[Federal Register Volume 62, Number 250 (Wednesday, December 31, 1997)]
[Notices]
[Pages 68303-68323]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-33968]


=======================================================================
-----------------------------------------------------------------------

NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 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 December 6, 1997, through December 18, 1997. 
The last biweekly notice was published on December 17, 1997 (62 FR 
66133).

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 30, 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

[[Page 68304]]

Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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 26, 1997.
    Description of amendments request: Carolina Power & Light Company 
(CP&L) has proposed amendments to the Technical Specifications (TS) for 
the Brunswick Steam Electric Plant Units 1 and 2 (BSEP 1 & 2) to revise 
certain instrumentation allowable values. The revised values were 
calculated using a methodology and format consistent with that provided 
in NUREG-1433, Revision 1, ``Standard Technical Specifications General 
Electric Plants, BWR/4.'' The current TS are based on the uncertainty 
associated with the trip unit portion of the instrumentation circuitry. 
The proposed values are based on the uncertainty associated with the 
entire instrumentation loop (sensor and trip unit). The NRC has 
previously approved this methodology for BSEP 1 & 2 as part of a 5 
percent power uprate amendment dated November 1, 1996.
    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 affect accident mitigation instrumentation 
allowable values. The changes will not affect the accident 
mitigation instrumentation functions. No changes will occur in the 
way in which equipment is operated. Therefore, the probability of a 
previously evaluated accident can not be affected.
    The proposed changes establish the allowable values for certain 
functions in accordance with the CP&L setpoint methodology, which 
has been approved, by the NRC, for use at the BSEP. The proposed 
changes do not affect the actual instrument setpoints. The proposed 
allowable values were calculated by applying calibration based 
errors to the trip setpoint values; thereby establishing an 
operability limit associated with the entire loop of an 
instrumentation function to ensure sufficient margin to protect 
analytical limits. The changes do not affect the analytical limits 
associated with the involved instrumentation functions. The involved 
instrumentation will continue to perform its accident mitigation 
functions as designed. Therefore, the consequences of a previously 
evaluated accident are not increased.
    2. The proposed amendments would not create the possibility of a 
new or different

[[Page 68305]]

kind of accident from any accident previously evaluated.
    The proposed changes do not affect the actual instrument 
setpoints nor do they affect the accident mitigation instrumentation 
functions. No changes will occur in the way in which equipment is 
operated. The involved instrumentation will continue to perform its 
accident mitigation functions as designed. Therefore, the proposed 
license amendments can not create the possibility of a new or 
different kind of accident.
    3. The proposed license amendments do not involve a significant 
reduction in a margin of safety.
    The proposed changes affect accident mitigation instrumentation 
allowable values. The changes will not affect the accident 
mitigation instrumentation functions. No changes will occur in the 
way in which equipment is operated. The proposed changes establish 
the allowable values for certain functions in accordance with the 
CP&L setpoint methodology which has been approved, by the NRC, for 
use at the BSEP. The proposed allowable values were calculated by 
applying calibration based errors to the trip setpoint values; 
thereby establishing an operability limit associated with the entire 
loop of an instrumentation function to ensure sufficient margin to 
protect analytical limits. The changes do not affect the analytical 
limits associated with the involved instrumentation functions. The 
involved instrumentation will continue to perform its accident 
mitigation functions as designed. Therefore, the proposed license 
amendments do not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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 No. 50-400, Shearon Harris

Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North Carolina

    Date of amendment request: October 29, 1997.
    Description of amendment request: Technical Specifications (TS) 
3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 presently require a plant 
shutdown and declaring the redundant required feature inoperable, when 
the required feature powered from the operable A.C. source is 
inoperable. The proposed change clarifies the intent of this TS to 
permit the applicable redundant required feature TS to direct a plant 
shutdown when required. The proposed amendment changes the existing TS 
3.8.1.1.a.3, 3.8.1.1.b.4, and 3.8.1.1.d.2 to eliminate the separate 
requirement for plant shutdown and instead allows the applicable 
required redundant feature TS to direct the plant shutdown when 
required.
    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:

    This change does not involve a significant hazards consideration 
for the following reasons:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment will not introduce any new equipment or 
require existing equipment to function different from that 
previously evaluated in the Final Safety Analysis Report (FSAR) or 
TS. The changes are consistent with NUREG-1431 and the Commission's 
Final Policy Statement on Technical Specification improvements.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment will not introduce any new equipment or 
require existing equipment to function different from that 
previously evaluated in the Final Safety Analysis Report (FSAR) or 
TS. The changes are consistent with NUREG-1431 and the Commission's 
Final Policy Statement on Technical Specification improvements. The 
proposed amendment will not create any new accident scenarios, 
because the change does not introduce any new single failures, 
adverse equipment or material interactions, or release paths.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    Margin of safety for acceptable TS action times have been 
determined for each TS related system. The proposed change will not 
alter individual system TS action times. HNP [the Harris Nuclear 
Plant] proposes to change the requirement to shutdown after 
expiration of the completion time of an inoperable A.C. source 
concurrent with an inoperable required feature. Instead of requiring 
a shutdown, the required feature on the inoperable A.C. source will 
be declared inoperable and the individual TS will be implemented.
    In most cases with both redundant features inoperable, a plant 
shutdown will be required by TS 3.0.3. In the few instances where 
additional time is allowed by the individual TS for both redundant 
required features being inoperable, then an immediate plant shutdown 
would not be required. The allowed out of service time for loss of 
individual safety functions has been previously analyzed for HNP TS 
and NUREG-1431, Revision 1.
    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    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.

Florida Power and Light Company, et al.

[Docket Nos. 50-335 and 50-389]

St. Lucie Plant, Unit Nos. 1 and 2, St. Lucie County, Florida

    Date of amendment request: December 1, 1997.
    Description of amendment request: The proposed amendment revises 
the Unit 1 and Unit 2 Environmental Protection Plans (EPP) Section 4, 
``Environmental Conditions,'' and Section 5, ``Administrative 
Procedures,'' to incorporate the proposed terms and conditions of the 
Incidental Take Statement included in the Biological Opinion issued by 
the National Marine Fisheries Service (NMFS) on February 7, 1997. The 
proposed amendment also revises the wording in the Unit 1 EPP to make 
it consistent with the Unit 2 EPP.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the

[[Page 68306]]

probability or consequences of an accident previously evaluated.
    The changes are administrative in nature and would in no way 
affect the initial conditions, assumptions, or conclusions of the 
St. Lucie Unit 1 or Unit 2, accident analyses. In addition, the 
proposed changes would not affect the operation or performance of 
any equipment assumed in the accident analyses.
    Based on the above information, we conclude that the proposed 
changes would not significantly increase the probability or 
consequences of an accident previously evaluated.
    (2) Use of the modified specification would not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The changes are administrative in nature and would in no way 
impact or alter the configuration or operation of the facilities and 
would create no new modes of operation. We conclude that the 
proposed changes would not create the possibility of a new or 
different kind of accident.
    (3) Use of the modified specification would not involve a 
significant reduction in a margin of safety.
    As indicated in the discussion of Criterion 1, the changes are 
administrative in nature and would in no way affect plant or 
equipment operation or the accident analysis. We conclude that the 
proposed changes would not result in a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Community College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34981-5596.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
P.O. Box 14000, Juno Beach, Florida 33408-0420.
    NRC Project Director: Frederick J. Hebdon.

IES Utilities Inc.

[Docket No. 50-331]

Duane Arnold Energy Center, Linn County, Iowa

    Date of amendment request: October 30, 1996.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from current Technical 
Specifications (CTS) to improved Technical Specifications (ITS), would 
modify the Surveillance Requirements (SRs) recommended in NUREG-1433 
LOC 3.5.1 by revising the combinations (Conditions C, D, G, and I of 
ITS 3.5.1) of emergency core cooling systems/subsystems that may be out 
of service. The combinations are supported by the Duane Arnold Energy 
Center (DAEC) Loss-of-Coolant Accident (LOCA) analysis.
Condition C
    ITS 3.5.1  Action C establishes Required Actions and Completion 
Times for the situation when one core spray (CS) subsystem and one or 
two residual heat removal (RHR) pump(s) are inoperable. The proposed 
specification is less restrictive than CTS 3.5.A.4, which allows one 
RHR pump to be inoperable for 30 days, and CTS 3.5.A.5, which allows 
two RHR pumps (i.e., the low pressure coolant injection (LPCI) 
subsystem) to be inoperable for up to 7 days, provided the remaining 
RHR (i.e., LPCI) active components, both CS subsystems, the containment 
spray subsystem, and the diesel generators are verified to be operable. 
The CTS does not allow one CS subsystem and one or two RHR pump(s) to 
be inoperable at the same time. The LOCA analysis presented in NEDC-
31310P, (Duane Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant 
Accident Analysis), indicates that an adequate level of protection is 
provided by the remaining operable ECCS subsystems. The accident 
analysis also demonstrates that in this condition, the peak clad 
temperature remains below the regulatory limit. However, another single 
failure may place the plant in a condition where adequate core cooling 
may not be available during a DBA-LOCA. Therefore, a Completion Time of 
72 hours has been proposed to either restore the inoperable CS 
subsystem or the inoperable RHR pump(s).
Condition D
    ITS 3.5.1  Action D establishes Required Actions and Completion 
Times for the situation when two CS subsystems are inoperable. The 
proposed specification is less restrictive than CTS 3.5.A.2, which 
allows only one CS subsystem to be inoperable. CTS 3.5.A.6 would 
require the plant to be in Hot Shutdown within 12 hours and Cold 
Shutdown within the following 24 hours if both CS subsystems were 
inoperable. With two CS subsystems inoperable, the LOCA analysis 
presented in NEDC-31310P, (Duane Arnold Energy Center SAFER/GESTR-LOCA 
Loss-of-Coolant Accident Analysis), indicates that the remaining 
operable low pressure ECCS subsystem consisting of LPCI with four RHR 
pumps operable (only 3 pumps required), provides adequate protection. 
However, another single failure may place the plant in a condition 
where adequate core cooling may not be available during a Design Basis 
Accident LOCA. Therefore, a Completion Time of 72 hours has been 
proposed to restore one CS subsystem to operable status.
Condition G
    ITS 3.5.1 Action G establishes Required Actions and Completion 
Times for the situation when HPCI and one RHR pump are inoperable. The 
proposed specification is less restrictive than CTS 3.5.D.2, which 
allows continued operation if HPCI is inoperable only if both CSs, 
LPCI, ADS, and RCIC are verified to be operable. While the LPCI 
subsystem is technically operable with only 3 of 4 RHR pumps operable, 
the CTS is currently interpreted by DAEC to require all 4 RHR pumps to 
be operable for the requirements of CTS 3.5.D.2 to be met, as a single 
RHR pump has more makeup capability than the HPCI System. Thus for 
mitigating small and intermediate break LOCAs, one LPCI pump, in 
combination with ADS, is more than adequate core cooling. The condition 
of when HPCI and one RHR pump are inoperable is bounded by the analysis 
in NEDC-31310P, Duane Arnold Energy Center, SAFER/GESTR-LOCA Loss-of-
Coolant Accident Analysis. Since the remaining operable low pressure 
ECCS subsystems are more than capable of performing their intended 
function, and RCIC and ADS are Operable, the proposed Action G 
maintains LOCA analysis assumptions for ECCS Operability. The proposed 
ITS condition allows 7 days to restore the HPCI System or the RHR pump 
to operable status. The licensee considers the 7 day Completion Time 
reasonable in that the LOCA analysis demonstrates that in this 
condition, the peak clad temperature remains below the regulatory 
limit. The 7 day Completion Time also provides the benefit of 
potentially avoiding an unnecessary plant shutdown while the safety 
functions are still capable of being performed.
Condition I
    ITS 3.5.1 Action I establishes Required Actions and Completion 
Times for the situation when HPCI and one ADS valve are inoperable. The 
proposed Specification is less restrictive than CTS 3.5.D.2, which 
allows continued operation if HPCI is inoperable only if both CSs, 
LPCI, ADS, and RCIC are verified to be operable. While ADS is capable 
of performing its design function with only 3 of 4 valves operable, per 
NEDC-31310P, Duane Arnold Energy Center, SAFER/GESTR-

[[Page 68307]]

LOCA Loss-of-Coolant Accident Analysis, the CTS requires all 4 ADS 
valves to be operable for the requirements of CTS 3.5.D.2 to be met. 
The proposed specification is less restrictive than CTS 3.5.F.2, which 
allows continued operation when one ADS valve is inoperable only if 
HPCI is verified to be operable. Since all low pressure ECCS subsystems 
remain capable of performing their design function and ADS is still 
capable of performing its design function, ITS 3.5.1 Action I maintains 
LOCA assumptions to ensure an adequate level of protection is 
maintained. The proposed condition allows 72 hours to restore the HPCI 
system or the ADS valve to operable status, since another single 
failure (i.e., loss of another ADS valve), may place the plant in a 
condition where adequate core cooling may not be available during a 
small or intermediate break LOCA.
    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:

For Condition C

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will allow one Core Spray subsystem and one 
or two RHR pump(s) to be inoperable for up to 72 hours. The ECCS 
subsystems affected by this change are not assumed to be initiators 
of analyzed events. Therefore, the proposed change does not increase 
the probability of any accident. The role of these ECCS subsystems 
is in the mitigation of accident consequences. The proposed change 
does not allow unlimited continuous operation with the plant in a 
condition where an additional single failure could result in a loss 
of ECCS function. The proposed change does not increase the 
consequences of an accident because accident analysis presented in 
NEDC-31310P, Duane Arnold Energy Center SAFER/GESTR-LOCA Loss-of-
Coolant Accident Analysis, indicates that an adequate level of 
protection is maintained by the ADS System and the remaining 
Operable ECCS subsystems when one Core Spray subsystem and one or 
two RHR pump(s) are inoperable. Therefore, this change will not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change will not involve any physical changes to 
plant systems, structures, or components (SSCs), or the manner in 
which these SSCs are operated, maintained, modified, tested or 
inspected. The change ensures the remaining ECCS capability is 
adequate to mitigate the consequences of accidents. Therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change does not significantly reduce the margin of 
safety because accident analysis presented in NEDC-31310P, Duane 
Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident 
Analysis, indicates that the plant is protected by the ADS System 
and the remaining ECCS subsystems when one Core Spray subsystem and 
one or two RHR pump(s) are inoperable. The accident analysis 
demonstrates that in this condition, the peak clad temperature 
remains below the regulatory limit. However, with one Core Spray 
subsystem and one or two RHR pump(s) inoperable, another single 
failure may place the plant in a condition where adequate core 
cooling may not be available during a DBA-LOCA. Therefore, a 
Completion Time of 72 hours has been assigned to either restore the 
inoperable Core Spray subsystem or the RHR pump. In addition, this 
change provides the benefit of potentially avoiding an unnecessary 
plant shutdown (due to a Completion Time being provided for one Core 
Spray subsystem and one or two RHR pump(s)) when the remaining ECCS 
subsystems and the ADS are capable of mitigating potential events. 
Therefore, this change does not involve a significant reduction in a 
martin safety.

For Condition D

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will allow both Core Spray subsystems to be 
inoperable for up to 72 hours. The ECCS subsystems affected by this 
change are not assumed to be initiators of analyzed events. 
Therefore, the proposed change does not increase the probability of 
any accident. The role of these ECCS subsystems is in the mitigation 
of accident consequences. The proposed change does not allow 
unlimited continuous operation with the plant in a condition where 
an additional single failure could result in a loss of ECCS 
function. The proposed change does not increase the consequences of 
an accident because accident analysis presented in NEDC-3131OP, 
Duane Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident 
Analysis, indicates that an adequate level of protection is 
maintained by the ADS System and remaining Operable ECCS subsystem 
when two Core Spray subsystems or inoperable. Therefore, this change 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change will not involve any physical changes to 
plant systems, structures, or components (SSCs), or the manner in 
which these SSCs are operated, maintained, modified, tested, or 
inspected. The change ensures the remaining ECCS capability is 
adequate to mitigate the consequences of accidents. Therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change does not significantly reduce the margin of 
safety because accident analysis presented in NEDC-31310P, Duane 
Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident 
Analysis, indicates that the plant is protected by the ADS System 
and the remaining ECCS subsystem when two Core Spray subsystems are 
inoperable. The accident analysis demonstrates that in this 
condition, the peak clad temperature remains below the regulatory 
limit. However, with both Core Spray subsystems inoperable, another 
single failure may place the plant in a condition where adequate 
core cooling may not be available during a DBA-LOCA. Therefore, a 
Completion Time of 72 hours has been assigned to restore one 
inoperable Core Spray subsystem. In addition this change provides 
the benefit of potentially avoiding an unnecessary plant shutdown 
(due to a Completion Time being provided for both Core Spray 
subsystems inoperable) when the remaining ECCS subsystem and the ADS 
are capable of mitigating potential events. Therefore, this change 
does not involve a significant reduction in a margin of safety.

Condition G

    1. Does the change involve a significant increase in the 
probability or consequences or an accident previously evaluated?
    The proposed change will allow the HPCI System and one RHR pump 
to be inoperable for up to 7 days. The ECCS subsystems affected by 
this change are not assumed to be initiators of analyzed events. 
Therefore, the proposed change does not increase the probability of 
any accident. The role of these ECCS subsystems is in the mitigation 
of accident consequences. The proposed change does not allow 
unlimited continuous operation with the plant in a condition where 
an additional single failure could result in a loss of ECCS 
function. The proposed change does not increase the consequences of 
an accident because accident analysis presented in NEDC-31310P, 
Duane Arnold Energy Center SAFER/GESTRA-LOCA Loss-of-Coolant 
Accident Analysis, indicated that an adequate level of protection is 
maintained by the ADS System and the remaining Operable ECCS 
subsystems when HPCI and one RHR pump are inoperable. Therefore, 
this change will not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change will not involve any physical changes to 
plant systems, structures, or components (SSCs), or the manner in 
which these SSCs are operated, maintained, modified, tested, or 
inspected. The change ensures the remaining ECCS capability is 
adequate to mitigate the consequences of accidents. Therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluate.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change does not significantly reduce the margin of 
safety because accident

[[Page 68308]]

analysis presented in NEDC-31310P, Duane Arnold Energy Center SAFER/
GESTR-LOCA Loss-of-Coolant Accident Analysis, indicates that the 
plant is protected by the ADS System and the remaining ECCS 
subsystems when HPCI and one RHR pump are inoperable. The accident 
analysis demonstrates that in this condition, the peak clad 
temperature remains below the regulatory limit. However, with both 
HPCI and one RHR pump inoperable, another single failure may place 
the plant in a condition where adequate core cooling may not be 
available during an accident. Therefore, a Completion Time of 7 days 
has been assigned to either restore the inoperable HPCI System or 
the RHR pump. In addition, this change provides the benefit of 
potentially avoiding an unnecessary plant shutdown (due to a 
Completion Time being provided for the HPCI System and one RHR pump 
inoperable) when the remaining ECCS subsystems and the ADS are 
capable of mitigating potential events. Therefore, this change does 
not involve a significant reduction in a margin of safety.

Condtion I

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will allow the HPCI system and one ADS valve 
to be inoperable for up to 72 hours. The ECCS subsystems affected by 
this change are not assumed to be initiators or analyzed events. 
Therefore, the proposed change does not increase the probability of 
any accident. The role of these ECCS subsystems is in the mitigation 
of accident consequences. The proposed change does not allow 
unlimited continuous operation with the plant in a condition where 
an additional single failure could result in a loss of ECCS 
function. The proposed change does not increase the consequences of 
an accident because accident analysis presented in NEDC-31310P, 
Duane Arnold Energy Center SAFER/GESTER-LOCA Loss-of-Coolant 
Accident Analysis, indicates that an adequate level of protection is 
maintained by the remaining ADS valves (the ADS design function is 
maintained) in combination with the remaining Operable ECCS 
subsystems when HPCI and one ADS valve are inoperable. Therefore, 
this change will not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or difference 
kind of accident form any accident previously evaluated?
    The proposed change will not involve any physical changes to 
plant systems, structures, or components (SSCs) or the manner in 
which these SSCs are operated, maintained, modified, tested, or 
inspected. The change ensures the remaining ECCS capability in 
adequate to mitigate the consequences of accidents. Therefore, this 
change will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Does this change involve a significant reduction in a margin 
of safety?
    The proposed change does not significantly reduce the margin of 
safety because accident analysis presented in NEDC-31310P, Duane 
Arnold Energy Center SAFER/GESTR-LOCA Loss-of-Coolant Accident 
Analysis, indicates that the plant is protected by the remaining ADS 
valves and the low pressure ECCS subsystems when HPCI and one ADS 
valve are inoperable. The accident analysis demonstrates that in 
this condition, the peak clad temperature remains below the 
regulatory limit. However, with both HPCI and one ADS valve 
inoperable, another single failure (i.e., of an ADS valve) may place 
the plant in a condition where adequate core cooling may not be 
available during a small or intermediate break LOCA. Therefore, a 
Completion Time of 72 hours has been assigned to either restore the 
inoperable HPCI System or the ADS valve. In addition, this change 
provides the benefit of potentially avoiding an unnecessary plant 
shutdown (due to a Completion Time being provided for the HPCI 
System and one ADS valve inoperable) when the remaining ECCS 
subsystems and ADS valves are capable of mitigating potential 
events. Therefore, this 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: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869.
    Acting NRC Project Director: Richard P. Savio.

Indiana Michigan Power Company

[Docket Nos. 50-315 and 50-316]

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: August 1, 1997 (AEP:NRC:0906H).
    Description of amendment requests: The proposed amendments would 
revise Technical Specification surveillance 4.7.1.2.b. to delete the 
requirement that the test be performed at a specified secondary steam 
supply 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:

Criterion 1

    The proposed changes will not significantly increase the 
probability or consequences of an accident previously evaluated.
    This is an administrative change intended to clarify the 
technical specification. There will be no change to the test 
procedure as a result of this clarification. The proposed change 
better correlates with the accident requirements for which TDAFP 
[turbine driven auxiliary feed pump] flow is required, and the 
change is consistent with the present requirement of testing the 
TDAFP at a secondary side pressure greater than 310 psig.

Criterion 2

    The proposed changes will not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The proposed change does not physically modify the plant, nor does 
it result in the installation of equipment which could introduce a 
new failure mechanism.

Criterion 3

    The proposed change does not involve a significant reduction in 
a margin of safety. The proposed change does not affect the 
performance of the TDAFP. Thus, the TDAFP remains capable of 
providing the required flow under accident conditions, and no safety 
margins are reduced.
    This is an administrative change intended to clarify the 
technical specification. There will be no change to the test 
procedure as a result of this clarification

    .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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Richard P. Savio, Acting.

Indiana Michigan Power Company

[Docket Nos. 50-315 and 50-316]

Donald C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of amendment requests: August 11, 1997 (AEP:NRC:1265).
    Description of amendment requests: The proposed amendments would 
revise the Technical Specifications (TS) to allow the filling of the 
emergency core cooling system (ECCS) accumulators without declaring 
ECCS equipment inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

[[Page 68309]]

Criterion 1

    This amendment request does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated because the proposed changes to the T/S represent the 
possibility of an event that has such a low probability as to not be 
considered credible. A calculation was performed that demonstrated 
the CDF resulting from the accumulator fill line operation with all 
of the conditions assumed above is approximately 3 x 
10--10 per year. This is well below the NEI guidelines of 
1 x 10-6 for acceptable risk for a given evolution. 
Therefore, based on probabilistic considerations and the robust 
design of the pumps, we conclude the risk associated with this 
proposed change will not result in a significant increase in the 
probability or consequences of a previously evaluated accident.

Criterion 2

    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
The change does not involve a physical change to the plant, but does 
involve a change in the plant operating configuration. The 
possibility of a LBLOCA [large break loss of coolant accident] 
occurring during the accumulation fill evolution has been evaluated 
and determined to not be credible. Westinghouse has confirmed the 
accumulator fill line was not modeled in the accident analyses due 
to the extremely short duration of the fill operation and the 
extremely small amount of flow that the fill line is capable of 
passing. The overall effect this configuration would have on the 
capability of the SI [safety injection] pump to perform its design 
function, should a LBLOCA occur during the extremely brief window of 
opportunity, is negligible and would not create a new type of 
accident.

Criterion 3

    This proposed change does not involve a significant reduction in 
a margin of safety, as the risk from the postulated sequence of 
events is insignificant. Additionally, engineering evaluation has 
determined that the real response of an SI pump under the postulated 
conditions would not be severe. The rugged construction of the 
pumps, and the design margin built into them, are factors that 
support the engineering judgment that the affected pump would 
continue to operate for some time, at some capacity beyond the 
manufacturer's design limit. As a result of exceeding the limit, the 
pump may experience some cavitation and require additional 
corrective maintenance, but would be expected to deliver a 
significant fraction of its design flow.

    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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, MI 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Richard P. Savio, Acting.

