebdon

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: August 18, 1997
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.1.6, Atmospheric Steam Relief 
Valves, to ensure the automatic feature of the steam generator power 
operated relief valve remains operable during Modes 1 and 2. In 
addition, the proposed change adds a surveillance requiring that a 
channel calibration on the steam generator power operated relief valve 
be performed every 18 months.
    Basis for proposed no significant hazardsconsideration 
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:
    The South Texas Project proposed to revise Technical 
Specification 3.7.1.6 to ensure the automatic feature of the Steam 
Generator Power Operated Relief Valve remains operable during Modes 
1 and 2. The South Texas Project has evaluated this proposed 
amendment and determined that it involves no significant hazards 
considerations based on the following:
    A. The proposed change does not 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 SBLOCA [small-break loss-of-coolant accident] event 
presented in the Updated Final Safety Analysis Report remain 
unchanged. The calculated Peak Clad Temperature remains 
substantially below the 2200 deg.F acceptance limit of 
10[]CFR[]50.46.
    B. The proposed change does not 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 the Technical Specification Amendment 
proposal.
    C. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change results in the calculated Peak Clad 
Temperature 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 South Texas Project has concluded that the 
proposed change does not involve a significant hazards 
considerations.
    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, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: James W. Clifford, Acting

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

    Date of amendment request: September 2, 1997
    Description of amendment request: The proposed changes to the 
Technical Specifications (TSs) would modify the maximum allowed 
containment pressure specified in TS 3.6.1.4, ``Containment Systems 
Internal Pressure,'' from 2.1 psig to 1.0 psig. The TS Bases, Section 
3/4.6.1.4, would also be revised to reflect the new maximum allowed 
containment pressure.
    Basis for proposed no significant hazardsconsideration 
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:
    The proposed change does not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will reduce the maximum allowed value for 
containment pressure specified in Technical Specification 3.6.1.4, 
``Containment Systems Internal Pressure.'' This change will improve 
the margin between the peak containment pressure following a main 
steam line break (most limiting accident for peak containment 
pressure at Millstone Unit No. 2) and the containment design 
pressure limit of 54 psig. Reducing the initial containment pressure 
will result in a reduction in peak containment pressure.
    To ensure the assumption of a lower initial containment pressure 
is maintained, a change to Technical Specification 3.6.1.4 is 
necessary.
    The proposed change to Technical Specification 3.6.1.4 will 
allow one of the initial assumptions used in the analysis for peak 
containment pressure following a main steam line break to be 
changed. However, this change will not affect how any of the plant 
systems function to mitigate design basis accidents and will not 
require any changes to mitigation procedures. The acceptance 
criteria of a peak containment pressure less than the design limit 
of 54 psig remains the same. Therefore, this change does not 
significantly increase 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 change does not alter the way any structure, 
system, or component functions and does not alter the manner in 
which the plant is operated. It does not

[[Page 50008]]

