[Federal Register Volume 62, Number 68 (Wednesday, April 9, 1997)]
[Notices]
[Pages 17223-17252]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-8916]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from March 17, 1997 through March 28, 1997. The 
last biweekly notice was published on March 26, 1997 (62 FR 14457).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, and should cite the 
publication date and page number of this Federal Register notice. 
Written comments may also be delivered to Room 6D22, Two White Flint 
North, 11545 Rockville Pike, Rockville, Maryland from 7:30 a.m. to 4:15 
p.m. Federal workdays. Copies of written comments received may be 
examined at the NRC Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC. The filing of requests for a hearing and 
petitions for leave to intervene is discussed below.
    By May 9, 1997, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert

[[Page 17224]]

opinion which support the contention and on which the petitioner 
intends to rely in proving the contention at the hearing. The 
petitioner must also provide references to those specific sources and 
documents of which the petitioner is aware and on which the petitioner 
intends to rely to establish those facts or expert opinion. Petitioner 
must provide sufficient information to show that a genuine dispute 
exists with the applicant on a material issue of law or fact. 
Contentions shall be limited to matters within the scope of the 
amendment under consideration. The contention must be one which, if 
proven, would entitle the petitioner to relief. A petitioner who fails 
to file such a supplement which satisfies these requirements with 
respect to at least one contention will not be permitted to participate 
as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Docketing and 
Services Branch, or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001, 
and to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1) (i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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: March 5, 1997.
    Description of amendments request: The proposed amendments would 
incorporate a new Technical Specification (TS) for instrumentation 
associated with automatic isolation of a pathway for release of non-
condensible gases from the main condenser. At power levels of 5 percent 
or less, mechanical vacuum pumps are used to remove non-condensible 
gases from the condenser using a pathway to the release stack that 
bypasses the normal holdup and filter train. The proposed TS will 
require that four channels of the main steam line radiation--high 
isolation function be capable of tripping the mechanical vacuum pumps 
and closing an isolation valve in the release pathway. Surveillance 
requirements are included in the TS to ensure the isolation 
instrumentation will perform its intended function.
    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 change incorporates a new Technical Specification 
3/4.3.8, ``Condenser Vacuum Pump Isolation Instrumentation.'' This 
specification will require that the main steam line radiation--high 
isolation function be capable of tripping the condenser vacuum 
pump(s) and isolate the associated common isolation valve. Four 
instrumentation channels of this function are required to be 
operable when the unit is in OPERATIONAL CONDITION 1 or 2 with a 
condenser vacuum pump in operation. Adding the requirement to trip 
the condenser vacuum pumps does not affect the probability of an 
accident previously evaluated. The probability of component failure 
of the proposed design for condenser vacuum pump isolation devices 
is the same as that of the original licensing basis. As a result, 
the capability to isolate the condenser vacuum pump will not be 
significantly impacted.
    CP&L contracted Scientech-NUS to recalculate the main control 
room doses resulting from a control rod drop accident assuming main 
steam line radiation monitors isolate the condenser vacuum pump(s) 
and determined the dose to be 23.2 rem thyroid and 0.05 rem whole 
body, which is less than the General Design Criterion (GDC) 19/
Standard Review Plan (SRP) Section 6.4 limits of 30 rem thyroid and 
5 rem whole body. The offsite doses at the exclusion area boundary 
after 2 hours are 0.16 rem thyroid and 0.015 rem whole body, which 
is less than the SRP Section 15.4.9 limits. The low population zone 
(LPZ) dose is estimated to be about 1 rem thyroid, which is also 
well below regulatory limits. Therefore, the proposed [amendments 
do] not increase the consequences of an accident previously 
evaluated.
    2. The proposed amendment[s] would not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed [amendments add] new requirements to ensure the 
capability to trip the condenser vacuum pump(s). The proposed 
[changes do] not affect the operability of equipment designed to 
mitigate the consequences of an accident nor [do they] create a 
potential to initiate a new type of accident. 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 license [amendments do] not involve a 
significant reduction in a margin of safety.
    The safety-related main steam line radiation monitors provide a 
highly reliable means to detect radioactivity resulting from a 
control rod drop accident and will provide automatic trip of the 
condenser vacuum pumps and isolation of the associated isolation 
valve. Use of the main steam line radiation monitors for this 
application is consistent with the original Brunswick Steam Electric 
Plant design for condenser pump and associated valve isolation. CP&L 
contracted Scientech-NUS to recalculate the main control room doses 
resulting from a control rod drop accident assuming main steam line 
radiation monitors isolate the condenser

[[Page 17225]]

vacuum pump(s) and determined it to be 23.2 rem thyroid and 0.05 rem 
whole body, which is less than the GDC 19/SRP Section 6.4 limits of 
30 rem thyroid and 5 rem whole body. The offsite doses at the 
exclusion area boundary after 2 hours are 0.16 rem thyroid and 0.015 
rem whole body, which is less than the SRP Section 15.4.9 limits. 
LPZ dose is estimated to be about 1 rem thyroid, which is also well 
below regulatory limits. Therefore, the proposed [changes do] not 
involve a significant reduction in a margin of safety.

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

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: February 18, 1997.
    Description of amendment request: The proposed change revises the 
Plant System Turbine Cycle Technical Specification (TS) 3/4.7.1 by 
revising the power range high neutron flux setpoint values in TS Table 
3.7-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The high flux setpoints are being revised to provide additional 
margin against secondary side overpressurization for LOL/TT [loss-
of-load/turbine trip] events. The proposed revision will not create 
any loss or reduction in redundancy or diversity in the reactor 
protection systems that would increase the probability of a 
previously evaluated accident. The high flux setpoints are being 
revised to ensure that the consequences of a previously evaluated 
accident do not increase.
    Therefore, there would be no 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.
    No new or previously unanticipated failure mechanisms are 
introduced by the proposed change. No new failure modes have been 
created by the proposed change. No new credible event or initiating 
factor is introduced. Reactor power is limited to ensure that the 
secondary system is not overpressurized.
    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.
    The margin of safety as defined in the basis of the Technical 
Specification does not decrease. This change is proposed to ensure 
that the secondary system pressure will be limited to within 110% of 
its design pressure during the most severe anticipated operational 
transient. The revised high flux setpoints are intended to bound the 
allowable operating configurations of TS Table 3.7-1.
    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: Mark Reinhart, Acting.