Niagara Mohawk Power Corporation

[Docket No. 50-410]

Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: October 7, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TS) to change the setpoints of 
Surveillance Requirements (SRs) 4.9.6.a, 4.9.6.f, and 4.9.6.g for the 
refueling platform main hoist. Specifically, each refueling platform 
crane or hoist used for handling control rods or fuel assemblies within 
the reactor pressure vessel would be demonstrated operable by:
    a. Demonstrating operation of the overload cutoff on the main hoist 
when the load exceeds 1600 +100/-0 pounds (rather than 1200 +50/-50 
pounds).
    f. Demonstrating operation of the loaded interlock on the main 
hoist when the load exceeds 700 +50/-0 pounds (rather than 485 +50/-50 
pounds).
    g. Demonstrating operation of the redundant loaded interlock on the 
main hoist when the load exceeds 700 +50/-0 pounds (rather than 550 
+50/-50 pounds).
    The proposed amendment, in effect, would authorize replacement of 
the existing triangular refueling platform mast with a round, heavier 
mast (General Electric Model NF-500) which includes an installed 
camera/TV system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change revises the setpoints for three TS SRs based 
on modifications to the refueling platform mast. The new mast is 
essentially a direct replacement for the existing mast, with the 
exception that the new mast is approximately 400 lbs. heavier, which 
directly affects the setpoints. No change in the frequency or manner 
in which the surveillances are performed is proposed. Refueling 
interlocks will continue to function as designed. No changes to the 
methods in which plant systems are operated are required. The same 
design criteria and standards were applied to the new mast, 
including the seismic capability of the refueling platform with the 
heavier mast. Therefore, none of the precursors of previously 
evaluated accidents are affected, and no new failure modes are 
introduced.
    Based on the additional weight of the new mast and camera/TV 
system, the revised GESTAR [General Electric GESTAR II document 
NEDE-24011-P-A-11-U5] criteria for fuel rod damage (more 
conservative threshold level), the use of GE11 [9x9] fuel for the 
bundle drop analysis, the number of damaged fuel rods has increased 
slightly for the potential fuel handling accident. The results of 
this increase were evaluated and dispositioned against the bounding 
calculation to show that the current USAR [updated safety analysis 
report] analysis bounds the revised radiological consequences which 
remain well within the GDC [General Design Criterion] 19 and 
10CFR[part]100 limits. The systems that are available to mitigate 
the consequences of any accident have not been affected and are 
still capable of performing their required functions. Therefore, 
this change will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change revises the setpoints for three TS SRs based 
on installation of a new refueling platform which is heavier than 
the current mast. No change in the frequency or manner in which the 
surveillances are performed has occurred. Refueling interlocks will 
continue to function as designed. No changes to the methods in which 
plant systems are operated are required. The same design criteria 
and standards were applied to the new mast, including the seismic 
capability of the refueling platform with the heavier mast. The 
basic function and operation of the refueling platform is unchanged. 
The uptravel stop and downtravel mechanical cutoff setpoints are not 
being changed and will continue to ensure that adequate water 
shielding is maintained. As such, the change does not introduce any 
new failure modes or conditions that may create a new or different 
kind of accident. Therefore, this change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed change revises three TS SR setpoints based on 
installation of a new refueling platform mast. No change in the 
frequency or manner in which the surveillances are performed has 
occurred. Refueling interlocks will continue to function as 
designed. No changes to the methods in which plant systems are 
operated are required. The same design criteria and standards were 
applied to the new mast, including the seismic capability of the

[[Page 68310]]

refueling platform with the heavier mast. The addition of a camera/
TV system will provide enhanced visibility for fuel handling 
activities and additional assurance that the grapple is oriented 
over the correct fuel bundle.
    The additional weight of the new mast has been evaluated and the 
operability requirements as described in the TS and TS Bases are 
unchanged. The modification and revised setpoints do not change the 
function of the refueling platform main hoist. The revised setpoints 
will continue to assure the lifting capacity of the main hoist will 
not be sufficient to result in damage to core internals or the 
reactor pressure vessel in the event that they are accidentally 
engaged.
    The necessary systems are still available to mitigate any 
potential radiological consequences of the increased number of 
damaged fuel rods. The radiological consequences remain within the 
bounds of the current safety analysis and well below the GDC 19 and 
10CFR[Part]100 limits. Therefore, the change does not involve any 
significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Niagara Mohawk Power Corporation

[Docket No. 50-410]

Nine Mile Point Nuclear Station, Unit 2, Oswego County, New York

    Date of amendment request: October 31, 1997.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) to support installation of the 
General Electric Nuclear Measurement Analysis and Control (NUMAC) Power 
Range Neutron Monitor (PRNM) System. The TS changes apply to Sections 
2.2, ``Limiting Safety System Settings''; 3/4.3.1, ``Reactor Protection 
System Instrumentation'' and its corresponding Bases; and 3/4.3.6, 
``Control Rod Block Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: The NUMAC-PRNM will monitor groups of Local Power Range 
Monitor (LPRM) signals and, together with the Oscillation Power Range 
Monitor (OPRM), initiate a reactor scram upon identifying neutron flux 
oscillations characteristic of a thermal-hydraulic instability. The 
NUMAC-PRNM will replace the existing Average Power Range Monitor (APRM) 
System and will ultimately support the activation of the OPRM. The 
proposed modification is in response to Generic Letter 94-02, ``Long-
Term Solutions and Upgrade of Interim Operating Recommendations for 
Thermal-Hydraulic Instabilities in Boiling Water Reactor.'' Except for 
minor deviations, the proposed TS changes are consistent with General 
Electric Licensing Topical Report (LTR), NEDC-32410P-A, ``Nuclear 
Measurement Analysis and Control Power Range Neutron Monitor (NUMAC-
PRNM) Retrofit Plus Option III Stability Trip Function,'' which was 
approved by the NRC staff September 5, 1995. Changes with respect to 
response time testing requirements would be based on Supplement 1 to 
NEDC-32410P-A, approved by the NRC staff December 26, 1996.
    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 operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    As discussed in NEDC-3241OP-A, the NUMAC-PRNM modification and 
associated changes to the TS involve systems that are intended to 
detect the symptoms of certain events or accidents mitigating 
actions. The worst case failure of the systems involved would be a 
failure to initiate mitigative actions (i.e., scram or rod block), 
but no failure can cause an accident and therefore the probability 
of precursors of any accidents previously evaluated is not 
increased. The NUMAC-PRNM system performs the same operations as the 
existing equipment, reduces the need for tedious operator action 
during normal conditions and allows the operator to focus more on 
overall plant conditions. Automatic self-test and increased operator 
information available with the NUMARC-PRNM system is likely to 
reduce the burden during off-normal conditions as well. The NUMAC-
PRNM system is compatible with the environmental conditions at the 
mounting location (e.g., temperature, humidity, seismic, 
electromagnetic fields) such that system performance will not be 
degraded when compared to the system being replaced. Therefore, the 
proposed change will not result in a significant increase in the 
probability of any accidents previously evaluated.
    The proposed changes to the RPS [reactor protection system] and 
Control Rod Block instrumentation TSs are necessitated by the NUMAC-
PRNM replacement. As discussed in the evaluation, in the 4 APRM 
channel configuration, any two of the four APRM channels and one 2-
out-of-4 voter channel in each RPS trip system are required to 
function for the APRM safety trip function to be accomplished. 
Therefore, the proposed TS change requires that 3 of the 4 APRM 
channels be operable. This assures at least two APRM channels to 
each of the 2-out-of-4 voter channels are available in the event of 
a single APRM channel failure and one APRM is bypassed. Also, the 
proposed TS requires a minimum of two 2-out-of-4 voter channels per 
RPS trip system (i.e., all four voter channels). This assures that 
at least one voter channel per trip system is available even in the 
event of a single voter channel failure. Surveillance testing 
requirements were revised to take advantage of certain features of 
the NUMAC-PRNM (digital) replacement of the existing analog APRM 
system. These advantages included improved accuracy, stability , 
self-testing, reduced drift, and constant time for digital 
processing. Testing of the RPS and Control Rod Block instrumentation 
will continue to be performed as described in the evaluation to 
assure that the reliability and performance of these systems will 
not be adversely affected.
    The proposed NUMAC-PRNM replacement system has been specifically 
designed to assure that the system response times meet the current 
acceptance limits (worst case). As a result, due to statistical 
variations resulting from the sampling and update cycles, the 
response time is typically faster than required in order to assure 
the required response time is always met. The architecture of the 
NUMAC-PRNM system has reduced segmentation compared to the existing 
PRM system. Examples of the reduced segmentation are combining 
previously separate functions, several input channels sharing an 
input board, and a central loop processor for many channels. The 
replacement equipment includes up to 5 LPRM inputs on a single 
module compared to one per module on the current system. Up to 17 
LPRM signals are processed through one preprocessor. The 
recirculation flow signals are processed in the same hardware as the 
LPRM processing. The net effect of these architectural aspects is 
that there are some single failures that cause a greater loss of 
``sub-functionality'' than in the current system. However, other 
architectural and functional aspects have an offsetting effect. 
Redundant power supplies are used so that a single failure of AC 
power has no effect on the overall NUMAC-PRNM system functions while 
still resulting in a half scram, as does the current system. 
Continuous automatic self-test also assures that if a single failure 
does occur, it is much more likely to be detected immediately. The 
net effect is that from a total system level, there is no increased 
risk of loss of critical functionality or reduction in safety 
margins due to the architecture of the replacement system.
    Failure analysis indicates that a software common cause failure 
is not a significant contributor to the unavailability of the NUMAC-
PRNM. However, in spite of that conclusion, means are provided 
within the system to mitigate the effects of such a failure and 
alert an operator. Therefore, such a failure, even if it occurred, 
will not increase

[[Page 68311]]