introduce any new failure modes and conservatively alters an 
assumption made in the main steam line break safety analysis.
    Therefore, the change 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.
    This proposed change will reduce the maximum allowed value for 
containment pressure specified in Technical Specification 3.6.1.4, 
``Containment Systems Internal Pressure.'' This change will improve 
the margin between the peak containment pressure following a main 
steam line break (most limiting accident for peak containment 
pressure at Millstone Unit No. 2) and the containment design 
pressure limit 54 psig. Starting at a lower initial containment 
pressure will result in a lower peak containment pressure. To ensure 
the assumption of a lower initial containment pressure is 
maintained, a change to Technical Specification 3.6.1.4 is 
necessary.
    This more restrictive change in the maximum allowed containment 
pressure will result in the use of a lower initial containment 
pressure in the analysis of a main steam line break accident. 
However, the analysis acceptance criteria of a peak accident 
containment pressure less than 54 psig, will remain the same. 
Therefore, there is no significant reduction in a margin of safety 
as defined in the Bases of Technical Specification 3.6.1.4.
    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, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270 NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: September 2, 1997
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to: (1) Combine TS 3.6.2.1, 
``Containment Spray System,'' and TS 3.6.2.2, ``Containment Air 
Recirculation System,'' into one specification which would reduce the 
allowed outage time for one inoperable containment spray (CS) train or 
one inoperable containment air recirculation (CAR) cooler from 30 days 
to 7 days; increase the allowed outage time for two inoperable CAR 
coolers from 48 hours to 7 days; add an allowed outage time of 48 hours 
(instead of entering TS 3.0.3) for one inoperable CS train and two 
inoperable CAR coolers or three or four inoperable CAR coolers; provide 
specific guidance on when to enter TS 3.0.3; and expand the applicable 
TS Bases to reflect these changes; (2) Modify the definition of 
containment integrity and TS 3.6.1.1, ``Containment Integrity,'' to 
indicate that the operability of the automatic isolation valve system 
is satisfied by the use of the containment isolation trip push buttons 
in Mode 4, and expand the TS Bases to reflect these change; (3) Add an 
exception to the reactor coolant flow rate surveillance requirement, TS 
4.1.1.3, whenever there is a reduction in reactor coolant system 
boration while in Modes 2 and 3 because the reactor coolant pumps are 
required to be in operation; (4) Delete the reactor coolant system 
leakage surveillance requirements, TS 4.4.6.2.a and TS 4.4.6.2.b, which 
require monitoring the containment atmosphere particulate radioactivity 
and containment sump inventory, respectively; (5) Modify emergency core 
cooling system surveillance requirement, TS 4.5.2.e, to allow the use 
of alternative methods to verify that the throttle valves in Table 4.5-
1 are in the correct position and expand the TS bases to address the 
alternative methods; (6) Modify TS 5.5.1,'' Emergency Core Cooling 
Systems,'' by deleting the word ``original'' since the design has been 
modified; and (7) Make editorial changes to terminology and item 
numbering.
    Basis for proposed no significant hazardsconsideration 
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:
    The proposed changes do not involve an SHC [significant hazards 
consideration] because the changes would not:
    1.Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to combine Technical Specifications 3.6.2.1 
and 3.6.2.2 into one specification reduces the allowed outage time 
for one inoperable containment spray (CS) train or one inoperable 
containment air recirculation (CAR) cooler from 30 days to 7 days; 
increases the allowed outage time for two inoperable CAR coolers 
from 48 hours to 7 days; adds an allowed outage time of 48 hours 
(instead of entering Technical Specification 3.0.3) for one 
inoperable CS train and two inoperable CAR coolers, or three or four 
inoperable CAR coolers; and provides specific guidance when it is 
necessary to enter Technical Specification 3.0.3 will not affect how 
these systems function to mitigate design basis accidents. 
Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed changes to modify the definition of containment 
integrity, modify the Technical Specification 3.6.1.1, ``Containment 
Integrity,'' and expand the Bases to explain why automatic 
containment isolation valves are operable in Mode 4 have no affect 
on any containment isolation valve or Engineered Safety Feature 
Actuation System (ESFAS) component. These components will still 
function as designed to mitigate design basis accidents. Therefore, 
this change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to provide an exception to Surveillance 
Requirement 4.1.1.3 when the plant is in Modes 1 and 2 will not 
result in any new approach to plant operation, it simply removes the 
requirement to perform an unnecessary surveillance. The minimum 
coolant flow through the core during a reduction in Reactor Coolant 
System (RCS) boron concentration will still be met. Therefore, this 
change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    The proposed change to delete Surveillance Requirements (SRs) 
4.4.6.2.a and 4.4.6.2.b does not reduce the operability requirements 
for any equipment used to monitor RCS leakage. The equipment covered 
by these 2 SRs, containment atmosphere particulate radioactivity 
monitors and containment sump inventory monitor, provide early 
indication that RCS leakage exists, but do not provide the specific 
information (amount of leakage) necessary to verify operation within 
the leakage limits contained in Technical Specification 3.4.6.2, 
``Reactor Coolant System Leakage.'' Operability of the containment 
atmosphere particulate radioactivity monitors and containment sump 
inventory monitor is verified by SRs 4.4.6.1.a and 4.4.6.1.b. 
Therefore, this change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The proposed change to Surveillance Requirement 4.5.2.e. to 
allow the use of alternate methods does not reduce operability or 
surveillance requirements for any of the Emergency Core Cooling 
System (ECCS) throttle valves. Therefore, these ECCS throttle valves 
will continue to function as designed to mitigate design basis 
accidents. Therefore, this change does not significantly increase 
the probability or consequences of an accident previously evaluated.
    The proposed change to Technical Specification 5.5.1 has no 
affect on how the ECCS operates. The ECCS will still function as 
designed to mitigate design basis accidents. Therefore, this change 
does not significantly increase the probability or consequences of 
an accident previously evaluated.

[[Page 50009]]

