[Federal Register Volume 62, Number 232 (Wednesday, December 3, 1997)]
[Notices]
[Pages 63970-63986]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-31522]


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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 November 7, 1997, through November 20, 1997. 
The last biweekly notice was published on November 19, 1997 (62 FR 
61836).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and should cite the publication date and 
page number of this Federal Register notice. Written comments may also 
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By January 2, 1998, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the

[[Page 63971]]

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.

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

    Date of amendments request: November 6, 1997.
    Description of amendments request: The proposed amendments change 
the Technical Specifications (TS) for the Brunswick Steam Electric 
Plant (BSEP) Units 1 and 2 to allow three 18-month diesel generator 
(DG) surveillance requirements (SR) to be performed during both plant 
operation (Operational Conditions 1 and 2) and shutdown (Operational 
Conditions 3, 4, and 5) rather than, as currently required, only during 
shutdown. The first SR is an inspection of the DG involving a partial 
disassembly. The second ensures that non-critical DG protective 
functions are bypassed on an Emergency Core Cooling system actuation 
signal. The third verifies that the DG operates for greater than or 
equal to 60 minutes while loaded to at least 3500 kw, which bounds the 
maximum expected post-accident diesel generator loading. The proposed 
amendments additionally remove an expired footnote from the BSEP Unit 2 
DG TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    10 CFR 50.92 provides standards for determining whether a 
significant hazards consideration exists. A proposed amendment to an 
operating license for a facility involves no significant hazards 
consideration if 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, (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. Carolina Power & Light Company has reviewed 
these proposed license amendment requests and has concluded that their 
adoption would not involve a significant hazards consideration. The 
basis for this determination follows.
    1. The proposed license amendments do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    The proposed license amendments add a footnote to SR 4.8.1.1.2.d to 
allow performance of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and SR 
4.8.1.1.2.d.5 in OPERATIONAL CONDITION 1, 2, 3, 4, or 5 rather than 
only during shutdown. The footnote requires the unit to be in 
OPERATIONAL CONDITION 3, 4, or 5 when performing SR 4.8.1.1.2.d.2, SR 
4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR 4.8.1.1.2.d.7 for its 
associated diesel generators. No such limitation is placed on SR 
4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, or SR 4.8.1.1.2.d.5.
    There is no relaxation of any limiting condition for operation 
(LCO) and no decrease in surveillance requirements as a result of the 
proposed amendments. As such, the proposed license amendments will not 
affect the ability of the diesel generators to perform their intended 
safety function. Performance of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and

[[Page 63972]]

SR 4.8.1.1.2.d.5, during power operations, will not adversely affect 
overall nuclear safety. Diesel generator capacity is such that any 
three of the four diesel generators can supply the required loads for 
the safe shutdown of one unit and a design basis accident on the other 
unit without relying on offsite power. The diesel generator is not tied 
to the emergency bus (E bus) during performance of SR 4.8.1.1.2.d.1 or 
SR 4.8.1.1.2.d.4. Therefore, performance of SR 4.8.1.1.2.d.1 and SR 
4.8.1.1.2.d.4, during power operation, will not affect the operability 
of any other safety-related systems nor will it create any 
perturbations of the electrical distribution system that could 
challenge plant operation.
    Performance of SR 4.8.1.1.2.d.5, during power operation, will not 
adversely affect overall nuclear safety. SR 4.8.1.1.2.d.5 is performed 
in a similar manner to SR 4.8.1.1.2.a.5, which requires that, at least 
once per 31 days on a staggered test basis, a diesel generator be 
synchronized to the E bus and loaded to 1750 kw for 15 minutes. The 
critical portions of these surveillances are when the diesel generators 
are being synchronized to the E bus or disconnected from the E Bus. As 
such, performance of SR 4.8.1.1.2.d.5 during power operation does not 
create an additional opportunity of a perturbation of the electrical 
distribution system that could challenge plant operation than currently 
exists as a result of the performance of SR 4.8.1.1.2.a.5. The existing 
design of the electrical distribution system ensures that a grid 
problem will not result in failure of a diesel generator when it is 
synchronized to the E bus. The E buses are normally supplied by offsite 
power, via a 4160 V balance of plant (BOP) bus, through a master/slave 
breaker combination. When performing SR 4.8.1.1.2.d.5, the diesel 
generator is started in manual mode and synchronized to the E bus. With 
a diesel generator synchronized to the E bus, the diesel generator is 
protected from a potential overload condition. Class 1E protective 
relaying, at the E bus, is aligned to the trip circuit of the slave 
breaker to protect the diesel from an overload condition should the 
normal source of power be lost. These relays sense E bus voltage, E bus 
frequency, and directional power from the E bus to the BOP bus. 
Actuation of any of these relays, with the diesel in manual, will trip 
the slave and master breakers to separate the diesel generator from the 
BOP bus. This separates the diesel generator from the potential 
overload condition. In addition, either a loss of offsite power or loss 
of coolant accident results in the diesel generator output breaker 
opening, E bus loads stripping, and the diesel generator reverting to 
automatic mode. This allows the diesel generator to tie back to the E 
bus and carry the E bus loads.
    The proposed license amendments reflect the clarification, 
previously made to Bases Section 3/4.8, ``Electrical Power Sources,'' 
in SR 4.8.1.1.2.d itself. Accordingly, SR 4.8.1.1.2.d.2, SR 
4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and SR 4.8.1.1.2.d.7 are performed for 
diesel generator 1 or 2 with BSEP, Unit No. 1 in OPERATIONAL CONDITION 
3, 4, or 5 and for diesel generator 3 or 4 with BSEP, Unit No. 2 in 
OPERATIONAL CONDITION 3, 4, or 5. Defining the term ``during shutdown'' 
as ``OPERATIONAL CONDITION 3, 4, or 5'' is consistent with the current 
TS requirements of SR 4.8.1.1.2.d. TS Table 1.2, ``OPERATIONAL 
CONDITIONS,'' defines five OPERATIONAL CONDITIONS for the BSEP. There 
are two OPERATIONAL CONDITIONS applicable to power operation with the 
unit critical (i.e., POWER OPERATION and STARTUP) and three OPERATIONAL 
CONDITIONS applicable to a subcritical, shutdown unit (i.e., HOT 
SHUTDOWN, COLD SHUTDOWN, and REFUELING). Therefore, ``during shutdown'' 
and ``in OPERATIONAL CONDITION 3, 4, or 5'' have equivalent meaning.
    Eliminating the expired BSEP, Unit No. 2 footnote to SR 
4.8.1.1.2.d.1 is an administrative change and, therefore, cannot 
increase the probability or consequences of an accident previously 
evaluated.
    Based on the above, the proposed license amendments do not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed license amendments will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The proposed license amendments to allow performance of SR 
4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and SR 4.8.1.1.2.d.5 in OPERATIONAL 
CONDITION 1, 2, 3, 4, or 5, rather than only during shutdown, do not 
affect the operation or response of any plant equipment, including the 
diesel generators, or introduce any new failure mechanism. Plant 
systems and equipment will continue to respond in accordance with 
design and as analyzed. There will not be a malfunction of a new or 
different type introduced by the proposed license amendments.
    The proposed license amendments reflect the clarification, 
previously made to Bases Section 3/4.8, in SR 4.8.1.1.2.d itself. 
Accordingly, SR 4.8.1.1.2.d.2, SR 4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and 
SR 4.8.1.1.2.d.7 are performed for diesel generator 1 or 2 with BSEP, 
Unit No. 1 in OPERATIONAL CONDITION 3, 4, or 5 and for diesel generator 
3 or 4 with BSEP, Unit No. 2 in OPERATIONAL CONDITION 3, 4, or 5. 
Defining the term ``during shutdown'' as ``OPERATIONAL CONDITION 3, 4, 
or 5'' is consistent with the current TS requirements of SR 
4.8.1.1.2.d. TS Table 1.2, ``OPERATIONAL CONDITIONS,'' defines five 
OPERATIONAL CONDITIONS for the BSEP. There are two OPERATIONAL 
CONDITIONS applicable to power operation with the unit critical (i.e., 
POWER OPERATION and STARTUP) and three OPERATIONAL CONDITIONS 
applicable to a subcritical, shutdown unit (i.e., HOT SHUTDOWN, COLD 
SHUTDOWN, and REFUELING). Therefore, ``during shutdown'' and ``in 
OPERATIONAL CONDITION 3, 4, or 5'' have equivalent meaning.
    Eliminating the expired BSEP, Unit No. 2 footnote to SR 
4.8.1.1.2.d.1 is an administrative change and, therefore, cannot create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Based on the above, the proposed license amendments do not create 
the possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    Bases Section 3/4.8, ``Electrical Power Systems,'' states that the 
operability of the alternating current (ac) and direct current power 
sources and associated distribution systems during operation ensures 
that sufficient power will be available to supply the safety-related 
equipment required for the safe shutdown of the facility and the 
mitigation and control of accident conditions within the facility. 
Diesel generator capacity is such that any three of the four diesel 
generators can supply the required loads for the safe shutdown of one 
unit and a design basis accident on the other unit without relying on 
offsite power. Performance of SR 4.8.1.1.2.d.1, SR 4.8.1.1.2.d.4, and 
SR 4.8.1.1.2.d.5 during power operation will not affect the operability 
of any other safety-related systems, nor will it create any 
perturbations of the electrical distribution system that could 
challenge plant operation. Class 1E protective relaying, at the E bus, 
protects the diesel from an overload condition should the normal source 
of power be lost while performing SR 4.8.1.1.2.d.5. There is no 
relaxation of any LCO as a result of the