the consequences of a previously evaluated accident. To reduce the 
likelihood of common cause failures of software controlled 
functions, thorough and careful verification and validation (V&V) 
activities are performed both for the requirements and the 
implementing software design. In addition, the software is designed 
to limit the loading that external systems or equipment can place on 
the system, thus significantly reducing the risk that some abnormal 
dynamic condition external to the system can cause an overload. For 
conservatism, however, despite, these V&V activities, common cause 
failures of software controlled functions due to residual software 
design faults are assumed to occur. Both the software and hardware 
are designed to manage the consequences of such failures. Safety 
outputs are designed to be fail safe by requiring dynamic update of 
output modules or data signals, where failure to update the 
information is detected by simple receiving hardware, which in turn, 
forces a trip. This aspect covers all but rather complex failures 
where the hardware or software executes a portion of the overall 
logic but fails to process some portion of the new information 
(inputs ``freeze'') or some portion of the logic (outputs 
``freeze''). To help reduce the likelihood of complex failures, a 
watchdog timer is used which is updated by a very simple software 
routine that in turn monitors the operational cycle time of all 
tasks in the system. The software design is such that as long as all 
tasks are updating at the design rate, it is likely that software 
controlled functions are executing as intended. Conversely, if any 
task fails too update at the design rate, that is a strong 
indication of at least some unanticipated condition. If such a 
condition occurs, its watchdog timer will not be updated, the 
computer will be restarted, and the outputs will detect an abnormal 
condition and provide an alarm.
    It is very difficult to quantify a software common cause failure 
rate. Analyses for the current system did consider common cause 
failures and assessed them to be at a rate of about 0.3 times the 
random failure rate. The reference analysis uses a field basis for 
the random rates. The analysis for the replacement design uses 
conservative estimates for failure rates of equipment that are 
actually a little higher than those assumed for the current 
equipment. The methodology being applied concludes that the common 
mode failure rate for the replacement system is somewhat higher than 
the current system. However, that is offset by more frequent 
surveillance tests performed by the self-test that result in an 
estimated slightly lower unavailability for the NUMAC-PRNM scram 
function compared to the current PRM system. The USAR, in general, 
considers the failure rate of the function, not that of sub-
components. On that basis, there will not be an increase, due to 
software common cause failure, in the probability of a malfunction 
analyzed in the USAR.I21The NUMAC-PRNM human-machine interface 
design does not introduce an increased burden or constraints on the 
operators' ability to adequately respond to an accident such that 
there would be more severe consequential effects. The information 
available to the operators is the same as with the current system. 
No actions are required by the operator to obtain information 
normally used and equivalent to that available with the current 
equipment. However, the replacement system does provide more direct 
accessible information regarding the condition of the equipment, 
including automatic self-test, which can aid the operator in 
diagnosing unusual situations beyond those defined in the licensing 
basis.
    The replacement system has a significantly lower power 
requirement and is generally smaller, reducing somewhat the seismic 
loading on the panels. The equipment qualification also includes EMI 
[electro magnetic induction] emissions which, combined with the fact 
that the replacement equipment is mounted in its own cabinet 
(replaces all of the current equipment), minimized the likelihood of 
significant impact on other existing equipment.
    The replacement equipment makes increased use of qualified 
optical methods to provide both safety and functional isolation 
between safety-related and nonsafety-related systems. Where fiber 
optic methods cannot be used, the isolation provided is comparable 
to or better than that provided in the current system.
    The net electrical and thermal load for the replacement system 
is less than that for the current system. Accordingly, the 
replacement system had adequate cabinet cooling and no forced 
cooling is required.
    The replacement system meets or exceeds all applicable 
requirements for separation, independence and grounding. The use of 
fiber optic connections between the APRM and RBM [rod block monitor] 
improves the separation and reduces the dependence of the system on 
common grounds. However, for noise rejection, the equipment design 
and manufacturing requirements assure improved grounding of the 
actual equipment.
    No change in wiring or grounding external to the panels 
containing the replacement equipment is necessary for correct 
operation of the replacement equipment.
    NEDC-3241OP-A, Section 3.2.3, discusses different plant 
configurations for recirculation flow channels, including the case 
where plants currently (before implementing the NUMAC PRNM system) 
have four flow channels. Absence of any discussion in the LTR 
related to separation for plants originally having four flow 
channels implies that those plants are expected to meet full 
separation requirements. The LTR includes a further statement that 
``The criterion is to maintain equal or better protection against 
single failures while allowing bypassing of the APRM channel that 
processes the flow signal.''
    The NMPC [Niagara Mohawk Power Corporation] NUMAC PRNM system 
has four recirculation low channels, but the flow input circuits for 
two of the four are not separated from each other outside the PRNM 
panel. As a result, a single failure that causes both of these flow 
signals to go high could, depending on the specific value, cause the 
APRM flow biased trip setpoint in two channels to go to the clamped 
setpoint. If, at the same time, a third channel is bypassed, the 
APRM flow-biased trip setpoint for the APRM system could be non-
conservative. (NOTE: The flow signals are compared to one another. 
Should the flow signals not be within specified limits, an alarm and 
a control rod block would be initiated.)
    Despite the fact that two of the four flow input circuits are 
not separated from each other outside the PRNM panel, the 
replacement system is judged to be adequate with the current field 
routing of flow signals and meets the LTR criteria. This conclusion 
is based on the fact that there is no credible fault in the circuits 
within the duct, in which the flow signals are routed, that can 
damage the other circuits. Also, there is no credible external fault 
that can damage the circuits inside the duct. Therefore, it is 
concluded that the separation between the two flow input circuits is 
adequate to meet the system single failure requirements in that no 
credible single failure will disable the flow inputs to more than 
one APRM channel. Additionally, there are no reload licensing 
transient analyses that take credit for the flow-biased simulated 
thermal power scram setpoint.
    The replacement design has been specifically designed to have 
the same or more conservative ``fail safe'' failure modes as the 
current system. For example, in the case of a single power bus 
failure, the current system loses about one half of the LPRM 
information and an output trip occurs. For the replacement system, 
that failure still results in an output trip, but no LPRM 
information is lost. In the current system, a static failure in 
several areas in the system could result in a ``fail-as-is'' state 
of the outputs. In the replacement system, dynamic coupling starting 
in the main processor and going to the final output virtually 
eliminates ``fail-as-is'' failure modes and replaces them with 
``fail tripped'' modes.
    The replacement system has the same loss of power failure mode 
as the current system relative to the trip outputs and for loss of 
AC [alternating current] power. For loss of DC [direct current] 
power, the replacement system in most cases continues to operate 
normally due to redundancy of the power supplies. Therefore, the 
consequences are no different or improved compared to those 
considered in the USAR.
    Both the current system and the replacement system automatically 
startup on application of power (or re-application). However, the 
replacement system may take slightly longer to reach normal 
operation due to initializing activities. However, no USAR 
evaluations take credit for rapid start of the PRM. Therefore, the 
slightly longer startup time from point of power application is 
bounded by the USAR analysis. Upon application of power, once the 
system is set up for the specific application, it automatically 
returns to those settings upon application of power. All such setup 
parameters are stored in non-volatile memory.
    Human-machine interfaces (HMI) failures in the current system 
could be related to misadjusted settings, incorrect reading of 
meters, and failure to return the equipment to the normal operating 
configuration. There are comparable failure modes for some of these 
in the digital system where an

[[Page 68312]]

erroneous potentiometer adjustment in the current system is 
equivalent to an erroneous digital entry in the replacement system. 
Certain potential ``failure to reconfigure'' errors in the current 
system have no counterpart in the replacement system because any 
``reconfiguration'' is automatically returned to normal by the 
system. Also, since parameters are available for review at any time, 
even if an error such as a digital entry error occurs, it is more 
likely that the error would be almost immediately detected by 
recognition that the displayed value is not the correct one. Failure 
analysis of the current system assumes certain rates of human error. 
The rates for the replacement system will be lower, and hence are 
bounded by the USAR analysis. The NUMAC-PRNM system has been 
approved as an acceptable neutron monitoring replacement by the NRC.
    Therefore, based on the above discussions, the proposed change 
will not result in a significant increase in the consequences of any 
accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    NMPC proposes to replace the existing RPS APRM system with the 
NUMAC-PRNM system and make associated changes to the RPS and Control 
Rod Block TS instrumentation sections. As discussed in NEDC-3241OP-
A, no new system level failure modes are created with the 
replacement system. The NUMAC-PRNM modification and associated 
changes to the TSs involve systems that are intended to detect the 
symptoms of certain events or accidents and initiate mitigating 
actions. The worst case failure of the systems involved would be a 
failure to initiate mitigative actions (i.e., scram or rod block), 
but no failure can cause an accident. This is unchanged from the 
current system. The proposed changes do not modify the basic 
functional requirements of the affected equipment, create any new 
system interfaces or interactions nor create any new system failure 
modes or sequence of events that could lead to an accident. The 
replacement system is more tolerant of degraded power than the 
current system. Software common cause failures can at most cause the 
system to fail to perform its safety function. As with system level 
failures, software failures could fail to initiate actions to 
mitigate the consequences of an accident, but would not cause one. 
Surveillance testing will continue to be performed to assure 
reliability and maintain current performance levels.
    The NUMAC-PRNM system is a digital system with software 
(firmware) control. As such, it has ``central'' processing points 
and software controlled digital processing where the current system 
has analog and discrete component processing. The result is that the 
specific failures of hardware and potentially common cause software 
are different from the current system. Also, automatic self-test 
results in some cases in a direct trip as a result of a hardware 
failure where the current system may have remained ``as is.'' 
However, when these are evaluated at the system level, there are no 
new effects. In general, the USAR assumes simplistic failure modes 
(relays for example) but does not specifically evaluate effects 
added by the NUMAC-PRNM such as self-test detection and automatic 
trip or alarm. The effects of software common cause failures are 
mitigated by hardware design and system architecture. The 
replacement system is fully qualified to operate in its installed 
location and will not affect other equipment. The NUMAC-PRNM system 
has been approved as an acceptable neutron monitoring replacement by 
the NRC. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The proposed modification and associated TS changes will not 
adversely affect the performance characteristics of the RPS and 
Control Rod Block instrumentation nor will it affect the ability of 
the subject instrumentation to perform its intended function. As 
stated in NEDC-3241OP-A, the replacement system has improved channel 
trip accuracy compared to the current system and meets or exceeds 
system requirements assumed in setpoint analysis. Also, the channel 
response time is within acceptable limits, the channel indicated 
accuracy is improved over the current system, and the replacement 
system does not cause a plant parameter for any analyzed event to 
fall outside of acceptable limits. The surveillance testing and 
frequencies proposed will assure reliability of the RPS and Control 
Rod Block instrumentation. In addition, the subject equipment was 
qualified, where appropriate, to assure its intended safety function 
is performed. Therefore, the proposed changes do not involve 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: S. Singh Bajwa.

Pacific Gas and Electric Company

[Docket Nos. 50-275 and 50-323]

Diablo Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo 
County, California

    Date of amendment requests: July 30, 1997.
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant Unit Nos. 1 and 2 to add a limiting condition for operation 
and surveillance requirements for a residual heat removal (RHR) pump 
trip on low refueling water storage tank (RWST) level to TS 3/4.3.2, 
``Engineered Safety Features Actuation System Instrumentation.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This change assures the availability of the refueling water 
storage tank (RWST) low-level trip of the residual heat removal 
(RHR) pumps by establishing limits on the time that a channel can be 
out of service to 72 hours and establishing surveillance criteria to 
verify the operation of the logic. The RHR system is used to respond 
to loss of coolant accidents (LOCAs) and other (e.g., secondary 
side) accidents that could result in initiation of a safety 
injection signal, and is not a precursor to any of these events as 
evaluated in safety analyses. Under accident conditions the RWST 
serves as the source of water for the emergency core cooling system 
(ECCS) pumps and the containment spray pumps. The RWST and the RHR 
pump trip are accident mitigation components and are not precursors 
for any accident evaluated in the safety analyses.
    The existing Technical Specification (TS) would allow one RWST 
level indication channel to be inoperable indefinitely, and has an 
allowed outage time (AOT) for two channels inoperable of up to seven 
days. Additionally, the existing TS does not apply to the RWST low-
level RHR pump trip logic. The new TS provides controls that require 
that all three RWST low-level trip channels be maintained operable 
while the plant is in Modes 1 to 4, and provides for an AOT for one 
channel inoperable for up to 72 hours, if the inoperable channel is 
placed in the cut-out mode within 6 hours. By placing the inoperable 
channel in the cut-out mode, the possibility of a channel failure 
causing an RHR pump failure to start at the onset of an accident is 
precluded even with a single active failure. This assures that the 
consequences of an accident are not increased.
    The change will have no affect on the probability of a physical 
failure of an RHR pump because it only ensures the presence of a 
pump trip signal when required. Therefore, there is no increase in 
the probability of failure of an RHR train to function as designed. 
This change will have no affect on the probability of any other ECCS 
equipment failure as it only affects the presence of a trip signal 
for the RHR pumps.

[[Page 68313]]

    The new TS 3.3.2 item would provide controls that require that 
all three RWST level channels be maintained operable while the plant 
is in operating Modes 1 to 4 (power operation through hot shutdown). 
By maintaining the three channels operable, the RHR pump actuation/
trip logic operability is assured so that the RHR and RWST can in 
all cases perform their intended accident mitigation functions 
following a design basis event as evaluated in the safety analyses.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The RHR system is used to respond to LOCAs and other (e.g., 
secondary side) accidents that could result in initiation of a 
safety injection signal. Under accident conditions the RWST serves 
as the initial source of water for injection by the RHR and other 
ECCS pumps, and is the source of water for the containment spray 
pumps. This change does not affect operation of the systems as it 
relates to their response to accident conditions. It provides 
additional assurance that the RHR pump trip logic will operate as 
designed by establishing administrative controls on the time the 
system is susceptible to a single failure. No new failure modes have 
been introduced.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The relevant margin of safety is based on the RHR pumps starting 
and then automatically stopping at the correct RWST water level. The 
new TS 3.3.2 item provides controls that require all three RWST 
level channels be maintained operable while the plant is in Modes 1 
to 4. By maintaining the three channels operable, the capability of 
the RHR pump actuation/trip logic to survive a single active failure 
is assured. Therefore, the trip logic operability is assured and the 
margin is preserved. This change also provides additional assurances 
that the remaining water in the RWST at the time of switchover is 
consistent with that assumed in the Final Safety Analysis Report and 
Safety Evaluation Reports.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

Pennsylvania Power and Light Company

[Docket Nos. 50-387 and 50-388]