    The proposed changes to add information to the Bases of the 
affected Technical Specifications, and make editorial changes to 
terminology and item numbering will have no affect on equipment 
operation. Therefore, all associated equipment will continue to 
function as designed to mitigate design basis accidents. Therefore, 
this change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    Thus, 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. They will not alter assumptions 
made in the safety analysis and licensing basis. The affected 
components and systems will still function as designed to mitigate 
design basis accidents.
    Therefore, these changes do 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 will not reduce the margin of safety since 
they have no impact on any safety analysis assumption. The proposed 
changes do not decrease the scope of equipment currently required to 
be operable or subject to surveillance testing, nor do the proposed 
changes affect any instrument setpoints or equipment safety 
functions. The requirement to check containment radiation and 
containment sump level every 12 hours has been eliminated. However, 
this equipment is still required to be operable, and the 
surveillance requirements to verify operability have not been 
changed. Therefore, this equipment will be available to provide 
early indication of RCS leakage.
    The effectiveness of Technical Specifications will be maintained 
since the changes will not alter the operation of any component or 
system. In addition, the changes are consistent with the new, 
improved Standard Technical Specifications (STS) for Combustion 
Engineering plants (NUREG-1432).
    Therefore, there is not 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: 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
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: September 3, 1997
    Description of amendment request: The proposed amendment would 
revise the Updated Final Safety Analysis Report (UFSAR) by changing the 
length of time the emergency diesel generators (EDGs) would operate 
following a loss-of-coolant accident (LOCA) based on the capacity of 
the onsite diesel fuel oil supply required by the current Technical 
Specifications (TSs). The UFSAR indicates that the diesel fuel oil 
supply tanks contain a sufficient amount of fuel to operate one EDG for 
about 7 days and the other EDG 1 hour following a LOCA based on the TS 
minimum limit of 24,000 gallons of diesel fuel oil stored onsite. 
Northeast Nuclear Energy Company (the licensee) has performed 
calculations indicating that both EDGs can initially operate, following 
a LOCA, for 24 hours and one EDG can continue to operate for an 
additional 3.5 days based on the TS requirement to have a minimum of 
24,000 gallons of fuel oil stored onsite. The licensee has determined 
that the difference in the EDGs operating time, as a result of the new 
calculations, constitutes an unreviewed safety question and requests 
approval to revise the UFSAR.
    Specifically, the proposed license amendment would revise the 
UFSAR, Section 8.3, ``Emergency Generators,'' to reflect the operating 
times for the EDGs based on the TS-required onsite fuel oil supply. 
Additional requirements would also be added indicating that the 
existing nonsafety-related underground fuel oil storage tank would be 
required to maintain about 17,700 gallons of fuel oil when the unit is 
operating in Modes 1 through 4. This requirement would be included in 
the Technical Requirements Manual, which also will require that the 
amount of stored fuel oil be verified by surveillance requirements 
similar to the TS-required surveillances for the safety-related fuel 
oil supply. This change will increase the total time that one EDG can 
continue to operate following a LOCA from 3.5 to 7 days. The Emergency 
Plan (EP) procedures require that an evaluation be performed within 4 
hours following a LOCA or loss of normal power (offsite power) to 
determine if additional fuel oil is needed from an offsite source. The 
licensee has a contract with a supplier for the delivery of fuel oil to 
the Millstone site. The EP procedures also require that load shedding 
recommendations be made within 24 hours. The recommendations will vary 
depending on the situation and are another way to extend the operating 
times for the EDGs.
    Basis for proposed no significant hazardsconsideration 
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:
    The proposed change does not involve an SHC [significant hazards 
consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change expands FSAR Section 8.3, ``Emergency 
Generators,'' to discuss the length of time the emergency diesel 
generators (EDGs) will operate following a loss of coolant accident 
(LOCA) and a loss of normal power (LNP), utilizing only onsite 
diesel fuel oil sources. The onsite sources include the Technical 
Specification required volume of 12,000 gallons in each diesel oil 
supply tank and an additional approximate 17,700 gallons that will 
be maintained in the underground diesel oil storage tank. This 
onsite volume of diesel fuel oil is sufficient to allow two EDGs to 
operate at rated load (2750 KW) for 24 hours following a design 
basis LOCA and LNP. The remaining diesel fuel oil will be sufficient 
for one EDG to continue operation at rated load for a total of 7 
days from event initiation.
    The proposed change to the FSAR has no effect on EDG operation 
and reliability. The EDGs will continue to operate as designed to 
supply the electrical loads assumed to mitigate the design basis 
accidents. Therefore, there is 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 change will not alter the plant configuration (no 
new or different type of equipment will be installed) or require any 
new or unusual operator actions. Plant operating procedures will be 
changed. However, the changes will not require the performance of 
any task not currently performed by the plant operators. Emergency 
Plan procedures already specify the action to provide load shedding 
recommendations within 24 hours of a LOCA and LNP, and to evaluate 
the need to order additional fuel from offsite sources within four 
hours after the accident.
    The proposed change does not alter the way any structure, 
system, or component

[[Page 50010]]

functions and does not alter the manner in which the plant is 
operated. It does not introduce any new failure modes and does not 
alter assumptions made in the safety analysis.
    Therefore, the change 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 length of time the emergency diesel generators (EDGs) will 
operate following a Loss of Coolant Accident and a Loss of Normal 
Power, utilizing only the onsite diesel fuel oil sources required by 
Technical Specifications has been recalculated. The new EDG run 
times do not agree with the current EDG run times contained in the 
Millstone Unit No. 2 Final Safety Analysis Report (FSAR), and 
therefore do not agree with the current Technical Specification 
Bases for 3.8.1.1, ``A.C. Sources - Operating,'' and 3.8.1.2, ``A.C. 
Sources - Shutdown.''
    This deviation does result in a reduction in the margin of 
safety as defined in the Technical Specification Bases for 3.8.1.1, 
``A.C. Sources - Operating,'' and 3.8.1.2, ``A.C. Sources - 
Shutdown.'' However, this proposed change will require additional 
diesel fuel oil to be maintained onsite in the non-seismic 
underground diesel oil storage tank. This will ensure sufficient 
diesel fuel oil will be maintained onsite to provide a 7 day supply, 
assuming a seismic event does not occur. Therefore, this is not a 
significant reduction in the margin of safety as defined in the 
Technical Specification Bases for 3.8.1.1, ``A.C. Sources - 
Operating,'' and 3.8.1.2, ``A.C. Sources - Shutdown.''
    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, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Deputy Director: Phillip F. McKee