[[Page 63973]]

proposed license amendments. If an additional ac power source becomes 
inoperable during the performance of SR 4.8.1.1.2.d.1, SR 
4.8.1.1.2.d.4, and SR 4.8.1.1.2.d.5, the units will be placed in the 
appropriate OPERATIONAL CONDITION in accordance with TS 3.8.1.1, ``A.C. 
Sources Operating.'' Therefore, the diesel generators' ability to 
perform their intended safety function, as described in Section 
8.3.1.1.6.1 of the BSEP Updated Final Safety Analysis Report, is not 
adversely affected by the proposed license amendments.
    The proposed license amendments are consistent with the guidance of 
Generic Letter 91-04, ``Changes In Technical Specification Surveillance 
Intervals To Accommodate A 24-Month Fuel Cycle,'' which concludes that 
TSs need not restrict surveillances to only being performed during 
shutdown provided that performance of the surveillance during power 
operations does not adversely affect safety.
    The proposed license amendments reflect the clarification, 
previously made to Bases Section 3/4.8, in SR 4.8.1.1.2.d itself. 
Accordingly, SR 4.8.1.1.2.d.2, SR 4.8.1.1.2.d.3, SR 4.8.1.1.2.d.6, and 
SR 4.8.1.1.2.d.7 are performed for diesel generator 1 or 2 with BSEP, 
Unit No. 1 in OPERATIONAL CONDITION 3, 4, or 5 and for diesel generator 
3 or 4 with BSEP, Unit No. 2 in OPERATIONAL CONDITION 3, 4, or 5. 
Defining the term ``during shutdown'' as ``OPERATIONAL CONDITION 3, 4, 
or 5'' is consistent with the current TS requirements of SR 
4.8.1.1.2.d. TS Table 1.2, ``OPERATIONAL CONDITIONS,'' defines five 
OPERATIONAL CONDITIONS for the BSEP. There are two OPERATIONAL 
CONDITIONS applicable to power operation with the unit critical (i.e., 
POWER OPERATION and STARTUP) and three OPERATIONAL CONDITIONS 
applicable to a subcritical, shutdown unit (i.e., HOT SHUTDOWN, COLD 
SHUTDOWN, and REFUELING). Therefore, ``during shutdown'' and ``in 
OPERATIONAL CONDITION 3, 4, or 5'' have equivalent meaning.
    Eliminating the expired BSEP, Unit No. 2 footnote to SR 
4.8.1.1.2.d.1 is an administrative change and, therefore, cannot 
involve a significant reduction in a margin of safety.
    Based on the above, the proposed license amendments do not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: James E. Lyons.

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

    Date of amendments request: November 6, 1997.
    Description of amendments request: The proposed amendments to 
Technical Specification (TS) Limiting Conditions for Operation (LCO) 
3.3.5.5, Instrumentation for Control Room Emergency Ventilation System 
(CREVS) and 3.7.2, Control Room Emergency Ventilation System, and 
associated Bases for the Brunswick Steam Electric Plant (BSEP) Units 1 
and 2 would be limited in duration (approximately 3 months) and would 
allow operation of both BSEP units to continue while upgrades to the 
control building ventilation system, including new air conditioning 
(AC) units, are being installed. Part of the planned work requires 
opening the ductwork at the evaporative (i.e. cooling) coils. Temporary 
barriers will be constructed to preserve the leakage integrity of the 
control room pressure boundary; however, the temporary barriers will 
not be seismically qualified. While the permanent AC units are out of 
service, temporary AC units will be utilized. During the upgrade 
installation, the AC for the control room will not be protected from 
certain external events (e.g., seismic events, environmental hazards 
such as tornadoes and hurricanes, radiological sabotage, and missile 
hazards), as required by the system design and licensing basis, and 
will not fully meet single failure criteria.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect any component of any of the 
barriers to radiation release, any of the systems which protect the 
core from overheating, nor any system used to shut down the reactor. 
The proposed changes do not affect any of the chlorination system 
piping or the tank car, which would be the initiating components of a 
chlorine release event. The proposed changes affect the CREVS and CREVS 
instrumentation, neither of which are accident or event causing 
systems. Therefore, the proposed changes do not increase the 
probability of an accident or toxic gas release previously analyzed in 
the Updated Final Safety Analysis Report (FSAR).
    The proposed changes do not affect the ability of the CREVS to 
mitigate the consequences of a design basis accident or event involving 
a release of radioactive material. In addition, the proposed changes do 
not significantly affect the ability of the system to mitigate the 
consequences of a toxic gas release. The following measures will be 
taken to minimize the consequences of accidents and events:
    Temporary isolation barriers will be constructed to provide 
integrity of the duct during design basis radiation release events. 
These temporary barriers will ensure that 10 CFR Part 50, General 
Design Criterion 19 for Control Room operator doses is met for all 
design basis radiation release accidents.
    During the time that the temporary barrier is used, the chlorine 
tank car will be removed from the exclusion area. Analyses have shown 
that with the chlorine tank car outside of the exclusion area, there is 
no threat to Control Room habitability. Removal of the chlorine tank 
car from the exclusion area is the current Technical Specification 
requirement for inoperability of the Control Room chlorine isolation 
mode.
    The temporary condensing units for the Control Room Air 
Conditioning system will be installed to high quality standards, and a 
spare condensing unit will be provided such that two units can be 
maintained functional. These units will each be powered from a separate 
division of Class 1E power. The operation of the units will be 
monitored to ensure that they are in good operating order.
    If two or more of the condensing units should fail, instructions 
have been provided to the operators for increased monitoring of 
temperatures, and mitigating actions are available to the operators if 
temperatures rise above a predetermined limit.
    Therefore, the consequences of an accident or an event involving a 
release

[[Page 63974]]

of radiation, toxic gas, or smoke will not be significantly increased. 
In addition, the change will not significantly affect the consequences 
of a seismic event or other severe natural phenomena, as previously 
analyzed in the Updated FSAR.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes involve adjustments to the LCO requirements 
for CREVS relative to protection from severe natural phenomena. The 
proposed changes do not introduce any new modes of plant operation. The 
proposed changes do not involve any new modes of system operation, 
except that temporary condensing units will be used in place of the 
permanent condensing units. The temporary condensing units will 
interface with the permanent Control Building Heating Ventilation and 
Air Conditioning system in a similar manner to the permanent system. 
The piping connections to the permanent system will be the same, and 
the controls interface will be the same. No new cross-ties will be 
created and no new piping will be run though the habitability boundary. 
Therefore, this change will not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed changes do not represent a significant change in the 
assumptions and inputs to the analyses for Control Room operator doses. 
No increase in the doses to the Control Room operators is expected 
after a seismic event or tornado, since the integrity of existing 
barriers to release of radioactive material are not affected. 
Therefore, this change does not result in a significant reduction in 
the margin of safety for a radiological event.
    The proposed change does not represent a change to the leakage 
criteria for the Control Room, or the Control Room ventilation 
ductwork, following either a toxic gas or external smoke event. The 
bounding analysis remains valid, unless the failure is caused by a 
tornado or seismic event. Due to the low probability of such an event 
occurring during the short time frame involved in this modification, 
the occurrence of such an event is not of significant concern.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: James E. Lyons.