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: February 21, 1997.
    Description of amendment request: The proposed change adds a 
definitive time limit to Technical Specification 3.3.2 in Action 16 of 
Table 3.3-3 to place an inoperable channel into bypass.
    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change does not affect the operation or design of 
the plant in any way. The requirement to place the channel into 
bypass already exists and this change simply provides a specific 
time limit. This logic circuit is not an initiator of any event and 
with no change in logic or operation there is no change in 
consequences.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed specific time limit does not involve any physical 
alterations or additions to plant equipment or alter the manner in 
which any safety-related system performs its function. 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.
    The proposed change replaces an indeterminate time period with a 
specific limit of six hours. Six hours is a reasonable period in 
which to complete this requirement and is identical to the time 
allowed for these functions in NUREG-1431 [Standard Specifications 
Westinghouse Plants]. 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: 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: Mark Reinhart, Acting.

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: February 27, 1997.
    Description of amendment request: The proposed change adds sleeve 
installation as an alternative to tube plugging for repairing degraded 
steam generator tubes to Technical Specification 3/4.4.5, Steam 
Generators.

[[Page 17226]]

    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 does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The only equipment affected by sleeving is the steam generator 
tubes. The most severe malfunction of a steam generator tube is a 
tube rupture. The consequences of a ruptured sleeve are no greater 
than the consequences of a ruptured tube. Sleeving does not increase 
the probability of a steam generator tube failure because the 
sleeved tube has been shown to have a significant safety factor for 
burst and collapse pressures as well as demonstrated acceptable 
resistance to corrosion and fatigue loading. Thus, a steam generator 
with sleeved tubes would perform in the same manner as one without 
sleeved tubes.
    A sleeved tube is functionally equivalent to an unsleeved tube 
except for less effective heat transfer due to the air gap and a 
slightly higher pressure drop due to the primary flow restriction. 
These differences are bounded by the current tube plugging limits.
    Analysis and testing have demonstrated that the sleeves are 
structurally adequate to withstand the load existing within the 
steam generator tubes whether the original tube is still intact or 
is breeched.
    There is no increase in the possibility for increased fatigue 
loadings. There is no possibility for the sleeve to become dislodged 
from its plugging location and enter the RCS [Reactor Coolant 
System] flow path.
    The plant safety analysis for tube plugging bounds tube 
sleeving.
    The proposed change has no significant effect on the 
configuration of the plant. The proposed change does not affect the 
way in which the plant is operated. Therefore, there would be no 
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.
    A sleeved tube is functionally equivalent to an unsleeved tube 
except for less effective heat transfer due to the air gap and a 
slightly higher pressure drop due to the primary flow restriction. 
These differences are bounded by the current tube plugging limits.
    The sleeved tube has been shown to have a significant safety 
factor for burst and collapse pressures as well as demonstrated 
acceptable resistance to corrosion and fatigue loading. Thus, a 
steam generator with sleeved tubes would perform in the same manner 
as one without sleeved tubes.
    The proposed change has no significant effect on the 
configuration of the plant. The proposed change does not affect the 
way in which the plant is operated. 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.
    The proposed revision to permit the installation of tube sleeves 
does not reduce the margin of safety as presently defined in 
Technical Specification BASES section 3/4.4.5. This margin of safety 
includes primary to secondary leakage limits and tube plugging 
limits which are not changed by the proposed amendment. The analyses 
and testing of the proposed sleeve design demonstrates that the 
structural integrity of the RCS is maintained. Design of the tube 
sleeve considers mechanical/structural aspects, water chemistry and 
metallurgical aspects as well as thermal/hydraulic considerations.
    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: Mark Reinhart, Acting.

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: March 10, 1997.
    Description of amendment request: The proposed changes to Technical 
Specification 3.5.1 provide an optional method of meeting surveillance 
requirements by allowing the use of instrument readings to meet 
surveillance 4.5.1.1.a.1, and adds a new Action c to cover a condition 
in which one accumulator has a boron concentration not within limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The accumulators are not initiators of any event and so the 
probability of occurrence of an event is unaffected by either of the 
proposed changes. The use of actual instrumentation readings to 
comply with the surveillance does not change the function or 
performance of the accumulators and thus does not affect any 
accident consequences. The increase in the allowed time to restore 
the boron concentration to within limits is consistent with allowed 
out of service times for other Emergency Safeguards equipment.
    It will not have a significant impact on subcriticality during 
reflood. Therefore, there will be no increase in the consequences of 
an accident.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes to the accumulator specification do not 
involve any physical alterations or additions to plant equipment or 
alter the manner in which any safety-related system performs its 
function. 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.
    The proposed change to the surveillance requirement provides an 
equivalent means of meeting the requirement. Since there is no 
change in either the accumulator limits or the surveillance 
frequency, there is no reduction in safety margin. The new Action c 
to address returning the boron concentration of a single accumulator 
to within limits allows an out of service time commensurate with the 
times allowed for other Engineered Safeguards Features. The boron 
concentration of one accumulator does not have a significant impact 
on subcriticality during reflood and thus does not involve a 
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: Mark Reinhart, Acting.

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: March 14, 1997.
    Description of amendment request: The amendment will revise the 
Final Safety Analysis Report to include the

[[Page 17227]]

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

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The changes described do not impact the probability of 
occurrence of accidents previously analyzed. Removal of the valve 
box covers and all but four of the cask closure head sleeve nuts has 
no impact on accident initiators. Dose assessments using maximum 
potential releases assuming failure of the spent fuel and 
radionuclide release through the gap between the cask closure head 
and the cask or damage to the valves show that no significant 
increase in consequences of an accident previously evaluated would 
occur. [Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.]
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Compromising the integrity of the cask by removing the valve box 
covers and closure head sleeve nuts in preparation for unloading the 
spent fuel from the cask does not create the possibility of a new 
type of accident or equipment malfunction. No safety-related 
equipment, safety function, or operations of plant equipment will be 
altered as a result of this change. Therefore, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The NRC basis for acceptance of a spent fuel cask drop is 
documented in Section 15.7.5 of the Safety Evaluation Report, NUREG-
1038, dated November 1983. It states, ``* * * no loss of cask 
integrity is postulated to occur in the event of a drop, and the 
staff concludes there will be no significant radiation released to 
the environment. The radiological consequences will be less than a 
small fraction of the 10 CFR 100 exposure guideline values.''
    As described in the proposed change, even though complete cask 
integrity may not be preserved in the event of a loaded cask drop 
with the valve box covers removed or with only four, rather than 32, 
closure head sleeve nuts installed, the radiological consequences 
calculated using conservative assumptions were determined to be a 
small fraction of the 10 CFR 100 values. 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: Mark Reinhart, Acting.

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

    Date of amendment request: June 20, 1996, as supplemented by 
letters dated December 30, 1996, and March 5, 1997.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) by incorporating NRC approved 
thermal limit licensing methodology in the list of approved 
methodologies used in establishing the fuel cycle specific thermal 
limits. In addition, the proposed amendment would correct mirror 
editorial items in the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits will be established consistent with NRC 
approved methods to ensure that fuel performance during normal, 
transient and accident conditions is acceptable. The proposed 
Technical Specifications amendment reflects NRC approved SPC 
methodology used to analyze normal operations, including anticipated 
operational occurrences (AOOs), and to determine the potential 
consequences of accidents.