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

    Date of amendment request: August 20, 1997
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TSs) to provide for: (1) the 
relocation of suppression pool volume references in Limiting Condition 
for Operation (LCO) 3.5.3 to the Hope Creek (HC) Updated Final Safety 
Analysis Report (UFSAR) and TS Bases as appropriate; (2) the revision 
of the suppression pool volume currently listed in LCO 3.5.3.b; (3) the 
relocation of the suppression pool volume references in LCO 3.6.2.1.a.1 
to the UFSAR and TS Bases; and (4) the revision to the suppression pool 
volume reference in TS 5.2.1 to reference the TS Bases section where 
this information will reside.
    Basis for proposed no significant hazardsconsideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS revisions involve: 1) no changes to the 
operation of any systems or components in normal or accident 
operating conditions; and 2) no significant changes to existing 
structures, systems or components. The installation of the new 
strainers will be justified separately using the provisions of 
10CFR50.59. The relocation of Technical Specification references to 
suppression pool volume to the UFSAR and/or TS Bases will not 
adversely impact the safety-related functions of the suppression 
pool or its supported systems since any changes to suppression pool 
volume will be subject to 10CFR50.59 provisions. The impact of the 
new strainers on ECCS [emergency core cooling system] performance in 
Operational Conditions 4 and 5 has been determined to be negligible, 
with less than a 0.3% decrease in suppression pool water volume at 
the minimum specified suppression pool water level limit. In 
addition, suppression pool volume is not a parameter involved in the 
initiation of any accident. Therefore these changes will not 
significantly increase the probability of an accident previously 
evaluated. To the extent practicable, these proposed changes were 
developed consistent with the changes approved by the NRC when 
developing NUREG-1433, ``Standard Technical Specifications, General 
Electric Plants, BWR/4'', with the intent of having the relocated 
information controlled in other plant documents subject to 
10CFR50.59 provisions. Since the plant systems associated with these 
proposed changes will still be capable of: 1) meeting all applicable 
design basis requirements; and 2) retain the capability to mitigate 
the consequences of accidents described in the HC UFSAR, the 
proposed changes were determined to be justified. Therefore, these 
changes will not involve a significant increase in the consequences 
of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Neither the relocation of Technical Specification references to 
suppression pool volume nor the revision of the suppression pool 
volume references for Operational Conditions 4 and 5 (COLD SHUTDOWN 
and REFUELING) will adversely impact the operation of any safety 
related component or equipment. Since the proposed changes involve: 
1) no changes to the operation of any systems or components; and 2) 
no significant changes to existing structures, systems or 
components, there can be no impact on the occurrence of any 
accident. To the extent practicable, these proposed changes were 
developed consistent with the changes approved by the NRC when 
developing NUREG-1433, ``Standard Technical Specifications, General 
Electric Plants, BWR/4'', with the intent of having the relocated 
information controlled in other plant documents subject to 
10CFR50.59 provisions. Furthermore, there is no change in plant 
testing proposed in this change request which could initiate an 
event. Therefore, these changes will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Removal and relocation of the Technical Specification references 
to suppression pool volume is consistent, to the extent practicable, 
with the changes approved by the NRC when developing NUREG-1433, 
``Standard Technical Specifications, General Electric Plants, BWR/
4''. The information retained in the Technical Specifications for 
minimum suppression pool water level and the information retained in 
the UFSAR and Technical Specification Bases will ensure that the 
suppression pool and supported components will remain capable of 
performing their intended safety functions. Any changes to 
suppression pool volume information retained in the UFSAR or 
Technical Specification Bases will be subject to the provisions of 
10CFR50.59 and a separate safety evaluation would be developed to 
support any proposed changes that would subsequently be made. The 
impact of the new strainers on ECCS performance in Operational 
Conditions 4 and 5 has been determined to be negligible, with less 
than a 0.3% decrease in suppression pool water volume in the minimum 
specified suppression pool water level limit. By retaining the 5 
inch minimum suppression pool water level limit within the TS, 
adequate provisions for: 1) NPSH [net-positive suction head] for 
ECCS pump suction; 2) recirculation volume; and 3) vortex prevention 
are maintained. Therefore, the changes contained in this request do 
not result in 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, NJ 08070
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21,

[[Page 50011]]

P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: August 19, 1997
    Description of amendment request: The proposed amendment would 
revise the Ginna Station Improved Technical Specifications (ITSs) by 
adding a note to the Containment Spray (CS) Limiting Condition for 
Operation (LCO) 3.6.6 which would allow the CS pumps in MODE 4 to be 
placed in pull-stop, and motor-operated valves (MOVs) 896A and 896B to 
have their DC control power restored with the valves placed in the 
closed position in order to perform interlock and valve testing of MOVs 
857A, 857B, and 857C. A time limit of 2 hours is placed on this 
configuration for each test.
    Basis for proposed no significant hazardsconsideration 
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 Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The change is 
to add a note to LCO 3.6.6 which allows the CS pumps to be placed in 
pull-stop and MOVs 896A and 896B to have power restored and closed 
in MODE 4. This does not increase the probability of any accident 
previously evaluated since the CS system provides mitigation 
capability only (i.e., does not initiate any accident). In addition, 
there is no design basis accident previously evaluated in MODE 4 
which would require the use of CS. Therefore, these changes do not 
involve a significant increase in the probability or consequences of 
an accident previously analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of plant 
safety because the CS function is not required for any design basis 
accident in MODE 4. In addition, time restraints [are] placed on the 
proposed plant configuration. As such, no question of safety is 
involved, and the change does not involve a significant reduction in 
a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005
    NRC Project Director: Alexander W. Dromerick, Acting Director