Consumers Energy Company, Docket No. 50-155, Big Rock Point Nuclear 
Plant, Charlevoix County, Michigan

    Date of amendment request: September 19, 1997 (Accession No. 
9709240373).
    Description of amendment request: The amendment request propose 
changes to the Facility Operating License and technical specifications 
(TS) to reflect the permanent cessation of power operations and 
permanent transfer of nuclear reactor fuel to the spent fuel pool 
(SFP). In particular, Consumers Energy requests to change: safety 
limits; limiting safety system settings; limiting control system 
settings; limiting conditions for operation; surveillance requirements; 
design features; and administrative controls. On November 12, 1997, 
Consumers Energy provided supplemental information regarding their no 
significant hazards determination, as requested by NRC request for 
additional information letter dated October 12, 1997. By letters dated 
June 26 and September 23, 1997, the licensee certified permanent 
cessation of power operations and permanent removal of all fuel from 
the reactor, respectively.
    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:
    The proposed change provides the applicable requirements to assure 
safe storage of spent nuclear fuel during decommissioning following the 
permanent cessation of power operations at the Big Rock Point Nuclear 
Plant (BRP) on August 30, 1997 [see Consumers Energy letter to NRC 
dated June 26, 1997] and permanent removal of all fuel from the reactor 
vessel on September 20, 1997 [see Consumer Energy letter to NRC dated 
September 23, 1997]. Decommissioning activities conducted using these 
controls do not present undue risk to the public, and do not impact 
common defense and security. As such, these changes will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    No accidents previously evaluated in the Updated Final Hazards 
Summary Report (UFHSR) will have their probability of occurrence 
increased because the proposed controls effectively preclude the 
occurrence of criticality, fuel temperature exceeding limits, or fuel 
handling accidents. The probability of plant accidents associated with 
power operations have been significantly reduced. Accidents associated 
with spent fuel handling, including cask and single bundle drop and 
spent fuel cooling capability loss events, are still pertinent and were 
reviewed using new data on pool inventory and revised 10 CFR 20 
radiological limit determinations. The probability of occurrence of 
accidents associated with storing 441 spent fuel assemblies in the SFP 
(current license limit) have not been affected by the changes in the 
proposed TSs.
    The consequences of a fuel handling and cask drop accidents were 
evaluated based on the removal of all fuel from the reactor and loading 
spent nuclear fuel in the SFP. The removal of all fuel from the reactor 
vessel to storage in the SFP and the subsequent decay of the fuel in 
the pool result in no increase in the probability of these accidents 
and continuously reduced consequences from these accidents.
    Analyses using the techniques in Branch Technical Position APCSB 9-
2 provide the heat rate from a freshly-removed full core off-load in 
the SFP whose racks are filled with a total of 441 fuel assemblies as 
the most limiting cooling condition. Existing cooling equipment under 
the current TSs provide sufficient cooling to preclude spent fuel pool 
temperatures reaching 150 degrees-Fahrenheit with a complete loss of 
spent fuel cooling for 72 hours. This precludes entry into an 
unanalyzed condition for the SFP and provides 3 days to recover cooling 
flow of ``approximately 30'' gallons per minute. Since this 
specification change is intended for implementation following 93 days 
after shutdown (approximately November 30, 1997), this analysis 
justifies the allowance of 24 hours to re-establish cooling flow 
provided in specification 3.1.2.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The permanent cessation of power operation and removal of fuel from 
the

[[Page 63975]]

reactor eliminates the possibility of the following categories of 
accidents and transients to create a hazard to the health and safety of 
the public: increase in heat removal by the secondary system; increase 
in reactor coolant inventory; decrease in heat removal by the secondary 
system; decrease in reactor coolant inventory; reactivity and power 
distribution anomalies; anticipated transient without scram; and, 
single loop operation. These revised TSs, in combination with 
requirements in the UFHSR, provide assurance that fuel handling and 
spent fuel cask drop accident, which represent the remaining specific 
pertinent accidents analyzed in the ``radioactive release from a 
subsystem of component'' category, will not occur. Because the revised 
TSs related to fuel handling, spent nuclear fuel storage, and handling 
of the spent fuel cask satisfy current license and UFHSR requirements, 
no new accidents are created.
    3. Involve a significant reduction in a margin of safety.
    The safety margins for analyzed accidents are maintained because 
the containment structures and redundant control established by the 
plant remain in place until the decay of spent fuel has reduced the 
source term to levels that analysis confirms do not require the 
containment features. ninety three days after permanent cessation of 
operations, the spent nuclear fuel at BRP will have decayed to the 
point where the added margin from this decay more than compensates for 
the removal of the containment as a safety feature, and allows relaxed 
controls for the cooling of the SFP.
    The Big Rock Point Plant Safety Committee has reviewed this 
Facility Operating License and TS change request and has determined 
this change does not involve an unreviewed safety question and, 
therefore, involves no significant hazards consideration. The proposed 
change has been reviewed by the BRP Nuclear Performance Assessment 
Department.
    The NRC staff has reviewed the licensee's analysis, as provided by 
licensee letters dated September 19 and November 12, 1997, and, based 
on this review, it appears that the three standards or 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: North Central Michigan College, 1515 
Howard Street, Petosky, MI 49770.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Seymour H. Weiss.

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: March 11, 1993; supplemented August 26, 
November 29, December 6, 1993, October 3, 1995, February 27, and 
September 3, 1997 (TSC 93-03).
    Description of amendment request: The proposed changes would 
replace the present Electrical Power Systems section of the Technical 
Specifications, Sections 3.7 and 4.6, by consolidating and rearranging 
the present specifications, incorporating new specifications, and 
formating the section similar to the Babcock and Wilcox Standard 
Technical Specifications. The proposed changes would address such 
concerns as Keowee hydro station operability, Lee gas turbine 
operability, overhead and underground emergency power path operability, 
Keowee and Keowee main step-up transformer outage requirements, 
surveillance requirements of various components and systems, Oconee 
distribution system requirements, protective instrumentation system 
requirements, operability of 125 VDC Vital Instrument and Control power 
and limiting condition for operation, inverter requirements, Oconee 
shutdown requirements related to various components, Keowee unit 
extended outage, dc power operability requirements, battery cell 
parameter requirements, and various editorial and related Bases 
changes.
    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.
    Duke Power Company (Duke) [currently Duke Energy Corporation] has 
made the determination that this amendment request involves a No 
Significant Hazards Consideration by applying the standards established 
by NRC regulations in 10 CFR 50.92. This ensures 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:
    Each accident analysis addressed within the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to the changes 
proposed within this amendment request. Changes included in this 
amendment request are provided to assure availability of electrical 
power systems for mitigation of Design Basis Accidents (DBAs). As 
described within the technical justification, the following types of 
changes are included:
    (1) Editorial and administrative changes associated with 
reformatting the Technical Specification requirements;
    (2) Additional restrictions not presently included in the Technical 
Specifications such as the addition of requirements for electrical 
power systems during cold shutdown and refueling, for the 230 kV 
switchyard degraded grid protection system and to delete the special 
inoperability period for the Keowee CX transformer;
    (3) Technical changes to current requirements to provide clarity 
and operational flexibility. These changes maintain the ability of the 
electrical power systems to mitigate the consequences of DBAs without a 
significant reduction in availability. These changes include the 
definition of emergency power paths to include the associated DC 
sources and auxiliary transformers, the combination of special 
inoperability periods for ``planned'' and ``unplanned'' reasons, and 
the ability to use the Keowee special inoperability period more than 
once in a three year period; and
    (4) Relocation of requirements which are unnecessary for the 
mitigation of DBAs to licensee controlled documents. Relocated 
requirements include surveillance requirements for the External Grid 
Trouble Protection system.
    Based on the above and the technical justification * * *, there is 
no significant increase in the probability of DBA as a result of this 
change, nor is there a significant increase in the consequences of a 
DBA as a result of this change since the proposed amendment assures 
availability of electrical power systems.
    (2) Create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated:
    The proposed changes make no physical changes to the plant 
configuration and do not adversely affect the performance of any 
equipment. Operation of ONS [Oconee Nuclear Station] in accordance with 
these Technical Specifications will not create any failure modes not 
bounded by previously evaluated accidents. Consequently, this change 
will not create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated.

[[Page 63976]]

    (3) Involve a significant reduction in a margin of safety:
    Margins of safety associated with these Technical Specifications 
have been evaluated. These changes include editorial and administrative 
changes associated with reformatting the Technical Specification 
requirements, additional restrictions not presently included in the 
Technical Specifications, technical changes to current requirements 
which maintain the ability of the electrical power systems to mitigate 
the consequences of DBAs, and relocation of requirements which are 
unnecessary for the mitigation of DBAs to licensee controlled 
documents. The design basis of auxiliary electrical systems is to 
supply the required ES [emergency system] loads of one Unit and safe 
shutdown loads of the other two units. The proposed amendment does not 
affect any safety limits, setpoints, or design parameters and assures 
the continued availability of electrical power systems; thus preserving 
the existing margin of safety. Therefore, there will be no significant 
reduction in any margin of safety.
    Duke has concluded based on the above, and the technical 
justification * * * 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.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC.
    NRC Project Director: Herbert N. Berkow.