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of amendment request: June 25, 1997.
    Description of amendment request: The amendments would modify the 
Susquehanna Steam Electric Station, Units 1 and 2 Technical 
Specifications to reflect an increase in the secondary containment 
bypass leakage. Specifically, Section 3.6.1.2 is changed to replace the 
leakage of 1.2 scf per hour for any one main steam line drain with 
25.43 scfh for secondary containment bypass leakage from all sources; 
Section 3.6.1.2 is changed to include the Main Steam Line Drain, high-
pressure coolant injection (HPCI) system drain, and reactor core 
isolation cooling (RCIC) system drain leakages as part of the 300 scfh 
leakage requirement; and Section 3/4.6.1.2 is changed to include a 
discussion which related the secondary containment bypass leakage TS to 
the radiological dose analyses.
    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.
    Of the potential accidents described in FSAR [Final Safety 
Analysis Report] Chapters 6 and 15, only a ``Decrease in Reactor 
Coolant Inventory'' as described in FSAR Section 15.6.5 is affected 
by the proposed action. The specific accident of concern is a design 
basis LOCA [loss-of-coolant accident] concurrent with a LOOP [loss-
of-offsite power] which results in RPV [reactor pressure vessel] 
depressurization and failure to recover RPV level above the FW 
[feedwater] spargers. For this accident, the current licensing basis 
offsite and control room dose analyses assume a secondary 
containment bypass leakage rate of 9 scfh and primary containment 
water (called ESF [engineered safety function]) leakage of 5 gpm. 
The current licensing basis analyses do not attribute this leakage 
to any specific pathway.
    The proposed action does not increase the probability of a 
previously analyzed accident in any way. The condition of concern is 
the result of an accident and as such does not contribute to the 
initiation of an accident as analyzed in the FSAR.
    Of concern is whether or not the proposed action significantly 
increases the consequences of an accident as previously evaluated. 
Calculations of off-site dose assuming SCBL [secondary containment 
bypass leakage] of 28 scfh, primary containment water leakage of 20 
gpm, and crediting suppression pool scrubbing show decreases in 
thyroid dose, but slight increases in whole body dose when compared 
with dose calculations performed to support the removal of the MSIV-
LCS [main steam isolation valve-leakage control system]. This result 
is expected because the effect of suppression pool scrubbing is 
factored into the revised licensing basis analysis. Suppression pool 
scrubbing is effective in reducing iodine release but has no assumed 
effect on the removal of noble gases. Since the methodology/
assumptions for scrubbing are acceptable to the NRC [Nuclear 
Regulatory Commission] per the guidance in SRP [Standard Review 
Plan] Section 6.5.5 and the values for decontamination factors are 
conservative, the judgment may be made that considerable margin is 
preserved within the analysis.
    Although the whole body dose with SCBL of 28 scfh and water 
leakage of 20 gpm is increased from the previously approved MSIV-LCS 
dose analysis, the increase is small (about 1 rem at the two hour 
site boundary; less than 0.1 rem 30 day LPZ [low population zone]). 
The total dose including the increase is still well below the 
10CFR100 whole body regulatory limit of 25 rem to which SSES 
[Susquehanna Steam Electric Station] was licensed. No change in 
operating procedures is anticipated. Calculated post accident 
control room thyroid dose decreases as a result of this change, and 
the increase in control room whole body dose is less than 0.05 rem, 
well below the 10CFR50, Appendix A, GDC [General Design Criterion] 
19 dose limits outlined in NUREG-0800. Thus, no appreciable effect 
on operator response will occur as a result of this change.
    The addition of the HPCI and RCIC Steam Line Drains to the Tech 
Spec for MSIV leakage is being performed as a result of the 
modification which eliminated the MSIV Leakage Control System (MSIV 
LCS). At the time this modification was performed, these lines were 
not identified as potential SCBL pathways. However, because leakage 
from the HPCI and RCIC drain lines are part of the same pathway to 
the condenser which is now used by the main steam line drains (MSLD) 
and included in the Technical Specifications, they must be combined 
with the MSIV's and MSLD to be less than 300 scfh. This change only 
affects the accounting of the various drain leakages in the valve 
testing program. The justification for this change is the same 
justification provided in the ITS [Improved Technical Specification] 
submittal (PLA-4488, August 1, 1996) which adds the MSLD to this 
Technical Specification. The test pressure change to allow testing 
at Pa was previously proposed in PLA-4502, September 23, 1996. One 
additional change to delete a footnote related to the removal of the 
MSIV Leakage Control System is

[[Page 68314]]

included because this system has been removed from Susquehanna SES.
    Since the increase in SCBL and primary containment water leakage 
result in only a small increase in the doses previously evaluated by 
the NRC and the other changes do not affect the dose analyses, the 
proposed change does not result in a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Because the FSAR analysis already assumes SCBL and ESF leakage 
occur and the other changes do not affect the type of accident[s] 
that are postulated to occur, the proposed change does not present 
the possibility of an accident of a different type. Additionally, 
the change in dose analysis methodology does not create an accident 
or malfunction of a different type since it only involves the 
analysis of the effects of such accidents or malfunctions.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This question addresses changes in system parameters only. Dose 
consequences are addressed in Section 1 above. The only Technical 
Specification dealing with SCBL is T.S. 3.6.1.2 which requires the 
leakage from any one Main Steam Line Drain (MSLD) Valve to be less 
than or equal to 1.2 scfh when tested at Pa (45.0 psig). As noted 
earlier, the current licensing basis accident dose analysis assumes 
a total of 9 scfh for bypass leakage and 5 gpm for primary 
containment water leakage but does not attribute them to any 
particular source. The proposed action increases the assumed SCBL 
from 9 to 28 scfh and water leakage from 5 gpm to 20 gpm. These 
leakage rates are insignificant in terms of SGTS [standby gas 
treatment system] flows or water loss from ECCS systems. These 
leakage rates do not affect building temperatures or pressures so 
that they become closer to acceptance limits. Likewise, no other 
system parameter values become closer to limits as a result of these 
changes in leakage. Consequently, the existing margin of safety 
between the licensing basis analysis and system parameter acceptance 
limits is not reduced. The changes to the HPCI, RCIC, and main steam 
line drain leakage only affect the accounting for the various 
leakages in the leakage testing program. The deletion of the 
footnote is administrative because the MSIV Leakage Control System 
has been removed from the Susquehanna SES. The change in test 
pressure was previously evaluated in PLA-4502, September 23, 1996. 
Thus, no decrease in margin of safety results.

    The NRC staff has reviewed the licensee's analysis and notes that a 
discussion of the administrative change to delete a footnote in Section 
3.6.1.2 is in the third section of the no significant hazards 
consideration. The staff finds that this administrative change also 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated and does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated. Based on this staff review, it appears 
that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, 
the NRC staff proposes to determine that the amendment request involves 
no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Pennsylvania Power and Light Company

[Docket No. 50-387]

Susquehanna Steam Electric Station, Unit 1, Luzerne County, 
Pennsylvania

    Date of amendment request: August 26, 1997.
    Description of amendment request: The amendment would modify the 
Susquehanna Steam Electric Station, Unit 1 Technical Specifications to 
change the definitions in Section 1.0 to make them applicable to 
ATRIUM-10 fuel (reflecting the new design), to include the Unit 1 Cycle 
11 flow dependent minimum critical power ratio (MCPR) Safety Limits in 
Sections 2.1.2 and 3.4.1.1.2, to change Section 5.3.1 to reflect the 
ATRIUM-10 design, and to include Siemens Power Corporation methodology 
topical reports and references to the methodology in Section 6.9.3.2.
    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 applicable sections of the FSAR [Final Safety Analysis 
Report] are Chapters 5, 6.3, 9, and 15 of the FSAR. Chapter 5 
discusses the results of the ASME [American Society of Mechanical 
Engineers] overpressure analysis for the reactor pressure boundary. 
Chapter 6.3 discusses the LOCA [loss-of-coolant accident]. Chapter 9 
discusses fuel storage and handling. Chapter 15 describes the 
transient and accident analyses, a majority of which have been 
dispositioned to be non-limiting. A discussion of the impact of the 
Technical Specification changes is provided below.
    The change to Definitions 1.2 and 1.3 makes the definitions 
applicable to ATRIUM TM-10. There are no effects on 
safety functions from this change.
    A cycle specific MCPR Safety Limit analysis was performed for 
PP&L [Pennsylvania Power and Light Company] by SPC [Siemien Power 
Corporation]. This analysis used NRC [Nuclear Regulatory Commission] 
approved methods described in Technical Specification Reference 13 
(ANF-524(P)(A), Revision 2 and Supplement 1 Revision 2), as modified 
by EMF-97-010(P), Rev. 1. The SAFETY LIMIT MCPR calculation 
statistically combines uncertainties on feedwater flow, feedwater 
temperature, core flow, core pressure, core power distribution, and 
the uncertainty in the Critical Power Correlation. The SPC analysis 
used cycle specific power distributions and calculated MCPR values 
such that at least 99.9% of the fuel rods are expected to avoid 
boiling transition during normal operation or anticipated 
operational occurrences. The SAFETY LIMIT MCPRs are specified as a 
function of core flow. The resulting two-loop and single-loop values 
(Technical Specification Sections 2.1.2 and 3.4.1.1.2) are included 
in the proposed change. Thus, the cladding integrity and its ability 
to contain fission products are not adversely affected.
    The MCPR methodology for ATRIUM TM-10 fuel (SPC 
report EMF-97-010(P), Rev. 1), included in the revised Technical 
Specifications via reference (Section 6.9.3.2) and previously 
approved by the NRC for Unit 2 Cycle 9, describes conservative 
methods for developing the MCPR Safety Limits and Operating Limits 
for the U1C11 reload of ATRIUM TM-10 fuel in the 
Susquehanna Steam Electric Station. This methodology conservatively 
accounts for a flow dependence in the ATRIUM TM-10 
critical power test data as well as an increased correlation 
uncertainty for high local peaking factor rods. The results of using 
this methodology are core flow dependent MCPR Safety Limits plus 
conservative MCPR Operating Limits for Unit 1 Cycle 11. The 
resulting MCPR Safety Limits and Operating Limits will continue to 
assure that at least 99.9% of the fuel rods are expected to avoid 
boiling transition during normal operation or anticipated 
operational occurrences. Thus, the cladding integrity and its 
ability to contain fission products are not adversely affected. The 
proposed change in MCPR methodology does not physically affect the 
plant or its systems.
    Using the approach discussed in EMF-97-010(P), Rev. 1, analyses 
of the Pump Seizure accident with the new MCPR methodology (SPC 
report EMF-97-010(P), Rev. 1) will demonstrate that the NRC 
acceptance criterion (i.e., small fraction of 10CFR100 dose limits) 
is met.
    The change to the Design Features (Section 5.3) increases the 
maximum allowable lattice average enrichment. Analyses have 
demonstrated that the ATRIUM TM-10 fuel will remain 
subcritical (k-effective < 0.95) in both the spent fuel pool and the 
new fuel vault. Thus, the change to maximum allowable lattice 
average enrichment has no impact on safety functions. The 
description

[[Page 68315]]

of a fuel assembly (Section 5.3) is also revised to reflect the 
ATRIUM TM-10 central water channel, and reference to an 
active fuel length of 150 inches was deleted. This change reflects 
the physical characteristics of the ATRIUM TM-10 fuel and 
has no impact on the probability or consequences of an event.
    Included in the revised Technical Specifications via reference 
(Section 6.9.3.2) are additional NRC approved methodology reports. 
The NRC approved topical reports contain methodology which is used 
to assure safe operation of Unit 1 with ATRIUM TM-10 
fuel. These methodologies assure that the core meets appropriate 
margins of safety for all expected plant operational conditions 
ranging from refueling and cold shutdown of the reactor through 
power operation. Thus, the results obtained from the analyses will 
provide assurance that the reactor will perform its design safety 
function during normal operation and design basis events.
    The BASES changes for Section 2.1.1 (THERMAL POWER, Low Pressure 
or Low Flow) reflect that the Safety Limit is valid for both 9x9-2 
and ATRIUM TM-10. BASES for Section 2.1.2 were changed to 
refer to Section 6.9.3.2 for applicable references.
    Therefore, the proposed action does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The changes to the Unit 1 Technical Specifications (Definitions, 
MCPR safety limits, Design Features, and inclusion of methodology 
references) to allow use of ATRIUM TM-10 fuel do not 
require any physical plant modifications, physically affect any 
plant components, or entail significant changes in plant operation. 
Thus, the proposed change does not create the possibility of a 
previously unevaluated operator error or a new single failure. The 
consequences of transients and accidents will remain within the 
criteria approved by the NRC. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The applicable Technical Specification Sections include 1.0, 
2.0, 3/4.4, 5.3, and 6.9.3.2.
    The changes to the Unit 1 Technical Specifications discussed in 
Item 1 above do not require any physical plant modifications, 
physically affect any plant components, or entail significant 
changes in plant operation. Therefore, the proposed change will not 
jeopardize or degrade the function or operation of any plant system 
or component governed by Technical Specifications. The consequences 
of transients and accidents will remain within the criteria approved 
by the NRC. The proposed MCPR Safety Limits and the NRC approved 
methods and revised MCPR methodology detailed in the references 
added to Section 6.9.3.2 maintain an equivalent margin of safety as 
defined in the BASES of the applicable Technical Specification 
sections.
    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz.

Southern California Edison Company, et al.