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

    Date of amendment request: August 2, 1996 (TXX-96434)
    Brief description of amendments: The proposed changes would 
increase the allowed outage time (AOT) for a centrifugal charging pump 
from 72 hours to 7 days.
    Basis for proposed no significant hazardsconsideration 
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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    There is no effect on the probability of an event; the only 
potential effect is on the capability to mitigate the event. The 
centrifugal charging pumps are credited in the Final Safety Analysis 
Report Chapter 15 LOCA analysis for ECCS injection and for the 
containment sump recirculation mode for the design-basis LOCA. 
Increasing the AOT for the centrifugal charging pumps does not 
affect analysis assumptions regarding functioning of required 
equipment designed to mitigate the consequences of accidents. 
Further, the severity of postulated accidents and resulting 
radiological effluent releases will not be affected by the increased 
AOT.
    A reliability analysis of the charging system found the change 
to have no significant impact on normal operation or on the RCP seal 
cooling function. Therefore, the change would not significantly 
increase in the probability of a seal LOCA.
    The change potentially affects only the availability of the 
charging system for accident mitigation and has no effect on the 
ability of other ECCS systems to perform their functions. Through 
the use of a probabilistic risk assessment, it was determined that 
the proposed change would have an insignificant effect on the core 
damage frequency.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different type of accident from any accident previously evaluated?
    Unavailability of one centrifugal charging pump for a finite 
period of time is currently allowed by the Technical Specifications. 
Increasing the AOT from 72 hours to 7 days would not change the 
method that TU Electric operates CPSES, thus would not create a new 
condition. Further, the proposed change would not result in any 
physical alteration to any plant system, and there would not be a 
change in the method by which any safety related system performs its 
function. The ECCS would still be capable of mitigating the 
consequences of the design-basis accident LOCA with the one 
centrifugal charging pump operable. No new unanalyzed accident would 
be created.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed change does not impact either the physical 
protective boundaries or performance of safety systems for accident 
mitigation. There is no safety analysis impact since the extension 
of the centrifugal charging pump AOT interval will have no effect on 
any safety limit, protection system setpoint, or limiting condition 
of operation. There is no hardware change that would impact existing 
safety analysis acceptance criteria, therefore there is no 
significant change in the margin of safety.
    In summary, the proposed change would not have a significant 
impact on the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and 
Bockius, 1800 M Street, N.W., Washington, DC 20036
    NRC Project Director: James W. Clifford, Acting

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

    The following notices were previously published as separate 
individual notices. The notice content was the same as above. They were 
published as individual notices either because time did not allow the 
Commission to wait for this biweekly notice or because the

[[Page 50012]]

action involved exigent circumstances. They are repeated here because 
the biweekly notice lists all amendments issued or proposed to be 
issued involving no significant hazards consideration.
    For details, see the individual notice in the Federal Register on 
the day and page cited. This notice does not extend the notice period 
of the original notice.

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

    Date of amendment request: September 5, 1997 (NRC-97-0107)
    Description of amendment request: The proposed amendment would add 
Special Test Exception 3/4.10.7, ``Inservice Leak and Hydrostatic 
Testing,'' that allows the performance of pressure testing at a reactor 
coolant temperature up to 212  deg.F while remaining in Operational 
Condition 4. This special test exception would also require that 
certain Operational Condition 3 specifications for Secondary 
Containment Isolation, Secondary Containment Integrity, Secondary 
Containment Automatic Isolation Dampers, and Standby Gas Treatment 
System operability be met. This change would also revise the Index, 
Table 1.2, ``Operational Conditions,'' and the Bases to incorporate the 
reference to the proposed special test exception. The licensee 
requested that this amendment be reviewed under exigent circumstances.
    Date of individual notice in the Federal Register: September 12, 
1997 (62 FR 48113)
    Expiration date of individual notice: October 14, 1997 NSHC 
comments: September 29, 1997
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226
    NRC Project Director: John N. Hannon

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: August 14, 1997
    Brief description of amendment request: The proposed amendments 
would revise the allowed tolerance of the reactor coolant system volume 
provided in Technical Specification 5.4.2 to account for steam 
generator tube plugging.
    Date of individual notice in the Federal Register: August 26, 1997 
(62 FR 45278)
    Expiration date of individual notice: September 25, 1997
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488

Notice Of Issuance Of Amendments ToFacility 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 Nos. 50-325 and 50-
324, Brunswick Steam Electric Plant, Units 1 and 2, Brunswick 
County, North Carolina

    Date of application for amendments: January 15, 1997, as 
supplemented on August 22, 1997.
    Brief description of amendments: The amendments revise the minimum 
and maximum allowed values in Technical Specification 3.6.2.1 for 
suppression chamber water volume. The amendments correct an error 
identified by Carolina Power & Light Company in the previous 
calculation of water volume and correct an error in the value listed in 
the associated TS Bases for Unit 1 for primary system operating 
pressure.
    Date of issuance: August 28, 1997
    Effective date: August 28, 1997
    Amendment Nos.: 186 and 217
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14458) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated August 28, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

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

    Date of application for amendments: December 4, 1996
    Brief description of amendments: The amendments revise the approach 
in Technical Specification 3/4.1.2 for determining a reactivity anomaly 
by changing from control rod density comparison to direct comparison of 
reactivity status.
    Date of issuance: September 5, 1997
    Effective date: September 5, 1997
    Amendment Nos.: 187 and 218
    Facility Operating License Nos. DPR-71 and DPR-62: Amendments 
change the Technical Specifications
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11484) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 5, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at

[[Page 50013]]

Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: January 24, 1997
    Brief description of amendments: The amendments revise the 
Technical Specification (TS) required surveillance calibration to be 
performed on the reactor water level instrumentation to reflect the 
modifications made to the Unit 3 instrumentation. The modifications 
were made during the recent Unit 3 refueling outage to improve the 
reliability of emergency core cooling system (ECCS) initiation on low 
low reactor water level. The surveillance requirement for calibration 
of the new level instrumentation is consistent with the ECCS low 
reactor water level initiation transmitter calibration requirements of 
NUREG 1433, ``Standard Technical Specifications, General Electric 
Plants, BWR/4'' for similar instrumentation. The same TS change for 
Unit 2 has been previously reviewed and approved by the NRC staff in 
Amendment No. 145 dated June 28, 1996. In addition minor editorial 
changes were made to the TS.
    Date of issuance: September 10, 1997
    Effective date: September 10, 1997, with full implementation within 
60 days.
    Amendment Nos.: 162 and 157
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 18, 1997 (62 FR 
19143) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 10, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450

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

     application for amendment: December 15, 1994, as revised July 25, 
1996, and supplemented December 13, 1996, and June 18, 1997
    Brief description of amendment: The amendment revises Technical 
Specification (TS) Section 6.0, Administrative Controls, by (1) 
removing requirements that are adequately controlled by existing 
regulations other than 10 CFR 50.36 and the TS and (2) relocating 
selected requirements from TS Section 6.0 to licensee-controlled 
documents or programs.
    Date of issuance: September 10, 1997
    Effective date: September 10, 1997, with full implementation within 
90 days. Implementation of this amendment shall include the relocation 
of the TS requirements to the appropriate licensee-controlled 
documents, as described in the licensee's application dated December 
15, 1994, as revised July 25, 1996, and supplemented December 13, 1996, 
and June 18, 1997, and evaluated in the staff's safety evaluation dated 
September 10, 1997.
    Amendment No.:  113
    Facility Operating License No. NPF-43. Amendment revises the TS.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29873) and August 14, 1996 (61 FR 42279). The December 13, 1996, and 
June 18, 1997, letters provided clarifying information within the scope 
of the original application and did not change the staff's initial 
proposed no significant hazards considerations determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 10, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161

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

    Date of application for amendment: March 10, 1997
    Brief description of amendment: The amendment modifies the 
Technical Specifications (TSs) by reducing the reactor coolant system 
specific activity limits in accordance with the NRC's guidance provided 
in Generic Letter 95-05, ``Voltage-Based Repair Criteria for 
Westinghouse Steam Generator Tubes by Outside Diameter Stress Corrosion 
Cracking.'' The definition of DOSE EQUIVALENT I-131 is replaced with 
the Improved Standard TS definition in the first sentence and an 
equation is added based on dose conversion factors derived from the 
International Commission on Radiation Protection (ICRP) ICRP-30. TS 
3.4.8, Specific Activity, is revised by reducing the DOSE EQUIVALENT I-
131 limit from 1.0 micro Ci/gram to 0.35 micro Ci/gram for the 48-hour 
limit and from 60 micro Ci/gram to 21 micro Ci/gram for the maximum 
instantaneous limit. Item 4.a in TS Table 4.4-12, Primary Coolant 
Specific Activity Sample and Analysis Program, TS Figure 3.4-1, and the 
Bases for TS 3/4.4.8 are also modified to reflect the reduced DOSE 
EQUIVALENT I-131 limit.
    Date of issuance: September 10, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No: 205
    Facility Operating License No. DPR-66. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24985) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 10, 1997. No 
significant hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

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

    Date of application for amendment: August 4, 1997, as supplemented 
August 16, 1997.
    Brief description of amendment: Temporary change to Technical 
Specification Surveillance Requirement (SR) 3.3.8.1. The change will 
allow the licensee to extend the frequency of SR 3.3.8.1 from 31 to 60 
days.
    Date of issuance: August 29, 1997
    Effective date: August 29, 1997
    Amendment No.: 157
    Facility Operating License No. DPR-72. Amendment temporarily 
revises Technical Specifications Surveillance Requirement 3.3.8.1.
    Date of initial notice in Federal Register: August 12, 1997 (62 FR 
43189) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated August 29, 1997. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

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 17, 1996, as supplemented June 14, 
1996, March 17, July 29, and July 30, 1997
    Brief description of amendments: The amendments modify Technical

[[Page 50014]]

Specification Section 3/4.4.5 Steam Generators, 3/4.4.6 Reactor Coolant 
System Leakage, and associated Bases to allow the installation of tube 
sleeves as an alternative to plugging to repair defective steam 
generator tubes.
    Date of issuance: September 4, 1997
    Effective date: September 4, 1997
    Amendment Nos.: Unit 1 - Amendment No. 90; Unit 2 - Amendment No. 
77
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 29, 1996 (61 FR 
25938) and April 9, 1997 (62 FR 17235). The June 14, 1996, and July 29, 
and July 30, 1997, submittals provided additional information that did 
not affect the staff's initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 4, 1997. 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

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: June 30, 1997
    Brief description of amendment: Technical Specification 
Surveillance Requirements 4.7.1.5.1 and 4.7.1.5.2 require the periodic 
testing of the main steam isolation valves (MSIVs) to demonstrate 
operability. The amendment (1) clarifies when the MSIVs are partial 
stroked or full closure tested, (2) adds a note to the Mode 4 
applicability of Technical Specification 3.7.1.5 to require that the 
MSIVs be closed and deactivated at less than 320 degrees F, (3) makes 
editorial changes, and (4) makes changes to the associated Bases 
sections.
    Date of issuance: September 3, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 148
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 ( 62 FR 
40853) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 3, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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: May 14, 1997, as supplemented by 
letter dated July 30, 1997
    Brief description of amendment: Technical Specification 
Surveillance Requirement 4.8.2.1.c.4 requires that each battery charger 
be tested to verify that it can supply a specified current at 125 
volts. The amendment increases the required test voltage.
    Date of issuance: September 5, 1997
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 149
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33130) The July 30, 1997, letter provided clarifying information that 
did not change the scope of the May 14, 1997, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated September 5, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385