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

    Date of amendment request: November 4, 1997.
    Description of amendment request: The proposed amendment would 
change Item 6.a.2, ``4.16 Emergency Bus (Start Diesel),'' of Table 3.3-
4 of Technical Specification 3.3.2.1. The proposed change would reduce 
the trip setpoint for starting the emergency diesel generators on 
emergency bus undervoltage from a trip setpoint of greater than or 
equal to 83 percent with a 12-cycle delay time to greater than or equal 
to 75 percent of nominal bus voltage with a time delay of less than 0.9 
second including auxiliary relay times. The proposed change would also 
reduce the allowable value from greater than or equal to 81 percent of 
nominal bus voltage to greater than or equal to 74 percent of nominal 
bus voltage with a time delay of less than 0.9 second including 
auxiliary relay times.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change replaces the current Engineered Safety Feature 
setpoint, allowable value and delay time for the diesel generator start 
on loss of power function. An analysis has been performed to develop 
the new values to minimize the diesel generator starts when a Reactor 
Coolant Pump (RCP) is being started or a fast bus transfer occurs. The 
heat generated by an increase in motor current, in response to reduced 
voltage, will be less than the heat generated during motor starting. 
The analysis results show that bus voltages may dip below the allowable 
setpoint value and then recover to the pick-up setpoint within the 
proposed delay time without stalling motors.
    The proposed change does not affect the design and reliability of 
any plant equipment; therefore, the probability of occurrence of a 
previously evaluated accident is not increased. The operation of the 
plant will not be changed as a result of this proposed amendment, 
except that fewer diesel generator starts will be initiated.
    This function anticipates the loss of voltage to protect equipment 
connected to the 4.16 Kv emergency bus. The UFSAR [Updated Final Safety 
Analysis Report] accident analyses do not take credit for this 
function; therefore, the consequences of an accident previously 
evaluated is not increased.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change to the trip setpoint, allowable value and delay 
time will continue to ensure that the safety-related equipment 
connected to the emergency bus is adequately protected from a low 
voltage condition. These setting changes will minimize the diesel 
generator starts due to voltage drops when an RCP is started or a fast 
bus transfer occurs.
    The new setpoint and time delay allow normal voltage drops to occur 
during expected plant operations without causing any thermal damage to 
safety-related equipment. The performance of the safety system will 
remain unchanged and will not alter any plant equipment, performance 
requirements or safety analysis. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin of 
safety?
    The proposed change does not involve a significant reduction in a 
margin of safety since an analysis has been performed to verify that 
safety-related equipment connected to the emergency bus is adequately 
protected from a low voltage condition with the proposed settings. The 
proposed changes do not affect the UFSAR design bases, accident 
assumptions, or technical specification bases. In addition, the 
proposed changes do not affect release limits, monitoring equipment or 
plant operating practices. Therefore, the proposed change will not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.
    Attorney for Licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: October 28, 1997.
    Description of amendment request: The amendment would (1) revise 
the frequency of conducting five Surveillance Requirements (SRs) and 
(2) add a 10 CFR Part 50, Appendix J Testing Program for Primary 
Containment Systems in the Technical Specifications (TSs) for Grand 
Gulf Nuclear Station, Unit 1 (GGNS). The five SRs are the following: SR 
3.6.1.1.1 for primary containment, SR 3.6.1.2.1

[[Page 63977]]

for primary containment air locks, and SRs 3.6.1.3.5, 3.6.1.3.8, and 
3.6.1.3.9 for primary containment isolation valves. The proposed 
revisions for each of the five SRs are to delete the references to SR 
3.0.2 not being applicable and change the surveillance frequency from 
being ``in accordance with 10 CFR 50, Appendix J, as modified by 
approved exemptions'' to ``in accordance with 10 CFR Part 50, Appendix 
J, testing program.'' The testing program would be added to Section 
5.0, Administrative Controls, of the TSs. Changes to the Bases of the 
TSs were also provided in the submittal.
    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:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.

[On April 26, 1995, the licensee was granted an exemption to Appendix J 
of 10 CFR Part 50 that allowed performance-based containment leak rate 
testing. This exemption will expire on the startup from Refueling 
Outage 9, currently scheduled for the spring of 1998. The licensee's 
proposed changes to the TSs are to adopt Option B, Performance-Based 
Requirements, that is now in Appendix J, but was not in Appendix J in 
1995 when the exemption was granted. The technical findings that 
support the rulemaking for Option B are in NUREG-1493, ``Performance-
Based Containment Leak Rate Test Program,'' dated September 1995. The 
licensee stated in its submittal that its current containment leak rate 
testing program meets the requirements of Option B.]
    Two initiating events were identified which could be affected by 
the proposed changes [in the submittal of October 28, 1997]. An 
interfacing system LOCA [(loss-of-coolant accident)] could be caused by 
significant leakage of both normally closed isolation valves in systems 
with high pressure/low pressure interfaces. Interfacing systems LOCAs 
were considered for the LPCI, LPCS, HPCS, and RCIC systems [(i.e., low 
pressure coolant injection, low pressure core spray, high pressure core 
spray, and reactor core isolation cooling)]. Because the frequency for 
testing of these valves will not be changed under this proposal, there 
is no increase in the probability or consequences of an accident 
[previously evaluated].
    The second event evaluated was a LOCA outside containment. In this 
case the probability for failure of the MSIVs [(main steam isolation 
valves)] and the feedwater isolation valves were calculated and 
combined with the frequency of a pipe break outside containment and the 
conditional probability of a core melt given a LOCA. The increase in 
core damage is extremely small and therefore does not significantly 
increase the probability of any previously evaluated accident. Further, 
because the testing frequency for MSIVs and feedwater isolation valves 
are not being changed, the LOCA outside containment events can be 
discounted.
    Failure of, or leakage through[,] a containment barrier can[,] 
however, increase the consequences of those accidents previously 
evaluated. Because the leakage probability for two valves in series to 
fail is very small and because all lines isolated by a single 
containment isolation valve always have a water seal and cannot act as 
a release pathway unless the integrity of the connected system is 
compromised, there is no significant increase in the consequences of 
any previously evaluated accident.
    Containment bypass can also increase the consequences of 
[previously] evaluated accidents. Accident sequences involving 
containment have been shown to be relatively insignificant by the GGNS 
IPE [(Individual Plant Examination]). The potential for [containment] 
bypass was analyzed. The analysis showed that the probabilities for 
bypass were dominated by failure to close scenarios. Many programs are 
in place at GGNS to monitor containment component performance[,] and to 
ensure that proper maintenance and repairs are made during the service 
life of the containment. Other routine surveillances are performed 
periodically to ensure that the valves will close on demand. In fact, 
all valves that are required to close for containment isolation and 
that are not maintained closed at all times during power operations are 
stroke tested quarterly or[,] at a minimum, during each refueling 
outage in accordance with ASME [(American Society of Mechanical 
Engineers) Boiler and Pressure Vessel Code,] Section XI, Subsection 
IWV.

[Based on the above, the proposed changes do not significantly increase 
the probability or consequences of an accident previously evaluated.]
    II. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The request involves the reduction in the local leak rate and the 
integrated leak rate testing frequencies [in accordance with Option B 
of Appendix J to 10 CFR Part 50]. Extending the test frequencies has no 
influence on, nor does it contribute in any way to, the possibility of 
a new or different kind of accident or malfunction from those 
previously analyzed. The method of performing the test is not changed. 
No new accident modes are created by extending the testing intervals. 
No safety-related equipment or safety functions are altered as a result 
of this change.

[Therefore, the proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated.]
    III. The proposed change does not involve a significant reduction 
in a margin of safety.
    The only margin of safety that has the potential of being impacted 
by the proposed changes involves the offsite dose consequences of 
postulated accidents which are directly related to containment leakage 
rate. The containment isolation system is designed to limit leakage to 
La which is defined by the GGNS TSs to be 0.437 percent by weight of 
the containment air [volume] per 24 hours at [the containment pressure 
of] 11.5 psig (Pa). The limitation on containment leakage rate is 
designed to ensure that total leakage volume will not exceed the value 
assumed in the accident analyses at the peak accident pressure (11.5 
psig, Pa).
    To provide additional conservatism, the measured overall integrated 
leakage rate is further limited to less than or equal to 0.75 La during 
performance of the periodic integrated leakage rate test and to less 
than or equal to 0.60 La for type B and C leakage rate tests [of 
Appendix J]. This is done to account for the possible degradation of 
the containment leakage barriers between [the Appendix J] tests. This 
acceptance criteria ensures that an acceptable margin of safety is 
being maintained and will not be altered by the proposed changes. The 
preservation of this margin will continue to provide for potential 
degradation of the leakage barriers between tests.
    No change in the method of testing is being proposed. The tests 
will continue to be done at full pressure (Pa) or greater [pressure]. 
The test pressure for primary containment isolation valves will 
continue to be applied in the same direction as would be required for 
the valve to perform its safety function (unless a different direction 
can be shown to be equivalent or conservative). Primary containment 
penetrations

[[Page 63978]]

which require Type B leakage rate tests will be performed in the same 
manner as before. The Type A test [of Appendix J] will continue to be 
performed at full pressure (Pa). Other programs are in place to ensure 
that proper maintenance and repairs are performed during the service 
life of the primary containment[,] and systems and components 
penetrating the primary containment.
    No change in the owners allowable leakage rate is being proposed. 
These conservative leakage rates ensure that[,] if every penetration 
were at its maximum allowable leakage rate, the total containment 
leakage would still be below 0.60 La. The effect of multiple 
penetration barriers is not considered which provides further 
conservatism.
    The assessment of risk analysis for the proposed changes concluded 
that the overall risk impact of the changes are neutral and essentially 
negligible. Any containment isolation barrier allowed to be tested at 
less frequent intervals [through performance-based testing of proposed 
Option B of Appendix J] will have demonstrated enhanced performance 
which minimizes the potential for increased leakage. The assessment 
further shows that there is reasonable assurance that an acceptable 
level of performance for the containment isolation function can be 
maintained. The overall risk impact for the proposed changes are small 
enough to be almost indeterminate. No change to the leakage rate 
specified in the TSs is being proposed.