[Docket Nos. 50-361 and 50-362]

San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego 
County, California

    Date of amendment requests: June 18, 1997.
    Description of amendment requests: The licensee proposes to revise 
Technical Specification (TS) 3.8.1, ``AC Sources--Operating'' and 
applicable Bases. This change will more clearly reflect safety analysis 
and testing conditions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change would revise Technical Specification (TS) TS 
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SRs) 
3.8.1.1, 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 
3.8.1.14, 3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and 3.8.1.20 and 
applicable Bases to more clearly reflect surveillance test 
conditions and system design requirements. Changes to the SRs 
include more restrictive voltage and frequency acceptability limits. 
The new requirements reflect the system design requirements in order 
to ensure Class 1E system operability, meet the requirements of the 
safety analysis, and to agree with the existing test surveillances.
    In addition, the discussion regarding design basis reactive 
power loading is eliminated since this cannot be readily controlled 
during testing.
    Operation of the facility would remain unchanged as a result of 
the proposed change and no assumptions or results of any accident 
analyses are affected. Therefore, the proposed change will not 
involve a 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 
evaluated.
    The proposed change would revise Technical Specification (TS) TS 
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SRs) 
3.8.1.1, 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 
3.8.1.14, 3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and 3.8.1.20 and 
applicable Bases to more clearly reflect surveillance test 
conditions and system design requirements.
    Operation of the facility would remain unchanged as a result of 
the proposed change. Therefore, the proposed change will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change would revise Technical Specification (TS) TS 
3.8.1, ``AC Sources--Operating,'' Surveillance Requirement (SRs) 
3.8.1.1, 3.8.1.2, 3.8.1.7, 3.8.1.10, 3.8.1.11, 3.8.1.12, 3.8.1.13, 
3.8.1.14, 3.8.1.15, 3.8.1.16, 3.8.1.17, 3.8.1.19, and 3.8.1.20 and 
applicable Bases to more clearly reflect surveillance test 
conditions and system design requirements. Changes to the SRs 
include more restrictive voltage and frequency acceptability limits. 
The new requirements reflect the system design requirements in order 
to ensure Class 1E system operability, meet the requirements of the 
safety analysis, and to agree with the existing test surveillances.
    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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.
    Attorney for licensee: T. E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern California Edison Company, et al.

[Docket Nos. 50-361 and 50-362]

San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego 
County, California

    Date of amendment requests: November 14, 1997 (supersedes February 
1, 1994, amendment request).

[[Page 68316]]

    Description of amendment requests: The licensee proposes to revise 
the licensing basis as described in the Updated Final Safety Analysis 
Report Section 3.5, ``Missile Protection,'' to allow the use of NUREG-
0800, ``Standard Review Plan'' methodology in evaluating tornado-
generated missiles. In particular, a probability based criteria is 
proposed to evaluate missile barrier requirements consistent with 
Section 3.5.1.4 of NUREG-0800.
    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.
    NUREG-0800, Standard Review Plan (SRP) Section 3.5.1.4, Revision 
0 and Section 3.5.1.5 Revision 1 provide a conservatively acceptable 
probability threshold for safety due to damage caused by postulated 
missile strikes. Section 3.5.1.4, Revision 0 uses 10-7 
per year for a tornado-generated missile strike, and Section 3.5.1.5 
Revision 1 uses 10-7 per year for exceeding 10 CFR Part 
100 limits.
    The proposed criteria of probability of damage to critical 
exposed equipment (as defined in San Onofre Updated Final Safety 
Analysis Report proposed Table 3.5-13) of 10-7 per year 
per unit is consistent with this guidance.
    The probability of damage to exposed critical components due to 
a postulated missile strike of 10-7 is so small as to be 
negligible. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This amendment request establishes a conservative criteria for 
tornado-generated missiles consistent with the SRP guidance and will 
not create a new or different kind of accident from any accident 
that has been previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This proposed change is consistent with the methodology and 
acceptance criteria of the SRP, and the SRP criteria ensures that 
there will be no undue risk to the health and safety of the public. 
Therefore, there will be no significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.
    Attorney for licensee: T.E. Oubre, Esquire, Southern California 
Edison Company, P. O. Box 800, Rosemead, California 91770.
    NRC Project Director: William H. Bateman.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia

Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

[Docket Nos. 50-424 and 50-425]

    Date of amendments request: August 8, 1997, as supplemented October 
10, 1997. This application and supplement supersedes the October 4, 
1996, application, noticed in the Federal Register on November 19, 1996 
(61 FR 58903), in its entirety.
    Description of amendments request: The proposed amendments would 
change the Technical Specifications to credit soluble boron in the 
spent fuel pool for maintenance of subcriticality and increase the 
allowable fuel enrichment to 5.0 percent U-235 as follows:

1. Revisions to the Table of Contents

    The Table of Contents would be revised to include two additional 
Technical Specifications 3.7.17, ``Fuel Storage Pool Boron 
Concentration,'' and 3.7.18, ``Fuel Assembly Storage in the Fuel 
Storage Pool'' and add Figures 3.7.18-1, 3.7.18-2, and 4.3.1-1 
through 4.3.1-9 describing burnup credit, checkerboard 
configurations and interface requirements. These changes would be 
added to support crediting soluble boron in the fuel storage pool 
criticality analyses.

2. Addition of Technical Specifications 3.7.17 and 3.7.18

    Technical Specifications 3.7.17, ``Fuel Storage Pool Boron 
Concentration,'' and 3.7.18, ``Fuel Assembly Storage in the Fuel 
Storage Pool,'' would be added to credit soluble boron in the fuel 
storage pool criticality analyses, and specify acceptable 
enrichment-burnup combinations for storage of fuel in the fuel 
storage pool.

3. Revision to Technical Specification 4.3.1.1

    Design Features Section 4.3.1.1 would be revised to reflect the 
increased maximum enrichment assumed in the fuel storage pool 
criticality analyses, add a requirement to maintain Keff 
less than 1.0 when fully flooded with unborated water, change the 
0.95 Keff requirement from ``if fully flooded with 
unborated water'' to ``when fully flooded with water borated to 450 
ppm (Unit 1) or 500 ppm (Unit 2),'' and to add a reference to 
Specification 3.7.18 for allowable enrichment-burnup combinations. 
Requirements for fuel that do not meet the requirements of 
Specification 3.7.18, would also be added to Section 4.3.1.1, 
including Figures 4.3.1-1 through 4.3.1-9 depicting acceptable 
enrichment-burnup requirements and checkerboard configurations.

4. Revisions to the Table of Contents (Bases)

    The Table of Contents would be revised to include two additional 
Technical Specification Bases Sections B 3.7.17 ``Fuel Storage Pool 
Boron Concentration'' and B 3.7.18 ``Fuel Assembly Storage in the 
Fuel Storage Pool.''

5. Addition of Bases for Technical Specifications 3.7.17 and 3.7.18

    Two additional Technical Specification Bases Sections B 3.7.17, 
``Fuel Storage Pool Boron Concentration'' and B 3.7.18, ``Fuel 
Assembly Storage in the Fuel Storage Pool'' would be added to credit 
soluble boron in the fuel storage pool criticality analyses.

    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 radiological consequences of 5.0 weight percent U-235 fuel 
on accidents previously evaluated in the Vogtle FSAR [Final Safety 
Analysis Report] are not significant. Increasing the enrichment up 
to and including 5.0 weight percent U-235 has minor effects on the 
radiological source terms and subsequently the potential releases 
both normal and accidental are not significantly affected. 
Evaluations performed (WCAP-12610-P-A, Reference 5 [of the 
licensee's application]) considered the source term, gap fraction, 
and the accident doses for a maximum fuel enrichment of 5.0 weight 
percent U-235. It was concluded that operating with and storing fuel 
with 5.0 weight percent U-235 enrichment may result in minor changes 
in the normal annual releases of long half-life fission products 
that are not significant. Also, the radiological consequences of 
accidents are minimally affected due to the very small changes in 
the core inventory and the fact that the currently assumed gap 
fractions remain bounding.
    The use of the slightly higher enrichment for VEGP [Vogtle 
Electric Generating Plant] fuel will not result in burnups in excess 
of those currently allowed for VEGP. The cycle design methods and 
limits will remain the same as are currently licensed. Therefore, 
the use of fuel with the higher enrichment will not result in 
conditions outside those currently allowed for VEGP.
    There is no increase in the probability of a fuel assembly drop 
accident in the fuel storage pool when considering the presence of 
soluble boron in the pool water for criticality control. The 
handling of the fuel assemblies in the fuel storage pool has always 
been performed in borated water.
    Fuel assembly placement will be controlled pursuant to approved 
fuel

[[Page 68317]]

handling procedures and will be in accordance with the spent fuel 
rack storage configuration limitations in the Technical 
Specifications. The consequences of a misplaced assembly have been 
included in the analysis supporting this revision to the Technical 
Specifications.
    There is no increase in the consequences of the accidental 
misloading of a fuel assembly into the fuel storage pool racks 
because criticality analyses demonstrate that the pool will remain 
subcritical following an accidental misloading of an assembly. There 
are no credible dilution events that reduce the subcriticality 
margin below the 5% margin recommended in NRC guidance (references 
1, 2, and 3 [of the licensee's application]). Even if the fuel 
storage pool were diluted to a boron concentration of 0 ppm the No 
Soluble Boron 95/95 analysis demonstrates that the pool will remain 
subcritical. The proposed Technical Specifications limitations will 
ensure that an adequate fuel storage pool boron concentration will 
be maintained.
    There is no increase in the probability of the loss of normal 
cooling to the fuel storage pool water due to the presence of 
soluble boron in the pool water for subcriticality control, because 
a concentration of soluble boron similar to the proposed limit has 
always been maintained in the fuel storage pool water.
    The loss of normal cooling to the fuel storage pool will cause 
an increase in the temperature of the fuel storage pool water. This 
will cause a decrease in water density which would normally result 
in an addition of negative reactivity. However, since Boraflex is 
not considered to be present, and the fuel storage pool water has a 
high concentration of boron, a density decrease causes a positive 
reactivity addition. The amount of soluble boron required to offset 
this postulated accident was evaluated for the allowed storage 
configurations. The amount of soluble boron necessary to mitigate 
these accidents and ensure that the Keff will be 
maintained less than or equal to 0.95 has been included in the fuel 
storage pool boron concentration. Because adequate soluble boron 
will be maintained in the pool water, the consequences of a loss of 
normal cooling to the fuel storage pool will not be increased.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The potential for criticality accidents in the fuel storage pool 
are not new or different types of concerns. The potential 
criticality accidents have been reanalyzed in the Criticality 
Analysis report (Enclosure 5 [of the licensee's application]) to 
demonstrate that the pool remains subcritical.
    Soluble boron has been maintained in the fuel storage pool water 
since its initial operation. The possibility of a fuel storage pool 
dilution is not affected by the proposed change to the Technical 
Specifications. Therefore, the implementation of Technical 
Specification controls for the soluble boron will not create the 
possibility of a new or different kind of accidental pool dilution.
    With credit for soluble boron now a major factor in controlling 
subcriticality, an evaluation of fuel storage pool dilution events 
was completed. The results of the evaluation concluded that no 
credible events would result in a reduction of the criticality 
margin below the 5% margin recommended by the NRC. In addition, the 
No Soluble Boron 95/95 criticality analysis assures that dilution to 
0 ppm will not result in criticality.
    Proposed Technical Specifications 3.7.17, 3.7.18 and 4.3.1.1 
which ensure the maintenance of the fuel storage pool boron 
concentration and storage configuration, do not represent new 
concepts. The actual boron concentration in the fuel storage pool 
has been maintained at a higher value than the proposed limits for 
the Unit 1 and 2 fuel storage pools for refueling purposes. The 
criticality analysis (Enclosure 5 [of the licensee's application]) 
determined that a boron concentration of 450 ppm (Unit 1) and 500 
ppm (Unit 2) results in a Keff [less than or equal to] 
0.95.
    There is no significant change in plant configuration, equipment 
design, or usage of plant equipment. The safety analysis for 
dilution accidents has been expanded; however, the criticality 
analyses assure that the pool will remain subcritical with no credit 
for soluble boron. Therefore, the proposed changes will not create 
the possibility of a new or different kind of accident.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    Proposed Technical Specifications 3.7.17, 3.7.18, and 4.3.1.1 
and the associated fuel storage pool boron concentration and storage 
requirements will provide adequate margin to assure that the fuel 
storage array will always remain subcritical by the 5% margin 
recommended by the NRC. Those limits are based on the criticality 
analysis (Enclosure 5 [of the licensee's application]) performed in 
accordance with the Westinghouse fuel storage rack criticality 
analysis methodology described in Reference 4 [of the licensee's 
application].
    While the criticality analysis utilized credit for soluble 
boron, the storage configurations have been defined using 
Keff calculations to ensure that the spent fuel rack 
Keff will be less than 1.0 with no soluble boron.
    Soluble boron credit is used to offset off-normal conditions 
(such as a misplaced assembly) and to provide subcritical margin 
such that the fuel storage pool Keff is maintained less 
than or equal to 0.95.
    The combination of the No Soluble Boron 95/95 Keff 
calculation which shows that the Keff will remain less 
than 1.0 when flooded with unborated water and the unavailability of 
the large volumes of water which are necessary to dilute the fuel 
storage pool to a Keff of > 0.95, provide a level of 
safety comparable to the conservative criticality analysis 
methodology required by References 1, 2, and 3 [of the licensee's 
application].
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of 
Georgia, City of Dalton, Georgia

Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

[Docket Nos. 50-424 and 50-425]

    Date of amendment request: September 4, 1997, as supplemented 
November 20, 1997.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) to change the capacity of the 
Vogtle Unit 1 spent fuel storage pool from 288 to 1476 assemblies, and 
would revise the design features description to reflect the criticality 
analyses and storage cell spacing. Specifically, the changes would be 
as follows:

    1. Figure 3.7.18-1 would be replaced with a revised figure based 
on the criticality analyses for the Unit 1 racks containing boral.
    2. The criticality information for Unit 2 would be placed 
unchanged into Section 4.3.1.2, and Section 4.3.1.1. would be 
revised to address Unit 1.
    3. Design Features Section 4.3.1.1.c would be revised to 
indicate 600 ppm as the required amount of soluble born to maintain 
Keff less than or equal to 0.95.
    4. Design Features Section 4.3.1.1.d would be revised to include 
the reference Keff that is equivalent to the combination 
of burnup and initial enrichment defined by Figure 3.7.18-1.
    5. Design Features Section 4.3.1.1.e would be revised to 
indicate that fuel assemblies with up to 5 weight percent U-235 may 
be stored in 3-out-of-4 checkerboard storage configurations; delete 
Figure 4.3.1-1; eliminate the reference to 2-out-of-4 storage for 
the Unit 1 pool and include the reference K acceptable for all cell 
storage in the Unit 1 fuel storage racks.
    6. Design Features Section 4.3.1.1.f would be revised to include 
the pitch of the Unit 1 fuel storage racks.