Licensing Methods and Models

    The proposed amendment is to support operation with NRC approved 
fuel and licensing methods supplied from Siemens Power Corporation 
[SPC]. In accordance with [Updated Final Safety Analysis Report] 
UFSAR Chapter 15, the same accidents and transients will be analyzed 
with the new fuel and methods. The latest NRC approved revision to 
the Siemens [loss-of-coolant accident] LOCA analysis methodology 
(Reference: ANF-91-048(P)(A), Advanced Nuclear Fuels Corporation 
Methodology for Boiling Water Reactors EXEM BWR Evaluation Model) 
will be used to evaluate the ATRIUM-9B and other co-resident fuel 
types. The other licensing analysis methods and models are also NRC 
approved. The approved methods and models are used to determine the 
fuel thermal limits (e.g., average planar linear heat generation 
rate, transient linear heat generation rate, minimum critical power 
ratio and linear heat generation rate). The SPC core monitoring code 
enables the site to monitor keff as well as control rod density 
to perform the reactivity anomaly surveillance. Therefore, the 
change in licensing analysis methods and models does not 
significantly increase the probability of an accident or the 
consequences of an accident previously identified. The support 
systems for minimizing the consequences of transients and accidents 
are not affected by the proposed amendment.

New Fuel Design

    The use of reload quantities of ATRIUM-9B fuel at Dresden does 
not involve a significant increase in the probability or 
consequences of any accident previously evaluated in the [Final 
Safety Analysis Report] FSAR. The ATRIUM-9B fuel is generically 
approved for use as a reload BWR fuel type (Reference: ANF-89-
014(P)(A) Revision 1 Supplement 1, Generic Mechanical Design for 
Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). Limiting 
postulated occurrences and normal operation have been analyzed using 
NRC-approved methods for the ATRIUM-9B fuel design to ensure that 
safety limits are protected and that acceptable transient and 
accident performance is maintained.
    The reload fuel has no adverse impact on the performance of in-
core neutron flux instrumentation or CRD response. The ATRIUM-9B 
fuel design will not adversely affect performance of neutron 
instrumentation nor will it adversely affect the movement of control 
blades relative to the current Dresden fuel type, the Siemens 
manufactured 9x9-2. The exterior dimensions of the ATRIUM-9B fuel 
have been evaluated by ComEd; the ATRIUM-9B fuel design provides 
adequate clearances relative to the co-resident 9x9-2 fuel. Thus, no 
increased interactions with the adjacent control blade or nuclear 
instrumentation are created. Additionally, given the above mentioned 
overall envelope similarities, no problems are anticipated with 
other station equipment such as the fuel storage racks, the new fuel 
inspection stand and the spent fuel storage pool fuel preparation 
machine. Therefore, the probability of adverse interactions between 
the ATRIUM-9B fuel and components in the core and fuel handling 
equipment is not significantly increased.
    The ATRIUM-9B design is neutronically compatible with the 
existing fuel types and core components in the Dresden core. SPC 
tests have demonstrated that the ATRIUM-9B fuel design is 
hydraulically compatible with the co-resident 9x9-2 fuel. The bundle 
pressure drop characteristics of the ATRIUM

[[Page 17228]]

9B bundle are similar to those of the 9x9-2 fuel design, hence core 
thermal-hydraulic stability characteristics are not adversely 
affected by the ATRIUM-9B design. Cycle stability calculations are 
performed by SPC. Therefore, the probability of thermal hydraulic 
instability is not significantly increased.
    Evaluations of the Dresden Emergency Procedures and UFSAR 
Chapter 15 AOOs are being performed to ensure that the use of the 
ATRIUM-9B fuel at Dresden does not alter any assumptions previously 
made in evaluating the radiological consequences of an accident at 
Dresden Units 2 and 3. Therefore, the radiological consequences of 
accidents are not significantly increased.
    Methods approved by the NRC are being used in the evaluation of 
fuel performance during normal and abnormal operating conditions. 
The ComEd and SPC methods to be used for the cycle specific 
transient analyses have been previously NRC approved. The proposed 
methodologies are administrative in nature and do not significantly 
affect any accident precursors or accident results; as such, the 
proposed change to the listing of the SPC methodologies for Dresden 
does not significantly increase the probability or consequences of 
any previously evaluated accidents.
    The description of the fuel is modified to include the water box 
design of the NRC approved ATRIUM-9B fuel type.
    Review of the above concludes that the probability of occurrence 
and the consequences of an accident previously evaluated in the 
safety analysis report have not been significantly increased.
* * * * *
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated:
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation.

Licensing Methods and Models

    The proposed Technical Specification amendment reflects 
previously approved SPC methodology used to analyze normal 
operations, including AOOs, and to determine the potential 
consequences of accidents. In accordance with FSAR Chapter 15, the 
same accidents and transients will be analyzed with the new fuel and 
method as have been previously performed. As stated above, the 
proposed changes do not permit modes of reactor operation which 
differ from those currently permitted; therefore, the possibility of 
a new or different kind of accident is not created. Plant support 
equipment is not affected by the proposed changes; therefore, no new 
failure modes are created.

New Fuel Design

    The basic design concept of a 9x9 fuel pin array with an 
internal water box has been used in various lead assembly programs 
and in reload quantities in Europe since 1986. WNP-2 has loaded 
reload quantities since 1991. Eight lead ATRIUM-9B assemblies were 
loaded into Dresden 2 during Cycle 15. Approximately 650 water box 
assemblies have been irradiated in the United States through 1995, 
with a substantially higher number being irradiated overseas. The 
NRC has reviewed and approved the ATRIUM-9B fuel design (Reference: 
ANF-89-014(P)(A) Revision 1 Supplement 1, Generic Mechanical Design 
for Advanced Nuclear Fuels 9X9-IX and 9X9-9X BWR Reload Fuel). The 
similarities in fuel design and operation between the ATRIUM-9B and 
the 9x9-2, and the previous Boiling Water Reactor experience with 
Siemens fuel, indicate there would be no new or different types of 
accidents for Dresden than have been considered for the existing 
fuel. Therefore, the use of ATRIUM-9B fuel at Dresden does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
* * * * *
    3. Involve a significant reduction in the margin of safety for 
the following reasons:
    The existing margin to safety is provided by the existing 
acceptance criteria (e.g., 10 CFR 50.46 limits). The proposed 
Technical Specification amendment reflects previously approved SPC 
methodology used to demonstrate that the existing acceptance 
criteria are satisfied. The revised LOCA methodology has been 
previously reviewed and approved by the USNRC for application to 
reload cores of BWRs. References for the Licensing Topical Reports 
which document this methodology, and include the Safety Evaluation 
Reports prepared by the USNRC, are added to the Reference section of 
the Technical Specifications as part of this amendment.