[The proposed changes to the TSs are in accordance with Option B of 
Appendix J of 10 CFR Part 50.]

[Therefore, the proposed changes do not involve a significant reduction 
in a margin of safety.]
    Based on the above evaluation, operation in accordance with the 
proposed amendment involves no significant hazards consideration.
    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, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Project Director: David A. Wigginton, Acting.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee, 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: September 30, 1997.
    Description of amendment request: The proposed amendment would 
eliminate certain license conditions of the Maine Yankee operating 
license that are no longer appropriate in the permanently defueled 
condition of the plant. These conditions include restrictions on the 
Fire Protection Program and implementation of leakage reduction, 
airborne iodine monitoring, secondary water chemistry, and cooling 
water discharge monitoring programs. By letter dated August 7, 1997, 
the licensee certified permanent cessation of power operations and 
permanent removal of fuel from the reactor vessel. Most of the 
provisions of the Maine Yankee operating license were established to 
ensure protection of the public health and safety during power 
operations. Maine Yankee has proposed to eliminate those license 
requirements that are not relevant to the permanently defueled plant 
condition to allow the Maine Yankee staff to focus on those provisions 
which are still appropriate during decommissioning.
    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:
    The proposed change does not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    The purpose of the proposed change is to eliminate requirements 
which are not appropriate in the permanently defueled plant condition. 
Since the plant has permanently ceased operation and will be maintained 
in a defueled condition, many provisions of the license related to 
operation of the plant are no longer appropriate. Elimination of these 
unnecessary requirements allows the plant staff to focus on those 
requirements which continue to be appropriate to the existing plant 
condition. The proposed change does not affect those Chapter 14 
accidents which are appropriate to the current plant conditions: fuel 
handling accident, spent fuel cask drop, and radioactive liquid waste 
system leaks and failures, and therefore, does not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The purpose of this proposed change is to eliminate requirements 
which are not appropriate in the permanently defueled plant condition. 
Since the plant has permanently ceased operation and will be maintained 
in a defueled condition, many provisions of the license related to 
operation of the plant are no longer appropriate. Elimination of these 
unnecessary requirements allows the plant staff to focus on those 
requirements which continue to be appropriate to the existing plant 
conditions. This proposed change does not affect storage of spent fuel 
and, therefore, does not create the possibility of a new or different 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The purpose of the proposed change is to eliminate requirements 
which are not appropriate in the permanently defueled plant condition. 
Since the plant has permanently ceased operation and will be maintained 
in a defueled condition, many provisions of the license related to 
operation of the plant are no longer appropriate. Elimination of these 
unnecessary requirements allows the plant staff to focus on those 
requirements which continue to be appropriate to the existing plant 
conditions. This proposed change does not affect storage of spent fuel 
and, therefore, does not involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, P.O. Box 408, Wiscasset, ME 04578.
    NRC Project Director: Seymour H. Weiss.

Maine Yankee Atomic Power Company, Docket No. 50-309, Maine Yankee 
Atomic Power Station, Lincoln County, Maine

    Date of amendment request: October 20, 1997.
    Description of amendment request: The proposed amendment would 
replace in their entirety the existing Technical Specifications 
incorporated

[[Page 63979]]

in Facility Operating License No. DPR-36 as Appendix A. Maine Yankee 
developed the revised Technical Specifications, titled Permanently 
Defueled Technical Specifications, to reflect the permanently shutdown 
and defueled status of the plant. Changes are proposed to the 
definitions, limiting conditions for operation, surveillance, and 
administrative control sections.
    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. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). A summary of the licensee's 
review is presented below:
    The proposed change does not,
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    This proposed change is consistent with the improved Standard 
Technical Specifications. The relocation of requirements from the 
technical specifications to the licensee controlled documents is 
consistent with the criteria set forth in 10 CFR 50.36 for the content 
of technical specifications. The removal of definitions, generic LCO 
actions and generic surveillance requirements has no impact on facility 
structures or equipment or the methods of operation of such structures 
or equipment. The deletion of design features and safety limits not 
applicable to the permanently shutdown and defueled status of the Maine 
Yankee reactor has no impact on the remaining applicable design basis 
accidents. The removal of LCO and Surveillance specifications which are 
related only to the operation of the nuclear reactor or only to the 
prevention, diagnosis or mitigation of transients or accidents 
primarily involving the reactor, do not affect the remaining applicable 
accidents previously evaluated. The critical safety functions involving 
core reactivity control, reactor heat removal, reactor coolant system 
inventory control and containment integrity are no longer necessary at 
the Maine Yankee facility. The postulated accidents involving damage to 
the reactor coolant system, main steam lines, main feed lines, steam 
generators or the reactor core and the subsequent release of 
radioactive material are no longer applicable at the Maine Yankee 
facility. Spent fuel pool cooling and makeup related equipment and 
support equipment including electrical power systems are not required 
to be continuously available since there is time available to effect 
repairs or establish alternate sources of makeup flow in the event of a 
loss of cooling and makeup flow to the spent fuel pool. The effect of 
radioactive decay since the shutdown of the reactor has reduced the 
consequences of the fuel handling accident to levels below those 
previously analyzed. The relevant parameters associated with spent fuel 
pool (level and boron concentration) that make up the initial 
conditions assumed in applicable analysis are included in the technical 
specifications. The deletion and modification of provisions of 
administrative controls do not directly affect the design of structures 
or equipment necessary for the safe storage of irradiated fuel or the 
methods used for handling and storage of such fuel in the spent fuel 
pool. The changes to the administrative controls are, in fact, 
administrative in nature and do not affect any accident applicable to 
the safe storage of irradiated fuel or the permanently shutdown and 
defueled condition of the reactor. Therefore, the proposed changes to 
the Maine Yankee Technical Specifications do not involve any increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes have no impact on facility structures or 
equipment affecting the safe storage of irradiated fuel or the methods 
of operation of such structures or equipment or handling and storage of 
such fuel. These changes are consistent with the improved Standard 
Technical Specifications and add to the clarity and ease of use of the 
proposed PDTS. The removal of technical specifications which are 
related only to the operation of the nuclear reactor or only to the 
prevention, diagnosis or mitigation of transients or accidents 
primarily involving the reactor, can not result in different or more 
adverse failure modes or accidents than previously evaluated because 
the reactor is permanently shutdown and defueled. The proposed deletion 
of provisions of the Maine Yankee Technical Specifications do not 
affect systems credited in the existing accident analyses for the 
remaining applicable postulated accidents at the Maine Yankee facility. 
The proposed technical specifications continue to require proper 
control and monitoring of safety significant parameters and activities. 
The proposed restrictions on boron concentration and level in the spent 
fuel pool are fulfilled by normal operating conditions and preserve 
initial conditions assumed in the analyses of postulated DBA's. 
Therefore, the proposed changes to the MYTS does not create the 
possibility of a new or different accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The deletion of provisions in the technical specifications which 
are not related to the storage of irradiated fuel or which are 
inconsistent with the scope of the improved Standard Technical 
Specifications will not affect the analyses of the design basis 
accidents remaining applicable to the Maine Yankee facility. The 
postulated design basis accidents involving the reactor are no longer 
possible due to the permanently defueled status of the Maine Yankee 
reactor. The requirements for systems, structures and components which 
have been deleted from the Maine Yankee Technical Specifications are 
not credited in the existing accident analysis for the remaining 
applicable postulated accidents and therefore do not contribute to the 
margin of safety associated with the accident analysis. Therefore, the 
proposed changes to the Maine Yankee Technical Specifications would 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: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.
    Attorney for licensee: Mary Ann Lynch, Esquire, Maine Yankee Atomic 
Power Company, P.O. Box 408, Wiscasset, ME 04578.
    NRC Project Director: Seymour H. Weiss.