[[Page 68318]]

    7. Design Features Section 4.3.3 would be revised to indicate 
the Unit 1 fuel storage pool capacity of 1476 fuel assemblies.
    8. The titles on Figures 4.3.1-4, 4.3.1-6, and 4.3.1-7 would be 
revised to reflect the elimination of 2-out-of-4 storage 
configuration requirements for the Unit 1 fuel storage pool.
    Changes to the TS Bases are also proposed.

    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 analyses methodologies are the same as previously 
approved for use by the NRC. The results of the analyses resulted in 
fuel pool boron concentrations, and fuel assembly storage 
limitations that are similar to those already submitted to the NRC. 
The increased number of fuel assemblies will remain less than the 
number previously accepted by the NRC for storage in VEGP [Vogtle 
Electric Generating Plant] Unit 2, which has a similarly designed 
and constructed facility, with the exception of the number of fuel 
storage locations.
    Therefore, based on the conclusions of the above analysis, the 
proposed changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The effects of accidents that could affect the fuel were 
analyzed for the fuel storage racks, however the types of accidents 
have not changed. The fuel to be stored in the Unit 1 pool is 
expected to meet the all cell storage requirements. The racks will 
be placed in the Unit 1 pool without lifting any loads over spent 
fuel. After installation of the new racks, the Unit 1 pool will have 
1476 storage locations which is well within the 2098 locations that 
the pool and structure is capable of storing, based on its 
similarity to the Unit 2 pool.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident.
    3. The changes to the technical specifications are necessary to 
incorporate the parameters resulting from the criticality analyses. 
The criticality analyses were performed using methods and criteria 
previously accepted by the NRC. The requirements are similar to the 
previously submitted requirements. The margins of safety provided by 
the previous technical specifications are not significantly affected 
because the new racks are based on the same acceptance values. The 
larger number of fuel assemblies to be stored in the Unit 1 pool 
remains well within the capability of the pool.
    Therefore, the proposed changes in this license amendment will 
not result in a significant reduction in the plant's margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Project Director: Herbert N. Berkow.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia

Vogtle Electric Generating Plant, Units 1 and 2, Burke County, Georgia

[Docket Nos. 50-424 and 50-425]

    Date of amendment request: November 20, 1997.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to provide for the following 
with regard to the Reactor Trip System (RTS) and Engineered Safety 
Feature Actuation System (ESFAS) instrumentation trip setpoints:

    1. The inequalities as they are applied to the Trip Setpoint 
column of Tables 3.3.1-1 and 3.3.2-1 would be deleted, and the 
column heading would be changed from ``Trip Setpoint'' to ``Nominal 
Trip Setpoint.''
    2. A footnote would be added to the new ``Nominal Trip 
Setpoint'' column of Tables 3.3.1-1 and 3.3.2-1 that would allow the 
trip setpoints to be set more conservative than the nominal value as 
necessary to respond to plant conditions.
    3. The Allowable Value for Table 3.3.1-1, Function 14.b, Turbine 
Trip--Turbine Stop Valve Closure, would be revised from ``[greater 
than or equal to] 96.7% open'' to ``[greater than or equal to] 90% 
open.''
    4. Footnotes l and m of Table 3.3.1-1 would be revised to refer 
to the ``Nominal Trip Setpoint'' and delete the inequalities applied 
to the trip setpoints.
    5. A superscript ``(a)'' would be deleted from the heading of 
the ``Trip Setpoint'' column on page 6 of 8 of Table 3.3.1-1.
    6. Notes 1 and 2 to Table 3.3.1-1, Overtemperature T 
and Overpower T, respectively, would be revised to refer to 
the ``Nominal Trip Setpoint.'' In addition, these notes will be 
revised to delete the inequalities from the values for the constants 
K1 through K6 (except for K5 
[greater than or equal to] 0 for decreasing temperature and 
K6 = 0 for T [less than or equal to] T''), and for T', 
T'', and P'.
    7. The inequality applied to the ESFAS Allowable Value for Steam 
Line Pressure--Low (Table 3.3.2-1, Function 1.e) would be changed 
from ``[less than or equal to]'' to ``[greater than or equal to].''
    Associated changes to the TS Bases are also proposed.

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    No. The proposed changes affect only the presentation of the 
trip set points for the RTS and ESFAS in the VEGP [Vogtle Electric 
Generating Plant] Units 1 and 2 TS. The calibration of the channels 
whose setpoints are specified in the TS will continue to be 
performed in a manner consistent with the setpoint methodology 
described in WCAP-11269 Rev. 1. There will be no adverse effect on 
the ability of those channels to perform their safety functions as 
assumed in the safety analyses. Since there will be no adverse 
affect on the trip setpoints or the instrumentation associated with 
those trip setpoints, there will be no increase in the probability 
of any accident previously evaluated. Similarly, since the ability 
of the instrumentation to perform its safety function is not 
adversely affected, there will [be] no increase in the consequences 
of any accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    No. The proposed change affects only the presentation of trip 
setpoint requirements in the TS. Plant operation will not be 
changed, and the response of safety related equipment as assumed in 
the accident analyses will not be adversely affected. Therefore, the 
proposed change does not involve a new or different kind of accident 
than any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety [?]
    No. As described above, the RTS and ESFAS instrumentation will 
remain capable of performing its safety function as assumed in the 
accident analyses. The treatment of trip setpoints as nominal values 
is consistent with the methodology used to establish those 
setpoints. As such, margin is not affected by the proposed change.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia.
    NRC Project Director: Herbert N. Berkow.

[[Page 68319]]

Vermont Yankee Nuclear Power Corporation

[Docket No. 50-271]

Vermont Yankee Nuclear Power Station, Windham County, Vermont

    Date of amendment request: October 10, 1997, as supplemented 
October 31, 1997.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to reflect the installation of 
a generator no-load disconnect to facilitate use of the main step-up 
transformer backfeed as the delayed access offsite power source. Also, 
the amendment would revise existing limiting conditions for operation 
and required action statements for operation with inoperable ac power 
sources to be consistent with current guidance.
    Specifically, the changes proposed are: (1) TS Limiting Conditions 
for Operation Section--Normal Operation, 3.10.A.4 (2) TS Limiting 
Conditions for Operation Section--Operation with Inoperable Components, 
3.10.B.3, (3) TS Surveillance Requirements--Normal Operation, 4.10.A.4, 
(4) TS Surveillance Requirements--Operation with Inoperable Components, 
Section 4.10.B.3, (5) Bases Section 3.10.A, (6) Bases Section 3.10.B, 
(7) Bases Section 4.10.A, and (8) Bases Section 4.10.B
    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 amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment removes credit for the Vernon Tie, 
Vermont Yankee's station blackout source of power, from the 
Technical Specifications and reflects the installation of the 
generator no load disconnect as part of the backfeed. Neither the 
backfeed through the main transformers nor the Vernon Tie are 
accident initiators; therefore, the change does not involve a 
significant increase in the probability of an accident previously 
evaluated. The change does not affect the capability, availability, 
maintenance or operation of the Vernon Tie. Installation of the 
generator no load disconnect switch is being implemented by a design 
change in order to enhance plant safety by reducing time necessary 
to establish the backfeed through the main transformer. A separate 
10 CFR 50.59 evaluation is being prepared to document that the 
modification does not create an unreviewed safety question.
    The proposed amendment also clarifies the allowable out of 
service times, and required actions; and updates surveillance 
requirements for the immediate and delayed access offsite power 
sources. These changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Modification of a technical specification out of service time and 
required action cannot affect the probability or consequences of an 
accident. Enhancing surveillance requirements to provide assurance 
that the backfeed can be achieved when required and to provide 
assurance that remaining power sources are available when an offsite 
source is unavailable improves plant safety and does not increase 
the probability or consequences of an accident.
    Therefore, the change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed amendment removes the Vernon Tie, Vermont Yankee's 
station blackout source of power, as a delayed access source from 
the Technical Specifications and reflects the improvements to the 
main transformer backfeed delayed access source because of 
installation of the generator no load disconnect. Neither the 
removal of the Vernon Tie from Technical Specifications nor the 
improvements to the delayed access power source (backfeed) can 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The proposed amendment also clarifies the allowed outage times, 
and action statements; and updates surveillance requirements for the 
immediate and delayed access offsite power sources. A clarification 
of a technical specification out of service time and required action 
cannot create a new or different kind of accident from any accident 
previously evaluated. Enhancing surveillance requirements to provide 
assurance that the backfeed can be achieved when required and to 
provide assurance that remaining power sources are available when an 
offsite source is unavailable improves plant safety and cannot 
create a new or different kind of accident from any accident 
previously evaluated.
    Therefore, this change would not create the possibility of a 
different type of accident than previously evaluated.
    (3) The proposed amendment will not involve a significant 
reduction in a margin of safety.
    The proposed amendment removes the Vernon Tie, Vermont Yankee's 
station blackout source of power, as a delayed source of offsite 
power from the Technical Specifications and reflects the 
improvements to the main transformer backfeed delayed access source 
because of installation of the generator no load disconnect. No 
existing safety margins are adversely affected. The backfeed is 
modified so that it may be established in sufficient time to 
``assure that specified acceptable fuel design limits and design 
conditions of the reactor coolant pressure boundary are not 
exceeded''. Vernon Tie will not be affected by the modification and 
remain available as a station blackout source; thus this change does 
not involve a significant reduction in the margin of safety.
    The proposed amendment also clarifies the allowed out of service 
times, and required actions; and updates surveillance requirements 
for the immediate and delayed access offsite power sources. A 
clarification of a technical specification out of service time and 
required action does not involve a significant reduction in the 
margin of safety in the Technical Specifications. Enhancing 
surveillance requirements to provide assurance that the backfeed can 
be achieved when required and to provide assurance that remaining 
power sources are available when an offsite source is unavailable 
improves plant safety and does not involve a significant reduction 
in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: Ronald Eaton, Acting Director.

Vermont Yankee Nuclear Power Corporation

[Docket No. 50-271]

Vermont Yankee Nuclear Power Station, Windham County, Vermont

    Date of amendment request: November 20, 1997.
    Description of amendment request: The proposed amendment would 
revise the existing requirements for the Auxiliary Electrical Power 
Systems as identified in Technical Specifications (TSs) 3/4.10.A and TS 
3.10.A.2.b. The specific changes are:
    (1) The requirements in TS 3.10.A.2.b. are revised to omit the 
allowance for Spare Charger AB to substitute for either Charger A or B.
    (2) The Bases in TS 3.10.A. are revised to omit the statements that 
justify Spare Charger AB to substitute for either Charger A or Charger 
B.
    The proposed changes provide more limiting requirements for 
operation with the standby battery charger in service.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

[[Page 68320]]

    Neither batteries, nor their chargers, are considered to be an 
initiator of any previously analyzed accident. Therefore, this 
change will not significantly increase the probability of any 
previously analyzed accident.
    At least one Battery System is required to be available to 
mitigate the consequences of a Design Basis Accident. This change 
removes an allowance which places the unit in a more vulnerable 
condition through the unrestricted use of the spare battery charger. 
Since this change limits such a condition, it maintains the 
assumptions of the safety analysis, and therefore, will not 
significantly increase the consequences of any previously analyzed 
accident.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) nor is operation of the currently installed equipment 
changed. The change will, however, limit a currently allowed 
configuration with the spare charger and is more conservative. Thus, 
this change will not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety.
    The proposed change continues to provide the previous margin of 
safety regarding the capability to withstand a single failure. At 
least one Battery System will continue to be available to provide 
the required safety function. The change will limit a currently 
allowed configuration with the spare charger and is thus more 
conservative. Therefore, this change will not significantly reduce a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Project Director: Ronald Eaton, Acting Director.