Licensing Methods and Models

    The proposed amendment does not involve changes to the existing 
operability criteria. NRC approved methods and established limits 
(implemented in the COLR) ensure acceptable margin is maintained. 
The ComEd and SPC reload methodologies for the ATRIUM-9B reload 
design are consistent with the Technical Specification Bases. The 
Limiting Conditions for Operation are taken into consideration while 
performing the cycle specific and generic reload safety analyses. 
USNRC approved methods are listed in Specification 6.9.A of the 
Technical Specifications.
    Analyses performed with USNRC-approved methodology have 
demonstrated that fuel design and licensing criteria will be met 
during normal and abnormal operating conditions. The same margins of 
safety will continue to be utilized by SPC (e.g., limits on peak 
cladding temperature, cladding oxidation, plastic strain). 
Therefore, there is not a significant reduction in the margin of 
safety.

New Fuel Design

    The exterior dimensions of the ATRIUM-9B fuel assembly result in 
equivalent clearances relative to the co-resident 9x9-2 fuel. Thus, 
no increased interactions with the adjacent control blade and 
nuclear instrumentation are created. The change does not adversely 
impact equipment important to safety; therefore the margin of safety 
is not significantly reduced.

    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 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: March 18, 1997.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) by increasing the High 
Pressure Coolant Injection (HPCI) isolation setpoint from greater than/
equal to 80 psig to greater than/equal to 100 psig. The licensee has 
requested the change to ensure consistency between the Updated Final 
Safety Analysis Report (UFSAR), design basis documents and the TS. The 
function of the setpoint is to assure the HPCI turbine steam supply is 
isolated in the event that the reactor scram supply pressure falls 
below the stall pressure of the HPCI turbine and the system seals are 
no longer effective in controlling the release of potentially 
contaminated steam.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    The Low Reactor Pressure isolation of the HPCI steam supply 
lines is provided to prevent damage to the HPCI turbine when the 
reactor steam pressure has decreased below that required to provide 
adequate motive force to operate the system. The steam supply 
isolation low reactor pressure setpoint is not an assumed initiator 
or contributor to any previously evaluated accident and therefore 
this change does not involve an increase in the probability of an 
accident previously evaluated at Dresden Station.
    The Lower Reactor Pressure isolation of the HPCI steam supply 
lines is described in the

[[Page 17229]]

plant safety analysis as a backup protection to other system and 
facility design features which provide assurance that accident 
transients will not result in failures of the system which 
contribute significantly to the consequences of the initiating 
accident. The low reactor pressure isolation signal provides backup 
to other isolation signals to ensure isolation will occur, 
minimizing the radiation dose as a result of steam leakage past the 
turbine seals in the event of a locked rotor due to damage from 
liquid carryover due to postulated swell in the reactor vessel.
    These analyses assume the isolation function occurs at 100 psig, 
and the proposed setpoint of greater than or equal to 100 psig is 
consistent and conservative with respect to these assumptions. 
Because the isolation function is not an accident initiator and the 
revised setpoint ensures that the isolation function continues to 
minimize radiological consequences, the consequences of any accident 
previously evaluated is not increased by the proposed changes.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed change administratively increases the Low Reactor 
Vessel Pressure trip setpoint which initiates HPCI isolation. This 
change does not result in any new or different modes of operation. 
The proposed change increases the setpoint at which the HPCI turbine 
steam supply will be isolated as the reactor vessel pressure 
decreases following a postulated accident. The proposed new setpoint 
is conservative with respect to the existing TS limit, i.e. the new 
limit of greater than or equal to 100 psig is consistent and 
permitted by the existing limit of greater than or equal to 80 psig. 
The change assures that the Trip Setpoint in the TS accurately 
reflects the design basis and UFSAR described limits.
    Because the proposed change does not result in any new modes of 
plant operation and administratively increases the system isolation 
setpoint in a conservative manner, the proposed change does not 
create the possibility of a new or different kind of accident from 
those previously evaluated.
    3. Involve a significant reduction in the margin of safety 
because:
    The Trip Setpoint provides assurance that the HPCI turbine 
cannot be operated with a steam supply pressure too low to drive the 
turbine and pump. The isolation assures that the turbine does not 
stall and minimizes the potential for the release of radioactivity 
which results from steam leakage past the turbine seals. The 
proposed change increases the setpoint, ensuring that the required 
isolation occurs at a higher pressure which is more conservative, 
i.e. it assures the turbine is isolated before the inlet steam 
pressure falls to the stall pressure of the HPCI turbine and leakage 
occurs. The greater than or equal to 100 psig limit is well below 
the range of reactor vessel pressure for which HPCI is required to 
perform its safety function. Therefore, the margin of safety 
provided by the function of the HPCI isolation on low reactor vessel 
pressure is increased by the proposed TS change, and this change 
will not involve a reduction in the margin of safety.
    As described, the proposed amendment for Dresden will not reduce 
the availability of systems required to mitigate accident 
conditions. Neither are new or significantly different modes of 
operation proposed. Therefore, the proposed change does not involve 
a significant reduction in the margin of safety.
    Guidance has been provided in ``Final Procedures and Standards 
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, 
for the application of standards to license change requests for 
determination of the existence of significant hazards 
considerations. This document provides examples of amendments which 
are and are not considered likely to involve significant hazards 
considerations.
    This proposed amendment does not involve any irreversible 
changes, a significant relaxation of the criteria used to establish 
safety limits, a significant relaxation of the bases for the 
limiting safety system settings or a significant relaxation of the 
bases for the limiting conditions for operations. Therefore, based 
on the guidance provided in the Federal Register and the criteria 
established in 10 CFR 50.92(c), the proposed change does not 
constitute a significant hazards consideration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment request: December 27, 1995, as supplemented 
September 4, October 18, and November 26, 1996.
    Description of amendment request: The proposed amendment would 
revise technical specifications (TS) related to electrical power 
systems. The proposed changes include revisions to limiting conditions 
for operation (LCO), LCO applicability and action statements, allowed 
outage times (AOT), surveillance requirements (SR), and administrative 
controls. The changes add new requirements, revise or delete existing 
requirements, relocate certain existing requirements to other licensee 
controlled documents, and editorially restructure the proposed 
requirements to closely emulate the electrical power system 
requirements of NUREG-1432, ``Standard Technical Specifications for 
Combustion Engineering Plants,'' (STS). The proposed requirements 
differ from the requirements of the STS where necessary to reflect 
features unique to the Palisades design. Each proposed change has been 
classified by the licensee as Administrative, Relocated, More 
Restrictive, or Less Restrictive.
    Basis for proposed no significant hazards consideration 
determination: A proposed amendment to an operating license for a 
facility involves no significant hazards consideration if operation of 
the facility in accordance with the proposed amendment would not: (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; (2) create the possibility of a new or 
different kind of accident from any previously evaluated; or (3) 
involve a significant reduction in a margin of safety. 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:

    Evaluation of ADMINISTRATIVE, RELOCATED, and MORE RESTRICTIVE 
changes:
    ADMINISTRATIVE changes and RELOCATED changes move requirements, 
either within the TS or to documents controlled under 10 CFR 50.59, 
or [clarify] existing TS requirements, without affecting their 
technical content. Since ADMINISTRATIVE and RELOCATED changes do not 
alter the technical content of any requirements, they cannot involve 
a significant increase in the probability or consequences of an 
accident previously evaluated, create the possibility of a new or 
different kind of accident from any previously evaluated, or involve 
a significant reduction in a margin of safety.
    MORE RESTRICTIVE changes only add new requirements, or revise 
existing requirements to result in additional operational 
restrictions. Since the TS, with all MORE RESTRICTIVE changes 
incorporated, will still contain all of the requirements which 
existed prior to the changes; MORE RESTRICTIVE changes cannot 
involve a significant increase in the probability or consequences of 
an accident previously evaluated, create the possibility of a new or 
different kind of accident from any previously evaluated, or involve 
a significant reduction in a margin of safety.
    Evaluation of LESS RESTRICTIVE changes:
    1. Do these LESS RESTRICTIVE changes involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Change 3 revised the requirement for operable AC sources, using 
more general wording than the existing TS. The existing LCO requires 
that two explicitly specified transformers be operable; the proposed 
LCO requires that two qualified offsite circuits be operable. The 
proposed LCO will allow

[[Page 17230]]

substitution of Safeguards Transformer 1-1 for Station Power 
Transformer 1-2 as a required AC source, but the quantity and 
quality of required offsite AC sources is unaffected. Since the 
capability and qualification of Safeguards Transformer 1-1 are 
equivalent to those of the Station Power transformer, neither the 
probability or consequences of an accident previously evaluated will 
be increased.
    Change 10 is less restrictive only in its allowance of a 72 hour 
AOT for an inoperable offsite source instead of the 24 hour AOT 
currently required. The change also makes a considerably more 
restrictive change by eliminating the allowance, based on submittal 
of a report, for continuous operation with Startup Transformer 1-2 
inoperable. Changing an AOT, alone, cannot increase the probability 
or consequences of an accident previously evaluated.
    Change 14 allows, for an inoperable DG [diesel generator], 
verification that no common cause failure is involved in lieu of 
test starting the other DG. The intent of the test starting 
requirement is to verify that there is no common cause failure which 
also makes the other DG inoperable. The proposed action statement 
thereby accomplishes the same objective as that it replaces. Since 
the proposed action statement accomplishes the same objective as the 
one it replaces, operation in accordance with the proposed change 
will not increase the probability or consequences of an accident 
previously evaluated.
    Change 21 revises the SR for the DG starting test. [The ``Less 
Restrictive'' elements of the change eliminate the requirement to 
vary use of the A and B starting circuits for each monthly test, 
because the DG is not assumed to be single failure proof; and 
eliminate requirements that the DGs be manually started and that 
they be synchronized from the control room, because no practical 
alternatives exist for accomplishing these actions]. The proposed 
change does not alter any plant operating conditions, operating 
practices, equipment settings, or equipment capabilities. Therefore, 
operation of the facility in accordance with the proposed change 
will not involve an increase in the probability of an accident. 
Change 21 requires more rigorous testing of the DGs than required by 
the existing Technical Specifications. The more rigorous testing is 
intended to provide additional assurance that the DGs are capable of 
performing their design function and should, therefore, involve a 
reduction, rather than an increase, in the consequences of those 
accidents previously evaluated.
    Change 25 revises the SR for testing the fuel transfer system. 
The proposed change does not alter any plant operating conditions, 
operating practices, equipment settings, or equipment capabilities. 
Therefore, operation of the facility in accordance with the proposed 
change will not involve an increase in the probability of an 
accident. The only ``Less Restrictive'' feature of proposed SR is 
test interval extension from ``each month'' to ``each 92 days.'' 
Changing a surveillance frequency, alone, cannot increase the 
probability or consequences of an accident previously evaluated.
    Change 26 revises the station battery SRs. The proposed monthly 
and quarterly battery SRs contain all of the test requirements of 
the existing SRs with two exceptions: (1) The proposed interval for 
measuring each cell voltage is ``each 92 days'' instead of the 
existing ``every month'' and (2) the requirement to record the 
amount of water added has been deleted. Changing a surveillance 
frequency or deleting a maintenance record cannot increase the 
probability or consequences of an accident previously evaluated.
    2. Do changes create the possibility of a new or different kind 
of accident from any previously evaluated?
    Change 3 only involves the specified offsite power sources. 
Since the Loss of Offsite Power is already considered in the 
accident analyses, operating the facility in accordance with Change 
3 will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    Change 10 revises an AOT; Change 14 revises a required action; 
Change 21 revises a testing requirement; Changes 25 and 26 revise a 
surveillance interval; and Change 26 deletes the requirement for a 
maintenance record. None of these proposed changes alter any plant 
operating conditions, operating practices, equipment settings, or 
equipment capabilities. Therefore, operation of the facility in 
accordance with the proposed changes will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Do changes involve a significant reduction in a margin of 
safety?
    Change 3 does not alter the quantity or quality of offsite 
sources required to be available. Therefore, operating the facility 
in accordance with the proposed change will not involve a reduction 
in a margin of safety.
    Change 10 revises an AOT; Change 14 revises a required action, 
Change 21 revises a testing requirement; Changes 25 and 26 revise a 
surveillance interval; and Change 26 deletes the requirement for a 
maintenance record. These proposed changes do not alter any plant 
operating conditions, operating practices, equipment settings, or 
equipment capabilities. Therefore, operating the facility in 
accordance with the proposed change will not involve a reduction in 
a margin of safety.