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

    Date of amendment request: October 15, 1997.
    Description of amendment request: The proposed change to Technical 
Specification 3/4.4.3, Pressurizer, would replace the pressurizer 
maximum water inventory requirement with a pressurizer maximum 
indicated level requirement. The proposed amendment would also modify 
the associated Bases section and make editorial changes.
    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

[[Page 63980]]

consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 10 CFR 
50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10 CFR 50.92(c) are not satisfied. The 
proposed revision does not involve [an] SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    The Technical Specification maximum pressurizer inventory 
requirement in Technical Specification 3.4.3 is being changed to use 
the numerical value for the Reactor Trip setpoint on pressurizer high 
water level in Technical Specification Section 2.2. This changes the 
requirement from a volume to a level requirement, is consistent with 
the Improved Standard Technical Specifications for Westinghouse plants, 
and represents a more restrictive level requirement than the current 
technical specification. The bases change clarifies that the 89% level 
requirement only assures that there is a steam bubble in the 
pressurizer. Also, the bases change states that pressurizer level is 
maintained by automatic and procedural controls to provide assurance 
that the design basis analyses are valid. These changes do not modify 
plant operation. Lowering the maximum level requirement so that it is 
numerically consistent with the reactor trip setpoint, while clarifying 
the bases of the requirement, [cannot] involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    There are no hardware modifications associated with the change. The 
change does not modify the way that the plant is operated. The change 
modifies neither accident mitigation nor system response post-accident.
    Therefore, the proposed revision does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The change places a lower maximum pressurizer level requirement for 
the pressurizer. The change imposes the numerical setpoint value for 
the reactor trip on pressurizer high water level as the restriction on 
the pressurizer level. The change to the bases clarifies that the 89% 
level requirement only ensures the existence of a steam bubble and not 
the validity of the design basis analyses. The design basis non-LOCA 
[loss-of-coolant accident] analyses use the current programmed 
pressurizer level and the LOCA analysis uses 62% level for full power. 
Those events that are analyzed to address pressurizer filling concerns 
are initiated assuming a higher initial pressurizer water level that 
accounts for 6% level uncertainty. The bases change makes it clear that 
the pressurizer level required to assure the validity of the design 
basis analyses is maintained by the automatic and procedural controls 
and not the less than or equal to 89% level in the requirement.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, bases on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: November 11, 1997.
    Description of amendment request: The proposed amendment to 
Technical Specifications (TS) 3.9.1.2 and 3.9.13 and their Bases will 
allow crediting soluble boron for maintaining k-effective at less than 
or equal to 0.95 within the spent fuel pool (SFP) rack matrix following 
a seismic event of a magnitude greater than or equal to an operating 
basis earthquake (OBE).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed the proposed revision in accordance with 
10CFR50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The basis for this conclusion 
is that the three criteria of 10 CFR 50.92(c) are not satisfied. The 
proposed revision does not involve [an] SHC because the revision would 
not:
    1. Involve a significant increase in the probability or consequence 
of an accident previously evaluated.
    There is one spent fuel pool accident condition discussed in 
Chapter 15 of the FSAR [final safety analysis report]. The FSAR 
discusses a fuel handling accident which drops a fuel assembly onto the 
fuel racks during fuel movement. Degradation of the Boraflex panels in 
a post-seismic condition will have no effect on the probability of a 
fuel assembly drop onto the stored fuel, or the fuel racks. Changing 
the way Boraflex responds to a seismic event will have no impact on the 
probability of a seismic event. A misplaced fuel assembly can be 
postulated in the MP3 [Millstone Unit 3] fuel pool as a result of 
either equipment malfunction or operator error. Degradation of the 
Boraflex panels will have no effect on the probability of a fuel 
misplacement event. Therefore, the degradation of Boraflex in a post-
seismic condition does not involve an increase in the probability of an 
accident previously evaluated.
    A fuel handling accident could cause a radioactive release of 
fission gases, resulting in dose consequences. This radioactive release 
of fission gases is due to the failure of a certain number of fuel pins 
which are postulated to fail during the fuel handling accident. The 
number of fuel pins which are postulated to fail in this event is not 
changed by the degradation of the Boraflex panels in a post-seismic 
condition. There are no criticality issues with this fuel handling 
accident for the reason described next. Although conservative, should a 
fuel handling accident occur during or after a seismic event, even with 
no Boraflex credit, the proposed 1750 ppm [parts per million] of 
soluble boron is sufficient to ensure that K-effective of the SFP is 
maintained at less than or equal to 0.95. The 1750 ppm boron 
requirement also bounds any criticality concerns for a fuel handling or 
dropped load event due to the no Boraflex assumption. Therefore, this 
proposed change does not involve an increase in the probability or

[[Page 63981]]

consequences of an accident previously evaluated.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The change in the way Boraflex responds to a seismic event with the 
presence of 1750 ppm boron does not create a new accident. The use of 
soluble boron in the spent fuel pool is safe. There is no possibility 
of a dilution event during or following a seismic event up to the 
magnitude of an SSE [safe shutdown earthquake]. The normally filled 
piping systems in the vicinity of the spent fuel pool are fire 
protection, hot water heating, hot water preheating, domestic water, 
and component cooling. In addition, the roof drain system piping runs 
through the building. An engineering review of these systems has 
determined that the majority of the systems are leak tight and meet 
NU's [Northeast Utilities'] commitment to seismic II/I criteria for a 
seismic event up to and including an SSE. The analysis was performed 
consistent with the original design criteria for seismic II/I piping as 
documented in section 3.9.2 of the Millstone 3 Safety Evaluation Report 
(SER) Number 4.
    Portions of fuel building piping systems that may not be leak tight 
following an SSE, and that would not leak into the spent fuel pool 
based on location of the potential leak, are not possible sources of 
dilution.
    Two lines in the Hot Water Preheating system will be modified to 
meet the leak tight seismic II/I criteria and will not be possible 
sources of dilution.
    A new pipe support will be added to the roof drain piping to meet 
the seismic II/I criteria. With the new support installed, one portion 
of the drain piping will still not meet leak tight requirements. The 
inlet opening on the roof feeding this portion of the piping will 
therefore be capped. Since the location of the potential cracking in 
the drain piping lies above the connection to the balance of the drain 
piping, and the system is not under pressure, water flowing from other 
portions of the drain system will not flow up to and out of the 
potentially cracked portion. This precludes a possible source of 
dilution.
    Non borated water sources that are connected to the SFP will be 
isolated following a seismic event of greater than or equal to an OBE 
to prevent dilution. Therefore there is no possibility of a SFP boron 
dilution accident coincident with or following a seismic event up to an 
SSE, and credit for soluble boron is acceptable to meet the K-effective 
limit of 0.95 for the SFP. The crediting of soluble boron in the spent 
fuel pool to control K-effective following a seismic event does not 
create a new accident as boron dilution of the pool can be prevented by 
closing and administratively controlling the opening of dilution paths 
to the pool and initiating routine sampling requirements on SFP boron. 
At present the crediting of soluble boron following a fuel misplacement 
event is allowed for [in] the Millstone 3 [TS]. Analysis has shown that 
a seismic event of greater than an OBE level earthquake can cause 
Boraflex damage which can be more limiting than a fuel misplacement 
event. As such, the minimum boron requirement in the fuel pool will be 
increased from 800 ppm to 1750 ppm. As such, no new accident has been 
created because the crediting of boron following a malfunction/accident 
has always been allowed.
    Therefore, the proposed revision does not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety, as defined by MP3 Technical Specifications, 
is to ensure that the K-effective of the MP3 SFP is maintained less 
than or equal to 0.95 at all times. The proposed change does not credit 
soluble boron during normal operations, but allows crediting soluble 
boron at a new higher concentration for control of K-effective during 
malfunction conditions. There is no reduction in the margin of safety 
as the result of the degradation of Boraflex following a greater than 
OBE seismic event, because soluble boron will compensate for the loss 
of Boraflex. A value of 1750 ppm of soluble boron in the SFP at all 
times ensures that K-effective of the MP3 SFP is maintained less than 
or equal to 0.95 at all times, including this new malfunction of 
degraded Boraflex following a greater than OBE seismic event.
    Eliminating the credit for the reactivity [hold-down] effect of 
Boraflex panels in conjunction with 1750 ppm boron will have no effect 
on the probability of a seismic event. As the probability of a seismic 
event has not changed there is no increase in the probability of an 
accident or malfunction due to a seismic event. Following a seismic 
event, operators are presently required to make inspections of the 
plant to determine post seismic event plant conditions. As a result of 
this change, inspections will be required to review the status of the 
spent fuel pool and isolate potential dilution paths following a 
seismic event of greater than or equal to [an] OBE. These actions are 
consistent with present guidance in the seismic response procedure and 
do not create an undue burden on the operator. To compensate for the 
potential loss of Boraflex after a seismic event, the SFP is now 
required to be [] borated at all times to at least 1750 ppm to maintain 
the proper post seismic K-effective condition. As such, there is no 
mitigation equipment that has to operate in the spent fuel pool 
following a seismic event.
    Although the Boraflex in the fuel racks is assumed to fail in a 
seismic event greater than an OBE, the presence of soluble boron in the 
fuel pool water will compensate for the loss of Boraflex. Surveillance 
requirements on SFP boron will ensure that there will be boron present 
in the SFP and ensure that the SFP is not diluted below the minimum 
required boron concentration during normal operation.
    As the presence of SFP soluble boron during and after a seismic 
event maintains k-effective less than 0.95 there is no effect on the 
consequences of any accidents evaluated. As there are no new accidents 
created, there are no changes in the consequences of previously 
analyzed accidents, and there is no effect on the consequences of any 
accident. There is no reduction in the margin of safety as the result 
of the degradation of Boraflex following a greater than OBE seismic 
event, because during normal operations k-effective remains less than 
0.95 without reliance on soluble boron, and during malfunction and 
accident conditions soluble boron can be used to compensate for the 
loss of Boraflex to maintain K-effective less than 0.95.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.
    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