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: May 9, 1997, as supplemented by 
letter dated July 14, 1997
    Brief description of amendments: The proposed change revises the 
Peach Bottom Atomic Power Station, Units 2 and 3, technical 
specifications to extend the interval for replacing the primary 
containment purge and exhaust valve inflatable seals.
    Date of issuance: September 4, 1997
    Effective date: Both units, as of date of issuance, to be 
implemented within 30 days.
    Amendments Nos.: 220 and 223
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35851) The supplemental letter provided clarifying information that did 
not change the original no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 4, 1997. 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

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 application for amendments: September 25, 1996
    Brief description of amendments: These amendments (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).
    Date of issuance: September 3, 1997
    Effective date: As of the date of issuance, to be implemented 
within 30 days of issuance.
    Amendment Nos.: 168 and 142
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 19, 1996 (61 
FR 58904) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated

[[Page 50015]]

September 3, 1997. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701

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: April 25, 1997, as supplemented 
June 6, 1997
    Brief description of amendments: The amendments revise Technical 
Specification 3.5.2 to eliminate reference to the flow path from the 
residual heat removal system to the reactor coolant system hot legs. 
This flow path is being eliminated to prevent excessive flow through 
the residual heat removal system during all hot leg recirculation 
configurations assuming worst-case single failures that could result in 
excessive flow during hot leg recirculation following a loss-of-coolant 
accident.
    Date of issuance: September 11, 1997
    Effective date: Both units, as of the date of issuance, to be 
implemented within 60 days of issuance.
    Amendment Nos.: 200 and 184
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 14, 1997 (62 FR 
26574) The June 6, 1997, supplement provided 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 September 11, 
1997. No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079Sacramento Municipal Utility District, 
Docket No. 312, Rancho Seco Nuclear Generating Station, Sacramento 
County, California
    Date of application for amendment: December 9, 1993, as superseded 
December 19, 1995, and as supplemented on January 22, 1996.
    Brief description of amendment: This amendment changes the 
Technical Specifications to incorporate the revised 10 CFR Part 20, 
Standards for Protection Against Radiation. The amendment corrects 
references from Semiannual Radioactive Effluent Release Report to 
Annual Radioactive Effluent Release Report. The amendment also corrects 
references from NRC Region V to NRC Region IV.
    Date of issuance: August 22, 1997
    Effective date: August 22, 1997
    Amendment No.: 125
    Facility Operating License No. NPF-1: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 2, 1994 (59 FR 
10015) The information provided in the licensee's letters of December 
19, 1995 and January 22, 1996 contained editorial changes and did not 
involve significant changes to the original Federal Register notice. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated August 22, 1997. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: : Central Library, Government 
Documents, 828 I Street, Sacramento, California 95814

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: January 7, 1997, as 
supplemented July 2, 1997

    Brief description of amendments: The amendments revise plant 
Technical Specifications associated with surveillance requirements 
testing that requires manually actuating every safety/relief valve 
during each unit startup from a refueling outage.
    Date of issuance: September 5, 1997
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 208 and 150
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4350) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 5, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513

Tennessee Valley Authority, Docket No. 50-260 Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of application for amendment: June 2, 1995, revised March 3, 
1997, as supplemented May 13 and August 20, 1997 (TS 353)
    Brief description of amendment: The amendment provides technical 
specification (TS) changes for an upgrade of the power range neutron 
monitor instrumentation. Changes to thermal limits specifications were 
also proposed to implement average power range monitor and rod block 
monitor ts improvements, and maximum extended load line limit analyses.
    Date of issuance: September 11, 1997
    Effective Date: September 11, 1997
    Amendment No.: 249
    Facility Operating License No. DPR-52: Amendment revised the TS.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42609) The March 3, 1997 revision, as supplemented May 13 and August 
20, 1997, does not affect the staff's proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 11, 1997. No significant hazards 
consideration comments received: None.
    Local Public Document Room location: Athens Public library, 405 E. 
South Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket No. 50-390 Watts Bar Nuclear 
Plant, Unit 1, Rhea County, Tennessee

    Date of application for amendment: March 27, 1997, as supplemented 
May 28, June 4, and July 30, 1997.
    Brief description of amendment: The amendment pertains to Cycle 2 
core design changes and provides operational enhancements for reactor 
trip setpoints. Part 1 addresses an increase in the containment sump 
boron concentration during a large break loss-of-coolant accident and 
describes changes to Technical Specification (TS) 3.5.1 and 3.5.4 
regarding boron concentration. Part 2 addresses changes to TS Figure 
2.1.1-1, TS Table 3.3.1-1, and TS 3.4.1 on safety limits, the trip 
system and pressure, temperature and flow limits, respectively.
    Date of issuance: September 11, 1997
    Effective date: Sepember 11, 1997
    Amendment No.: 7
    Facility Operating License No. NPF-90: Amendment revises the TS.
    Date of initial notice in Federal Register: July 2, 1997 (62 FR 
35852) The July 30, 1997 submittal provided clarifying information 
which did not affect the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated September 11, 1997. No significant hazards 
consideration comments received: None