    The licensee's September 4, 1996, supplement stated that three of 
the proposed changes contained in the supplement were not addressed in 
the December 27, 1995, no significant hazards analysis. The changes 
involved TS requirements that would be deleted. Equivalent requirements 
would be incorporated in the FSAR or other documents subject to the 
controls of 10 CFR 50.59. The licensee's analysis of the issue of no 
significant hazards consideration for these changes is presented below:

    1. Do changes which relocate a requirement from the TS to 
documents which are controlled under 10 CFR 50.59 involve a 
significant increase in the probability or consequences of an 
accident previously evaluated?
    10 CFR 50.59 specifically prohibits [without obtaining prior NRC 
review and approval] changes to the facility as described in the 
safety analysis report, and to procedures described in the safety 
analysis report ``if the probability of occurrence or the 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated in the safety analysis report may be 
increased''. Since the conditions which limit changes performed 
under 50.59 are more restrictive than the conditions which define 
changes considered to involve a significant hazards consideration, 
relocation of a requirement from the TS to the FSAR [Final Safety 
Analysis Report] or to documents which are referenced by the FSAR 
cannot involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Do changes which relocate a requirement from the TS to 
documents which are controlled under 10 CFR 50.59 create the 
possibility of a new or different kind of accident from any 
previously evaluated?
    10 CFR 50.59 specifically prohibits [without obtaining prior NRC 
review and approval] changes to the facility as described in the 
safety analysis report, and to procedures described in the safety 
analysis report ``if a possibility for an accident or malfunction of 
a different type than any evaluated previously in the safety 
analysis report may be created''. Since the conditions which limit 
changes performed under 50.59 are more restrictive than the 
conditions which define changes considered to involve a significant 
hazards consideration, relocation of a requirement from the TS to 
the FSAR or to documents which are referenced by the FSAR cannot 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Do these changes which relocate a requirement from the TS to 
documents which are controlled under 10 CFR 50.59 involve a 
significant reduction in a margin of safety?
    10 CFR 50.59 specifically prohibits [without obtaining prior NRC 
review and approval] changes to the facility as described in the 
safety analysis report, and to procedures described in the safety 
analysis report ``if the margin of safety as defined in the basis 
for any technical specification is reduced''. Since the conditions 
which limit changes performed under 50.59 are more restrictive than 
the conditions which define changes considered to involve a 
significant hazards consideration, relocation of a requirement from 
the TS to the FSAR or to documents which are referenced by the FSAR 
cannot involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analyses 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: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

[[Page 17231]]

    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: March 10, 1997.
    Description of amendment request: The proposed amendment would 
modify Unit 1 Technical Specification (TS) 5.2.1 to add ZIRLO as fuel 
assembly material and add reference to Nuclear Regulatory Commission 
approved Topical Report, WCAP-12610, ``Vantage+ Fuel Assembly Reference 
Core Report'', to TS 6.9.1.12 for both units.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The methodologies used in the accident analyses have been 
modified to reflect the requirements provided in WCAP-12610, 
VANTAGE+ Fuel Assembly Reference Core Report. Reference to this NRC 
approved ZIRLO topical report has been added to Specification 
6.9.1.12, for both units to ensure the analytical methods used to 
determine the core operating limits are consistent with those 
previously approved by the NRC. The proposed changes do not change 
or alter the design assumptions for the systems or components used 
to mitigate the consequences of an accident. Use of ZIRLO fuel rod 
material does not adversely affect fuel performance or impact 
nuclear design methodology. Therefore, accident analysis results are 
not impacted.
    The operating limits will not be changed and the analysis 
methods to demonstrate operation within the limits will remain in 
accordance with NRC approved methodologies. Other than the changes 
to the fuel assemblies, there are no physical changes to the plant 
associated with this technical specification change. A safety 
analysis will continue to be performed for each cycle to demonstrate 
compliance with all fuel safety design bases.
    VANTAGE 5H fuel assemblies with ZIRLO fuel rods meet the same 
fuel assembly and fuel rod design bases as other VANTAGE 5H fuel 
assemblies. In addition, the 10 CFR 50.46 criteria are applied to 
the ZIRLO fuel rods. The use of these fuel assemblies will not 
result in a change to the reload design and safety analysis limits. 
Since the original design criteria are met, the ZIRLO fuel rods will 
not be an initiator for any new accident. The fuel rod material is 
similar in chemical composition and has similar physical and 
mechanical properties as Zircaloy-4. Thus, the fuel rod integrity is 
maintained and the structural integrity of the fuel assembly is not 
affected. ZIRLO improves corrosion performance and dimensional 
stability. No concerns have been identified with respect to the use 
of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel 
rods.
    The dose predictions in the safety analyses are not sensitive to 
the fuel rod material used; therefore, the radiological consequences 
of accidents previously evaluated in the safety analysis remain 
valid. A reload analysis is completed for each cycle, in accordance 
with NRC approved methodologies. Therefore, the proposed change does 
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?
    VANTAGE 5H fuel assemblies with ZIRLO fuel rods satisfy the same 
design bases as those used for other VANTAGE 5H fuel assemblies. All 
design and performance criteria continue to be met and no new 
failure mechanisms have been identified. The ZIRLO fuel rod material 
offers improved corrosion resistance and structural integrity.
    The proposed changes do not affect the design or operation of 
any system or component in the plant. The safety functions of the 
related structures, systems, or components are not changed in any 
manner, nor is the reliability of any structure, system, or 
component reduced. The changes do not affect the manner by which the 
facility is operated and do not change any facility design feature, 
structure, or system. No new or different type of equipment will be 
installed. Since there is no change to the facility or operating 
procedures, and the safety functions and reliability of structures, 
systems, or components are not affected, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The use of Zircaloy-4, ZIRLO, or stainless steel filler rods in 
fuel assemblies will not involve a significant reduction in the 
margin of safety because analyses using NRC approved methodology 
will be performed for each configuration to demonstrate continued 
operation within the limits that assure acceptable plant response to 
accidents and transients. These analyses will be performed using NRC 
approved methods that have been approved for application to the fuel 
configuration.
    Use of ZIRLO as fuel rod material does not change the VANTAGE 5H 
reload design and safety analysis limits. The use of these fuel 
assemblies will take into consideration the normal core operating 
conditions allowed in the technical specifications. For each reload 
core, the fuel assemblies will be evaluated using NRC approved 
reload design methods, including consideration of the core physics 
analysis peaking factors and core average linear heat rate effects.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the UFSAR [Updated 
Final Safety Analysis Report] or any plant technical specification 
BASES.