[[Page 63982]]

    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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

    Date of amendment request: October 3, 1997.
    Description of amendment request: Omaha Public Power District 
(OPPD) proposes to change the Fort Calhoun Station Unit No. 1 Technical 
Specifications (TS) by revising TS Surveillance Requirement 3.9, 
``Auxiliary Feedwater System,'' to clarify what flow paths are required 
to be tested. Additionally, OPPD proposes to revise the auxiliary 
feedwater pumps' surveillance requirements to delete the specific 
discharge pressure.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    A change to TS 3.9(2) is proposed to delete the specific discharge 
pressure specified for the Auxiliary Feedwater (AFW) pumps' 
surveillance. The developed head of the motor-driven and steam turbine-
driven AFW pumps is verified quarterly. These tests are in addition to 
those required by TS 3.3, which implements ASME Section XI Inservice 
Testing (IST) to evaluate a pump's performance against its pump curve 
to determine operability. The IST program is controlled by TS 3.3, and 
requires that testing of ASME Code Class 1, Class 2, and Class 3 pumps 
shall be performed in accordance with Section XI of the ASME Boiler and 
Pressure Vessel Code, as required by 10 CFR 50.55a(g), except where 
specific written relief has been granted by the NRC. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to TS 3.9(4) and the Basis Section only 
clarify the AFW flow paths that are required to be tested. The proposed 
change follows the recommendations of NUREG-0635, ``Generic Evaluation 
of Feedwater Transients and Small Break Loss-of-Coolant Accidents in 
Combustion Engineering Designed Operating Plants,'' Recommendation GS-
6(2). No physical changes are proposed, information is being added to 
clarify the testing required to meet the recommendations of NUREG-0635, 
therefore these proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    There will be no physical alterations to the plant configuration or 
changes in operating modes. The proposed change to delete the specific 
discharge pressure of the AFW pumps from the TS is consistent with the 
ASME Code Section XI requirements that are controlled by TS 3.3. 
Testing requirements of TS 3.3 require testing of ASME Code Class 1, 
Class 2, and Class 3 pumps in accordance with Section XI of the ASME 
Boiler and Pressure Vessel Code, as required by 10 CFR 50.55a(g), 
except where specific written relief has been granted by the NRC. The 
clarifications being provided to describe the flow paths only provide 
additional information for testing required to meet the recommendation 
of NUREG-0635.
    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes will not result in any physical alterations to 
the plant configuration or changes to the application of setpoints or 
limits. Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: November 4, 1997.
    Description of amendment request: The amendments would change the 
Emergency Diesel Generator (EDG) Technical Specification (TS) 3/4.8.1 
to (1) delete 18-month surveillance requirement 4.8.1.1.2.d.1, and (2) 
eliminate the accelerated testing requirement of Table 4.8-1. Both 
changes have been approved on other nuclear power facilities.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed change deleting the requirement for an 18 month diesel 
inspection is consistent with the improved Standard Technical 
Specifications (NUREG-1433) and does not result in any changes to the 
existing plant design. The Salem preventive maintenance program 
utilizes diesel generator performance history, engineering analyses and 
manufacturer's recommendations as appropriate for determining diesel 
generator inspection requirements. The Technical Specifications will 
continue to contain surveillance requirements that demonstrate the 
functional capability of the diesel generators. The change does not 
impact the ability of the diesel generators or the AC electrical power 
sources to perform their function, nor result in a significant increase 
in the consequences of any accident previously evaluated. The diesel 
generators will continue as designed.
    PSE&G has implemented the provisions of the maintenance rule for 
EDG's, including the appropriate regulatory guidance. This provides a 
program which assures EDG performance. The elements of this program 
include the performance of detailed root cause analysis of individual 
failures, effective corrective actions taken in response to individual 
failures, and implementation of preventive maintenance consistent with 
the Maintenance Rule. Additionally, the proposed changes (elimination 
of accelerated diesel generator testing requirements of TS 4.8.1.1.a in 
lieu of monthly testing and deletion of special

[[Page 63983]]

reporting requirements for diesel failures), do not delete the 
surveillance requirements but rather set their frequency at every 31 
days. Monitoring the effectiveness of EDG maintenance and continuing 
surveillance testing will ensure that the diesel generators will 
perform their intended functions and will minimize failures. As is 
noted in the recommendations of GL [Generic Letter] 94-01, because 
PSE&G is monitoring and maintaining EDG performance in accordance with 
the provisions of 10 CFR 50.65, there is no longer a need for special 
reporting requirements.
    Since the changes do not affect the assurance of diesel generator 
reliability or operability as discussed above, there is no significant 
increase in the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously analyzed.
    This request does not result in any change to the plant design or 
does it involve a significant change in current plant operation. The 
diesel generators are inspected utilizing diesel generator operating 
history, engineering analyses and manufacturer's recommendations as 
appropriate, and the remaining surveillance requirements continue to 
demonstrate the functional capability of the diesel generators.
    Changing the surveillance of frequency of TS 4.8.1.2.a to 31 days 
the existing frequency as determined by Table 4.8-1, does not create a 
new or different kind of accident. Deleting of special reporting 
requirements, appropriate in light of the monitoring and maintenance in 
conformance with 10 CFR 50.65, and reliance on the reporting 
requirements of 10 CFR 50.72 and 10 CFR 50.73, does not create the 
possibility of a new or different kind of accident.
    The proposed changes do not result in any change to the plant 
design nor do they involve a significant change in current plant 
design. No new failure modes will be introduced. Therefore, the 
proposed 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.
    The proposed request does not adversely impact the reliability of 
the diesel generators. As stated above, the diesel generator operating 
history, engineering analyses and the manufacturer's recommendations 
will be utilized as appropriate to perform diesel generator 
inspections. Additionally, other Technical Specification surveillance 
requirements will continue to demonstrate the functional capability of 
the diesel generators. The diesel generators will continue to perform 
their design functions.
    Noting the monitoring and maintenance being performed in 
conformance with 10 CFR 50.65, revision of the frequency of 
surveillance testing of 4.8.1.1.2.a does not adversely impact the 
reliability of the diesel generators. Deletion of the special reporting 
requirements of 4.8.1.1.4 does not impact the operability or the 
reliability of the diesel generators.
    This request does not involve an adverse impact on diesel generator 
operation or reliability. Since the diesel generator function is not 
affected by the proposed change, this request does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: John F. Stolz.

Southern Nuclear Operating Company, Inc, Docket Nos. 50-348 and 50-364, 
Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, Alabama

    Date of amendments request: October 16, 1997.
    Description of amendments request: The proposed amendments would 
revise the Farley Nuclear Plant (FNP) Units 1 and 2 Technical 
Specifications (TS) to increase the allowable number of charging pumps 
capable of injecting into the reactor coolant system (RCS) when the 
temperature of one or more of the RCS cold legs is 180 deg.F or less. 
The amendments would also modify the FNP TS to allow a maximum of two 
charging pumps to be capable of injecting into the RCS during pump swap 
operations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes to TS 3.1.2.3 allow two charging pumps to be 
capable of injecting into the reactor coolant system (RCS) for a period 
not to exceed 15 minutes while RCS cold leg temperature is at or below 
180 degrees F. The intent is to allow the operator to start a second 
pump long enough to ensure that it operates properly and then to 
promptly secure the pump that was originally running. This order of 
pump operation will allow seal injection flow to be maintained to the 
RCS pumps number one seal continuously, thus preventing loss of 
pressure to the seals and maintaining filtered water flow through the 
seals. The proposed revised bases address the potential for [an] RCS 
mass addition transient. Guidance is given to prevent the charging pump 
swap from being conducted while the RCS is in a condition conducive to 
an overpressure transient. The RCS should be in a non water solid 
condition and the residual heat removal (RHR) relief valves must be 
operable or the RCS must be vented while the pump swap evolution is in 
progress. The proposed revision to TS 3.1.2.3 allows 15 minutes to have 
two pumps capable of injecting into the RCS, although two pumps will be 
running only momentarily, the remaining time is needed to perform the 
charging pump circuit breaker racking operations needed to render one 
of the two pumps incapable of injecting into the RCS. The proposed 
actions statement 3.1.2.3b directs that immediate action be taken to 
render all but one pump inoperable should the allotted 15 minutes be 
exceeded. This action is more appropriate than is currently specified. 
These proposed changes include sufficient controls to prevent an RCS 
overpressurization event.
    Therefore, the proposed TS changes do not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    This proposed change involves no change to the physical plant. It 
allows for a very limited and controlled operational change. The change 
increases the potential for a mass addition transient while the RCS is 
[at or] below 180 degrees F; however, sufficient controls are proposed 
to prevent a cold overpressure event.