Vermont Electric and Power Company

[Docket Nos. 50-280 and 50-281]

Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: November 5, 1997.
    Description of amendment request: The proposed change to Technical 
Specifications 5.3 and 5.4 would reflect an increase in the maximum 
permitted fuel enrichment to 4.3 weight percent U235 from 
the current 4.1 weight percent U235. Fuel burnup limits and 
reactor operating power level would remain unchanged.
    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:

    Virginia Electric and Power Company has reviewed the Technical 
Specifications changes for Surry Units 1 and 2 against the criteria 
of 10 CFR 50.92. It has been concluded that use of fuel with the 
slightly higher initial enrichment does not involve a significant 
hazards consideration as defined in 10 CFR 50.92. An increase in the 
maximum initial fuel enrichment from 4.1 to 4.3 weight percent 
U235 will not:
    1. Involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated. 
The only accidents for which the probability of occurrence is 
potentially affected by the fuel enrichment involve criticality 
events during handling and storage. Analyses have demonstrated that 
the K-effective will be low enough to ensure subcriticality during 
both normal operation and under postulated accident conditions 
during the handling and storage of both new and spent fuel. 
Therefore, the probability of occurrence of criticality during fuel 
handling or storage is not increased. Safety analyses of record are 
based on inputs which bound the proposed increase in fuel 
enrichment. Since no changes to the fuel burnup limits are 
requested, the radiological consequences of previously evaluated 
accident scenarios will not be increased. Therefore, neither the 
probability of occurrence nor the consequences of any accident 
previously evaluated is significantly increased.
    2. Create the possibility for a new or different type of 
accident from any accident previously evaluated. Fuel with the 
higher initial enrichment will meet all applicable design criteria 
and will operate within existing Technical Specifications limits. 
Adherence to these standards and criteria precludes new challenges 
to components and systems that could introduce a new type of 
accident. All design and performance criteria will continue to be 
met. In addition, the use of a slightly higher initial fuel 
enrichment does not involve any alteration to plant equipment or 
procedures which would introduce any new or unique operational modes 
or accident precursors. Therefore, the possibility for a new or 
different kind of accident from any accident previously evaluated is 
not created.
    3. Involve a significant reduction in the margin of safety. 
Surry Units 1 and 2 will continue to operate in compliance with the 
Technical Specifications, ensuring that the plants continue to 
provide acceptable levels of protection for the health and safety of 
the public. The Technical Specifications are based upon 
assumption[s] made in the safety and accident analyses, including 
those relating to the fuel enrichment and the design of the fuel 
storage areas. Analyses have demonstrated that subcriticality will 
be ensured during fuel storage and handling accident scenarios for 
both new and spent fuel. Additionally, safety analyses of record for 
core operation will remain applicable for Surry Unit 1 and 2 cores 
which use fuel with the slightly higher U235 enrichment. 
Therefore, the regulated margin of safety as defined in the Bases to 
the Surry Technical Specifications is not reduced.
    Based on the preceding information, it has been determined that 
the use of fuel with an initial enrichment of up to 4.3 weight 
percent U235 satisfies the no significant hazards 
consideration criteria of 10 CFR 50.92.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied.
    Therefore, the NRC staff proposes to determine that the amendment 
request involves no significant hazards consideration.
    Local Public Document Room location: Swern Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: James E. Lyons.

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

[[Page 68321]]

amendments. If the Commission has prepared an environmental assessment 
under the special circumstances provision in 10 CFR 51.12(b) and has 
made a determination based on that assessment, it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

Commonwealth Edison Company

[Docket Nos. STN 50-454 and STN 50-455]

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

    Date of application for amendments: June 30, 1997, as supplemented 
on September 25, 1997.
    Brief description of amendments: The amendments grant partial 
credit for boron in the spent fuel pools to maintain the 
subcriticality.
    Date of issuance: December 4, 1997.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment Nos.: 94, 94, 86 and 86.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54868).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 4, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.

Duquesne Light Company, et al.

[Docket Nos. 50-334 and 50-412]

Beaver Valley Power Station, Unit Nos. 1 and 2, Shippingport, 
Pennsylvania

    Date of application for amendments: March 14, 1997, as 
supplemented. July 29, 1997, and August 13, 1997. The July 29, 1997, 
and August 13, 1997, letters provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination or expand the amendment request beyond the scope of the 
May 7, 1997, Federal Register notice.
    Brief description of amendments: These amendments relocate certain 
administrative control Technical Specifications (TSs) from the Beaver 
Valley Power Station, Unit Nos. 1 and 2 (BVPS-1 and BVPS-2), TSs to the 
licensee's operational quality assurance program description, which is 
presented in Section 17.2 of the BVPS-2 Updated Final Safety Analysis 
Report (UFSAR). Section 17.2 of the BVPS-2 UFSAR contains the quality 
assurance program description for both BVPS-1 and BVPS-2. The following 
TSs are being relocated to the quality assurance program description.

BVPS-2 TS 6.2.3 (Independent Safety Evaluation Group)
BVPS-1 and BVPS-2 TS 6.5.1 (Onsite Safety Committee)
BVPS-1 and BVPS-2 TS 6.5.2 (Offsite Review Committee)
BVPS-1 and BVPS-2 TS 6.8.2 (Procedures, Review and)
BVPS-1 and BVPS-2 TS 6.8.3 (Temporary Procedure Changes, Review and 
Approval)
BVPS-1 and BVPS-2 TS 6.10.1 (Records Retention, At least 5 Years)
BVPS-1 and BVPS-2 TS 6.10.2 (Records Retention, Duration of Operating 
License)

    Date of issuance: December 10, 1997.
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 209 and 87.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications, and Appendix C of the License.
    Date of initial notice in Federal Register: May 7, 1997 (62 FR 
24986).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 10, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001.

Entergy Operations, Inc.

[Docket No. 50-382]

Waterford Steam Electric Station, Unit 3, St. Charles Parish, Louisiana

    Date of amendment request: July 17, 1996, as supplemented by 
letters dated June 3, and July 7, 1997. Also, application dated April 
11, 1997.
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specification (TS) 3.7.1.3 by increasing the minimum 
required contained water volume in Condensate Storage Pool from 82 
percent to 91 percent indicated level. In addition, this amendment 
expands the applicability of TS 3.7.1.3 to include Mode 4 operational 
requirements. The amendment also deletes Action (b) in TS 3.7.1.3 and 
its associated surveillance requirement in Waterford 3 TSs.
    Date of issuance: December 18, 1997.
    Effective date: December 18, 1997, to be implemented within 60 
days.
    Amendment No.: 137.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 26, 1997 (62 FR 
14461), July 30, 1997 (62 FR 40849) and April 22, 1997 (62 FR 19624).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 18, 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.

Florida Power Corporation, et al.

[Docket No. 50-302]

Crystal River Unit No. 3 Nuclear Generating Plant, Citrus County, 
Florida

    Date of application for amendment: August 26, 1997.
    Brief description of amendment: The amendment involves a revision 
to the design basis of the Emergency Diesel Generator (EDG) Air 
Handling System at Crystal River 3 resulting from the EDG upgrade 
modification which increased the 200-hour and 2000-hour service ratings 
for each EDG.
    Date of issuance: December 12, 1997.
    Effective date: December 12, 1997.
    Amendment No.: 160.
    Facility Operating License No. DPR-31: Amendment revises the Final 
Safety Analysis Report.
    Date of initial notice in Federal Register: September 24, 1997 (62 
FR 50004).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 12, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal River, Florida 34428

[[Page 68322]]

Indiana Michigan Power Company

[Docket Nos. 50-315 and 50-316]

Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of application for amendments: September 19, 1997 
(AEP:NRC:1278).
    Brief description of amendments: The amendments modify Technical 
Specification 4.5.2.d.1 to delete the interlock that would close the 
Residual Heat Removal (RHR) suction valves if the Reactor Coolant 
System (RCS) pressure were to increase to 600 psig while retaining the 
interlock that would prevent the suction valves from opening while the 
RCS pressure is above the RHR system design pressure. This change 
maintains the open interlock function and allows continued deactivation 
of the isolation valves to assure RHR availability and provide low 
temperature overpressure protection.
    Date of issuance: December 10, 1997.
    Effective date: December 10, 1997, with full implementation within 
45 days.
    Amendment Nos.: 219 and 203.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54861).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 10, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

Pennsylvania Power and Light Company

[Docket Nos. 50-387 and 50-388]

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: October 7, 1996, as 
supplemented by letter dated May 9, 1997.
    Brief description of amendments: These amendments modify 
Susquehanna Steam Electric Station, Units 1 and 2, Technical 
Specifications Table 3.3.2-2 by revising the trip setpoints and 
allowable values for secondary containment isolation radiation 
monitors.
    Date of issuance: December 8, 1997.
    Effective date: December 8, 1997.
    Amendment Nos.: 170 and 143.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 18, 1996 (61 
FR 66716).
    The May 9, 1997, letter provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 8, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Pennsylvania Power and Light Company

[Docket Nos. 50-387 and 50-388]

Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: April 4, 1997, as supplemented 
April 14, June 6, and September 2, 1997.
    Brief description of amendments: These amendments clarify the scope 
of the surveillance requirements for response time testing of 
instrumentation in the reactor protection system, isolation actuation 
system, and emergency core cooling system in the Technical 
Specifications for each unit (Sections 4.3.1.3, 4.3.2.3, and 4.3.3.3).
    Date of issuance: December 8, 1997.
    Effective date: December 8, 1997.
    Amendment Nos.: 171 and 144.
    Facility Operating License Nos. NPF-14 and NPF-22: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 17, 1997 (62 FR 
17885).
    The April 14, June 6, and September 2, 1997, letters provided 
clarifying information that did not change the original proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 8, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Power Authority of the State of New York

[Docket No. 50-333]

James A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: December 14, 1995, as 
supplemented September 26, 1997.
    Brief description of amendment: The amendment proposes to change 
the James A. FitzPatrick Technical Specifications to incorporate the 
inservice testing requirements of Section XI of the American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code.
    Date of issuance: December 2, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 241.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1635).
    The September 26, 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 December 2, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Rochester Gas and Electric Corporation

[Docket No. 50-244]

R. E. Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: September 29, 1997, as 
supplemented October 8, 1997.
    Brief description of amendment: The amendment revises the Ginna 
Station Technical Specifications (TS) to allow referencing of revision 
of the Ginna Station pressure and temperature limits report for the 
reactor coolant system pressure and temperature limits and low 
temperature overpressure protection limits. The amendment also corrects 
a typographical error in the TSs.
    Date of issuance: December 9, 1997.
    Effective date: December 9, 1997.
    Amendment No.: 70.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.

[[Page 68323]]

    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59921).
    The September 29 and October 8, 1997, superseded in their entirety 
the applications dated December 13, 1996, April 24, 1997, and June 3, 
1997.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 9, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

Southern California Edison Company, et al.

[Docket Nos. 50-361 and 50-362]

San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, San Diego 
County, California

    Date of application for amendments: December 22, 1995, as 
supplemented by letter dated November 25, 1997.
    Brief description of amendments: These amendments revise License 
Conditions 2.E and 2.G for the San Onofre Nuclear Generating Station 
(SONGS), Units 2 and 3. The amendments delete the physical protection 
program reporting requirement from License Condition 2.G, and clarify 
in License Condition 2.E that not all documents composing the physical 
protection program plans necessarily contain safeguards information.
    Date of issuance: December 16, 1997.
    Effective date: December 16, 1997.
    Amendment Nos.: Unit 2--138; Unit 3--130.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Facility Operating Licenses.
    Date of initial notice in Federal Register: November 5, 1997 (62 FR 
59921). The November 25, 1997, letter provided additional clarifying 
information and did not change the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated December 16, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Virginia Electric and Power Company, et al.

[Docket Nos. 50-338 and 50-339]

North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of application for amendments: May 14, 1997, as supplemented 
October 15, 1997. The October 15, 1997, submittal provided clarifying 
information only, and did not change the proposed no significant 
hazards consideration determination.
    Brief description of amendments: The proposed action consists of 
changes to the Technical Specifications (TS) revising Surveillance 
Requirement 4.7.1.7.2.a for both units to clarify the testing and 
inspection methodology of the turbine governor control valves. The 
proposed changes also provide clarification in the TS Bases Section 3/4 
7.1.7 for the Turbine Valve Freedom Testing of the turbine governor 
control valves.
    Date of issuance: December 4, 1997.
    Effective date: December 4, 1997.
    Amendment Nos.: 207 and 188.
    Facility Operating License Nos. NPF-4 and NPF-7: Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: July 30, 1997 (62 FR 
40860).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 4, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

    Dated at Rockville, Maryland, this 24th day of December 1997.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 97-33968 Filed 12-30-97; 8:45 am]
BILLING CODE 7590-01-P