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

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

    Date of amendment request: July 17, 1996.
    Description of amendment request: The proposed amendment would 
reflect that the name of Louisiana Power & Light Company, which is 
licensed to own and possess Waterford 3, has been changed to Entergy 
Louisiana, Inc.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    The proposed change documents changing the legal name of the 
company. The proposed change will not affect any other obligations. 
The company will still own all of the same assets, they still serve 
the same customers, and all existing obligations and commitments 
will continue to be honored. Therefore, the proposed change will not 
involve a significant increase in the probability or consequences of 
any accident previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No
    The administrative changes in the Operating License requirements 
do not involve any change in the design of the plant. Therefore, the 
proposed change will not create the possibility of a new or 
different

[[Page 17232]]

kind of accident from any accident previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No
    The proposed change is administrative in nature and does not 
reduce the level of safety imposed by any current requirement. 
Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: October 16, 1996.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) action requirements 3.2.1 and 3.2.4 
and their associated surveillance requirements to extend the allowable 
time for the Core Operation Limit Supervisory System (COLSS) to be out 
of service by monitoring for adverse trends in the linear heat rate 
(LHR) and departure from nucleate boiling (DNBR) limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    The proposed change does not modify the requirement to operate 
within the alternate LHR and DNBR limits nor does it modify the 
actual LHR or DNBR limits themselves. In the case of exceeding a 
COLSS calculated [power operating limit] POL, Entergy agrees that 
corrective action should be initiated promptly to bring the LHR and 
DNBR within their respective limits and, in this case, a 15 minute 
time limit is appropriate. However, in the case of exceeding a [core 
protection calculator] CPC calculated operating limit following the 
loss of COLSS, it is clear that simply because COLSS execution was 
lost does not mean that the plant is operating outside the range of 
conditions assumed in the Chapter 15 Safety Analysis and, in this 
case, a 15 minute time limit is not appropriate. An increase from 2 
hours to 8 hours to regain the monitoring capabilities of COLSS 
would not significantly increase the probability of exceeding the 
actual LHR or DNBR power operating limits since the increase in 
COLSS out-of-service time will be compensated for by monitoring for 
adverse trends of the important CPC calculated parameters (DNBR 
Margin and LHR). Further, since the proposed change will result in 
maintaining steady-state conditions while monitoring for adverse 
trends, it will be easier for the operators to detect any abnormal 
occurrence that has the potential to degrade either the LHR or the 
DNBR.
    The primary consideration in extending the COLSS out of service 
time limit is the remote possibility of a slow, undetectable 
transient that degrades the LHR and/or DNBR slowly over the 8 hour 
period and is then followed by an [anticipated operational 
occurrence] AOO or an accident. The parameters normally monitored by 
COLSS which have the potential for degrading the LHR and DNBR if no 
corrective action is taken are: Reactor Coolant System (RCS) flow 
rate, axial and radial power distributions, core inlet temperature, 
core power, RCS pressure and azimuthal tilt. Of these parameters, 
core inlet temperature, core power, and RCS pressure are easily 
monitored by the plant operators using various safety-grade, 
Redundant Control Room indications and, therefore, changes in these 
parameters are readily apparent. Further, operating experience at 
Waterford 3 and other [Combustion Engineering] CE nuclear steam 
supply systems using the same reactor coolant pumps (RCPs) as 
Waterford has shown that measurable changes in RCP Ps 
(which COLSS uses to calculate RCS flow) are very rare and when they 
do occur, involve abrupt step changes in flow which are readily 
apparent; hence, the probability of a slow degradation in the RCS 
flow rate is exceedingly small. Thus, the parameters that 
comparatively (although still remote) pose the most potential for a 
degradation in the core thermal margin when COLSS is out of service 
relate to the axial and radial core power distributions and the 
azimuthal tilt. These parameters are discussed below.
    Axial xenon oscillations are a normal consequence of the 
Waterford 3 core design, particularly near the end of core life. As 
a result, Waterford 3 operations personnel are instructed, per 
operating procedure OP-10-001, General Plant Operations, to maintain 
strict control over the axial power shape in the core. Although the 
primary reason for axial shape control is to maintain an even fuel 
burnup throughout the core, it also results in maintaining the axial 
power shapes well within the limits assumed in the safety analysis. 
Typically, axial shape control practiced at Waterford 3 maintains 
the axial shape index (ASI) within 0.05 ASI units of the equilibrium 
shape index (ESI), which is normally very near 0.0.
    Hypothetically, the most severe situation which could be 
postulated to occur, although again remote, would be if COLSS 
execution was lost just when the plant operators were ready to take 
manual action to return the ASI value to within the ESI + 0.05 
control band. Since a full xenon oscillation takes approximately 26 
hours, there would be about 6 hours from the time that control 
action would normally be taken to the time that the ASI reached its 
peak value (i.e., it takes one quarter cycle for the ASI to travel 
from its ESI value to its peak value). Since abnormal operating 
procedure OP-901-501, PMC or Core Operating Limit Supervisory System 
Inoperable, will be revised to require the CPC calculated LHR and 
DNBR trends to be monitored every 15 minutes (see below), any 
significant change in the axial shape index will be apparent through 
a change in these CPC calculated values. Hence, due to the attention 
given the axial power distribution, both when COLSS is in service as 
well as when COLSS is out of service it is very improbable that a 
change in ASI during eight hours of steady-state operation with 
COLSS out of service could be either undetected or lead to a 
condition that placed the reactor outside the range of initial 
conditions that were assumed in the safety analysis.
    With regards to azimuthal tilt, there is very rarely any 
significant change in this parameter as long as all [control element 
assembly] CEAs are properly aligned. The only real contributor to a 
rapid increase in azimuthal tilt would be an inadvertent CEA drop; 
however, since the probability of a CEA drop is very low, the 
likelihood of this event occurring within the eight hour time limit 
is even lower. In the unlikely event that a CEA drop did occur, the 
Control Element Assembly Calculators (CEACs) provide a safety-grade, 
redundant means of alerting the operators that corrective action is 
necessary. Thus, the potential for a degradation in azimuthal tilt 
during eight hours of steady-state operation following the loss of 
COLSS is both highly unlikely and relatively easy to detect using 
instrumentation already available in the Control Room.
    As previously stated, upon approval of the proposed change plant 
personnel will revise abnormal operating OP-901-501, PMC or Core 
Operating Limit Supervisory System Inoperable, to monitor for 
adverse trends of the CPC calculated values of LHR and DNBR. 
Currently, this procedure requires that the monitoring frequency for 
LHR and DNBR be increased to once every 15 minutes on a loss of 
COLSS.
    Extending the time to restore the CPC calculated LHR and DNBR to 
within the acceptable operating range from 2 hours to 8 hours is 
being proposed to assure that COLSS can be restored thus decreasing 
the probability of an avoidable challenge to the reactor protection 
system (RPS) during a power reduction. It is possible that the 
required power reductions may exceed 25% near the end of the fuel 
cycle. These large power reductions result in a rapid increase in 
xenon concentration, changes in ASI, and a subsequent decrease in 
cold leg temperature (T-cold) that may be difficult to control. 
Accordingly, given the potential for