[[Page 63984]]

    Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction in a 
margin of safety.
    The proposed change includes controls sufficient to prevent a 
significant reduction in the possibility or consequences of an 
accident. The proposed change specifies that the pump swap evolution be 
performed under conditions that will prevent an adverse plant 
transient. In addition, the proposed revision provides appropriate 
operator action that does not currently exist. This change is 
consistent with NUREG 1431, Standard Technical Specifications-
Westinghouse Plants.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama.
    NRC Project Director: Herbert N. Berkow.

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

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

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: October 4, 1997.
    Brief description of amendment: The proposed amendment revises the 
description of the electrical controls for Operating Reactor Building 
Recirculation System Fan/Cooler contained in the Final Safety Analysis 
Report and Improved Technical Specification Bases.
    Date of publication of individual notice in the Federal Register: 
November 13, 1997 (62 FR 60921).
    Expiration date of individual notice: December 15, 1997.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

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: October 31, 1997.
    Brief description of amendment: The proposed amendment revises 
Operating License No. DPR-72, License Condition 2.C.(5) and deletes the 
requirement for installation and testing of flow indicators in the 
emergency core cooling system.
    Date of publication of individual notice in the Federal Register: 
November 12, 1997 (62 FR 60733).
    Expiration date of individual notice: December 12, 1997.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

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: October 31, 1997.
    Brief description of amendment: The proposed amendment involves 
revisions to the Crystal River 3 Technical Specifications (TS) relating 
to decay heat removal requirements in Mode 4.
    Date of publication of individual notice in the Federal Register: 
November 12, 1997 (62 FR 60735).
    Expiration date of individual notice: December 12, 1997.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428.

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: October 31, 1997.
    Brief description of amendment: The proposed amendment involves 
revisions to the Crystal River 3 Technical Specifications (TS) relating 
to the methodology for post-loss of coolant accident boron 
precipitation prevention.
    Date of publication of individual notice in the Federal Register: 
November 12, 1997 (62 FR 60731).
    Expiration date of individual notice: December 12, 1997.
    Local Public Document Room location: Coastal Region Library, 8619 
W., Crystal River, Florida 34428.

Notice of Issuance of Amendments to Facility Operating Licenses

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

[[Page 63985]]

Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document rooms for the particular facilities involved.

Duke Energy Corporation, et al., Docket No. 50-414, Catawba Nuclear 
Station, Unit 2, York County, South Carolina

    Date of application for amendment: May 27, 1997.
    Brief description of amendment: The amendment deletes references to 
steam generator tube sleeving and repair criteria that will not be used 
for the Westinghouse Model D5 steam generators in use at Catawba Unit 
2. Also, unused paragraph numbers have been deleted and a typographical 
error has been corrected.
    Date of issuance: November 13, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 154.
    Facility Operating License No. NPF-52: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33122).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of amendment request: October 7, 1997.
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications (TSs) by modifying TS 3.3.3.7.3, and 
Surveillance Requirements (SR) 4.3.3.7.3 for the broad range gas 
detection system. Also it makes some changes to the Bases in section 3/
4.3.3.7 to incorporate information associated with the existing toxic 
gas monitors.
    Date of issuance: November 14, 1997.
    Effective date: November 14, 1997, to be implemented within 60 
days.
    Amendment No.: 135.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 15, 1997 (62 FR 
53660).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 14, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of amendment request: February 6, 1996.
    Brief description of amendment: The proposed change will amend the 
Allowable Values of parameters in Table 3.3-4 of Waterford Steam 
Electric Station, Unit 3, (Waterford 3) Technical Specifications (TSs) 
to make it consistent with the identical parameters in Table 2.2-1 of 
TSs for Waterford 3. The proposed change will add Mode 4 to 
surveillance requirements of Table 4.3-2, Item 5.c (Safety Injection 
System Automatic Actuation Logic) that was inadvertently removed. 
Finally, the proposed change removes a reference to TS 3.3.3.2 in 
Surveillance Requirements TS 4.10.2.2 and 4.10.4.2 since Incore 
Detectors has been removed from the TSs.
    Date of issuance: November 20, 1997.
    Effective date: November 20, 1997, to be implemented within 60 
days.
    Amendment No.: 136.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 5, 1996 (61 FR 
28615).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 20, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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 23, 1996, as supplemented by 
letters dated October 1 and 15, 1996, and January 28, 1997.
    Brief description of amendments: The amendments reflect the 
approval of the transfer of the authority to operate South Texas 
Project, Units 1 and 2, under the licenses to a new operating company, 
South Texas Project Nuclear Operating Company.
    Date of issuance: November 17, 1997.
    Effective date: November 17, 1997.
    Amendment Nos.: Unit 1--Amendment No. 93; Unit 2--Amendment No. 80.
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical Specifications and the operating licenses.
    Date of initial notice in Federal Register: November 7, 1996 (61 FR 
57719).
    The additional information contained in the supplemental letter 
dated January 28, 1997, was clarifying in nature and thus, it was 
within the scope of the initial notice and did not affect the staff's 
proposed no significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 17, 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: May 5, 1997.
    Brief description of amendment: Technical Specification 
Surveillance 4.8.4.1 requires periodic testing of lower voltage circuit 
breakers for all containment penetration conductor overcurrent 
protective devices. The amendment modifies the requirements for 
determining the operability of lower voltage circuit breakers by using 
the manufacturer's curve of current versus time to test delay trip 
elements, clarifies the use of two pole in series testing, and expands 
the Bases description of the testing.
    Date of issuance: November 14, 1997.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 153.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30637).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 14, 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,

[[Page 63986]]

Norwich, Connecticut 06360, and the Waterford Library, ATTN: Vince 
Juliano, 49 Rope Ferry Road, Waterford, Connecticut 06385.

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

    Date of application for amendments: May 15, 1997, as supplemented 
August 29, October 20, October 24, and October 28, 1997.
    Brief description of amendments: The amendments revise certain 
Technical Specification (TS) limitations on reactor coolant system 
leakage and steam generator tube surveillance, and implement a voltage-
based repair criteria per requirements of NRC Generic Letter 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes 
Affected by Outside Diameter Stress Corrosion Cracking.'' In addition, 
the amendments correct a typographical error in TS Section 4.12.c.
    Date of issuance: November 18, 1997.
    Effective date: November 18, 1997, with full implementation of the 
Technical Specifications within 30 days. License Condition 5 of 
Appendix B shall be implemented immediately upon issuance of the 
amendments.
    Amendment Nos.: 133 and 125.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the licenses and the Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43371).
    The August 29, October 20, October 24, and October 28, 1997, 
supplements provided clarifying information that did not change the 
staff's initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 18, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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

    Date of application for amendment: August 19, 1997, as supplemented 
September 17, 1997.
    Brief description of amendment: The amendment revised the Ginna 
Station Improved Technical Specifications to correct an error in the 
required accumulator borated water volume specified in Surveillance 
Requirement 3.5.1.2.
    Date of issuance: November 10, 1997.
    Effective date: November 10, 1997.
    Amendment No.: 69.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 8, 1997 (62 FR 
52587).
    The September 17, 1997, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 10, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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 1, Lake County, Ohio

    Date of application for amendment: October 24, 1996, as 
supplemented June 16 and October 2, 1997.
    Brief description of amendment: This amendment revised the minimum 
critical power ratio safety limit to reflect the 10 CFR Part 21 
condition reported by General Electric in their letter to the NRC dated 
May 24, 1996.
    Date of issuance: November 7, 1997.
    Effective date: November 7, 1997.
    Amendment No.: 91.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6569). The June 16 and October 2, 1997, supplemental letters 
provided additional clarifying information and did not change the 
initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated November 7, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

    Dated at Rockville, Maryland, this 25th day of November 1997.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 97-31522 Filed 12-2-97; 8:45 am]
BILLING CODE 7590-01-P