[[Page 50016]]

    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, TN 37402

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear, 
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 
50-440 Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: January 31, 1997, supplemented 
August 6, 1997.
    Brief description of amendment: The amendment approves the use of 
Option B, ``Performance-Based Requirements,'' to 10 CFR Part 50, 
Appendix J, ``Primary Reactor Containment Leakage Testing for Water-
Cooled Power Reactors.''
    Date of issuance: September 9, 1997
    Effective date: September 9, 1997
    Amendment No.: 86
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 12, 1997 (62 FR 
11492). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 9, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, OES Nuclear, 
Inc., Pennsylvania Power Company, Toledo Edison Company, Docket No. 
50-440 Perry Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: May 2, 1997
    Brief description of amendment: The amendment allows the leakage 
rate of one or more main steam lines to be up to 35 standard cubic feet 
per hour (scfh), as long as the total leakage rate through all four 
main steam lines is less than or equal to 100 scfh.
    Date of issuance: September 11, 1997
    Effective date: September 11, 1997
    Amendment No.: 87
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33136). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated September 11, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendments: April 14, 1997 (TSCR 198)
    Brief description of amendments: These amendments revise Technical 
Specification Section 15.3.1, ``Reactor Coolant System,'' to eliminate 
the provisions for operation of the units at below 3.5 percent rated 
power with a single reactor coolant pump.
    Date of issuance: September 3, 1997
    Effective date: September 3, 1997, with full implementation within 
45 days
    Amendment Nos.: 178 and 182
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: May 21, 1997 (62 FR 
27802) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 3, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241

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

    Date of application for amendments: January 24, 1997, as 
supplemented on May 15 and August 5, 1997 (TSCR 193)
    Brief description of amendments: These amendments revise TS 15.5.4, 
``Fuel Storage,'' to increase fuel assembly enrichment limits to 5.0 
weight percent uranium-235 while maintaining Keff in the 
storage pools (spent fuel pool and new fuel storage racks) less than 
0.95. Date of issuance: September 4, 1997
    Effective date: September 4, 1997, with full implementation within 
45 days
    Amendment Nos.: 179 and 183
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30647) The August 5, 1997, submittal provided clarifying information 
within the scope of the original application and did not affect the 
staff's initial proposed no significant hazards considerations 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated September 4, 1997. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: The Lester Public Library, 
1001 Adams Street, Two Rivers, Wisconsin 54241

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the

[[Page 50017]]

plant's licensed power level, the Commission may not have had an 
opportunity to provide for public comment on its no significant hazards 
consideration determination. In such case, the license amendment has 
been issued without opportunity for comment. If there has been some 
time for public comment but less than 30 days, the Commission may 
provide an opportunity for public comment. If comments have been 
requested, it is so stated. In either event, the State has been 
consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By October 24, 1997, 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. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

[[Page 50018]]

Arizona Public Service Company, et al., Docket No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit No. 2, Maricopa County, 
Arizona

    Date of application for amendment: August 28, 1997, as supplemented 
by letter dated September 3, 1997.
    Brief description of amendment: The amendment revises Technical 
Specification Table 4.3-2 to allow for a one-time, five-day extension 
of the required surveillance interval for the main steam isolation 
system portion of the engineered safety feature actuation system logic.
    Date of issuance: September 4, 1997
    Effective date: September 4, 1997
     Amendment No.: 105
    Facility Operating License No. NPF-51: The amendment revised the 
Technical Specifications. Press release issued requesting comments as 
to proposed no significant hazards consideration: Yes. September 1, 
1997. Arizona Republic Newspaper (Arizona). Comments received: No. The 
Commission's related evaluation of the amendment, finding of exigent 
circumstances, consultation with the State of Arizona and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated September 4, 1997.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey 
Date of application for amendment: August 19, 1997, as supplemented 
August 20, 1997.

    Brief description of amendment: This amendment to the Technical 
Specifications increases the allowable band for control and shutdown 
rod demanded position versus indication position from plus or minus 12 
steps to plus or minus 18 steps when the power level is not greater 
than 85% rated thermal power.
    Date of issuance: September 10, 1997
    Effective date: As of date of issuance, to be implemented within 7 
days.
    Amendment No. 183
    Facility Operating License No. DPR-75: This amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes. The NRC published a public 
notice of the proposed amendment, issued a proposed finding of no 
significant hazards consideration, and requested that any comments on 
the proposed no significant hazards consideration be provided to the 
staff by the close of business on September 3, 1997, and stated that, 
should circumstances change during the notice period, such that a 
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 notice period, provided that its 
final determination is that the amendment involves no significant 
hazards consideration. The notice was published in the Wilmington News 
Journal on August 22, 1997, and in Today's Sunbeam on August 24, 1997. 
No public comments were received. The Commission's related evaluation 
of the amendment, finding of exigent circumstances, consultation with 
the State of New Jersey and final no significant hazards consideration 
determination are contained in a Safety Evaluation dated September 10, 
1997.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit - N21, P.O. Box 236, Hancocks Bridge, NJ 08038
    NRC Project Director: John F. Stolz
    Dated at Rockville, Maryland, this 17th day of September 1997.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 97-25210 Filed 9-23-97; 8:45 am]
BILLING CODE 7590-01-F