[[Page 17233]]

power reductions of this magnitude, it is appropriate to extend the 
time allowed to restore COLSS so that a power reduction may be 
unnecessary.
    Taken in total, the proposed changes will reduce the number of 
potentially unnecessary power reductions by allowing more time for 
COLSS to be restored along with the advantages of trend monitoring 
in detecting an adverse trend expeditiously. The proposed change 
will result in significant operational benefits while continuing to 
maintain a high degree of confidence that the core conditions remain 
well within the range of values assumed in the safety analysis.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No
    The proposed change does not alter the current power operating 
limits nor does it involve any changes to COLSS or CPC software. 
There has been no physical change to plant systems, structures or 
components nor will the proposed change affect the ability of any of 
the safety-related equipment required to mitigate AOOs or accidents. 
The only significant change associated with the proposed amendment 
involves changes to the operating procedures used when COLSS is out-
of-service. All revisions to operating procedures will be reviewed 
and approved by appropriate plant personnel as required by the 
Administrative Controls (Section 6) in the Waterford 3 Technical 
Specifications. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No
    The intent of [limiting conditions for operation] LCOs 3.2.1 and 
3.2.4 is to maintain the reactor within the range of initial 
conditions that was assumed in the Safety Analysis. Maintaining the 
LHR within the specified range ensures that in the event of a LOCA, 
the fuel cladding temperature will not exceed the 2200 deg.F limit 
imposed by 10CFR46 [10 CFR Part 46]. Maintaining the DNBR within the 
specified range ensures that no AOO will result in a violation of 
the [Specified Acceptable Fuel Design Limits] SAFDLs and that no 
postulated accident will result in consequences more severe than 
those described in Chapter 15 of the [Final Safety Analysis Report] 
FSAR. Since there has been no change to the requirement to operate 
the reactor within the LHR and DNBR limits and no change to the 
actual LHR and DNBR limits themselves, the accident analyses 
described in Chapter 15 of the FSAR will not be affected and will 
therefore remain bounding.
    The proposed change will reduce the number of potentially 
unnecessary power reductions along with the rate at which the power 
reductions are accomplished. Maintaining steady-state conditions for 
up to eight hours after the loss of COLSS while monitoring the CPC 
LHR/DNBR for trends, provides plant personnel with a reasonable 
period of time to return COLSS to service while continuing to 
maintain a high degree of confidence that the core conditions remain 
well within the range of values assumed in the safety analysis. In 
fact, monitoring for trends in LHR and DNBR Margin increases the 
margin of safety by allowing the anticipation of degradation in LHR 
or DNBR Margin. Moreover, by reducing the number of plant transients 
there will be an attendant reduction in probability of an AOO and 
subsequent RPS actuation. Therefore, the proposed change will not 
involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 
L Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: October 16, 1996.
    Description of amendment request: The following changes to the 
Waterford Steam Electric Station, Unit 3, Technical Specifications are 
proposed: 1) Relocation of certain administrative controls to the 
Quality Assurance Program Manual (QAPM) as described in Nuclear 
Regulatory Commission Administrative Letter 95-06, ``Relocation of 
Technical Administrative Controls related to Quality Assurance''; 2) 
Change shift coverage from 8-hour day, 40-hour weeks to an option of 8 
or 12 hour days and nominal 40-hour weeks; 3) Make certain editorial 
changes to the titles of certain organizational positions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No
    The conditions as they exist in the present Technical 
Specifications do not have an affect on either the probability or 
consequences of a previously evaluated accident. These changes also 
will have no impact to increase either the probability or 
consequences of a previously evaluated accident.
    The proposed changes will have no affect on design basis 
accidents nor will the change directly affect any material condition 
of the plant that could directly contribute to causing or mitigating 
the effects of an accident.
    Therefore, the proposed changes will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No
    The proposed changes will not alter the operation of the plant 
or the manner in which it is operated. The changes do not involve a 
design change and do not introduce any new failure modes.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in margin of safety?
    Response: No
    The proposed changes are administrative in nature and affect 
only Section 6.0 of the Technical Specifications. The Waterford 3 
margins of safety are defined in Sections 2 through 5 and are 
unaffected by these changes. Moving the reviews from the TS to the 
QAPM will have no affect on the margin of safety because reviews 
will still be performed. The only difference is the reviews will be 
administratively controlled by the QAPM. The QAPM is controlled by 
10CFR50.54 so no changes can be made which would lessen these 
commitments (i.e., remove or reduce the requirement for procedure 
reviews) without prior NRC approval.
    Changing from an 8 hour to an 8 or 12 hour shift will not have 
an adverse impact on personnel performance. The NRC study documented 
in NUREG CR-4248 has identified that personnel errors have decreased 
and productivity has increased where this change has been 
implemented.
    Therefore, the proposed changes will not involve a significant 
reduction in a margin of safety.

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

[[Page 17234]]

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed change modifies 
Technical Specification 3/4.5.2, ``ECCS Subsystems Modes 1, 2, and 3.'' 
The proposed change adds a surveillance requirement to verify the 
Emergency Core Cooling System (ECCS) piping is full of water at least 
once per 31 days. A change to the Technical Specification Basis 3/4.5.2 
and 3/4.5.3 has been included to support this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    The proposed change will not affect the assumptions, design 
parameters, or results of any accident previously evaluated. The 
proposed change does not add or modify any existing equipment. The 
proposed change adds a new surveillance requirement which will 
minimize the likelihood of a pressure transient occurring during 
system startup and provide increased assurance that the ECCS will 
perform its design basis function when needed. The new [low pressure 
safety injection] LPSI and [high pressure safety injection] HPSI 
vent valves which may be manipulated during this surveillance will 
be administratively controlled and will be locked close when not in 
use to prevent the possibility of a flow diversion. This 
surveillance requirement is consistent with NUREG 1432.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    While new vent lines are being installed under 10CFR50.59, this 
proposed change adds only a new surveillance requirement to 
Technical Specification 3/4.5.2 and therefore does not involve 
modifications to any existing equipment. The new vent valves, when 
required, will be operated and controlled in the same manner as 
existing LPSI and HPSI vent valves. The new LPSI and HPSI vent 
valves will be administratively controlled and will be locked close 
when not in use. This surveillance requirement is consistent with 
NUREG 1432.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    The functionality of ECCS is maintained such that it is capable 
of performing its design function as assumed in the Updated Final 
Safety Analysis Report. Verifying the ECCS is full of water at least 
once per 31 days will minimize the likelihood of a pressure 
transient occurring during system startup and provide increased 
assurance that the ECCS will perform its design basis function when 
needed. This surveillance requirement is consistent with NUREG 1432.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed change modifies 
Technical Specification (TS) surveillance requirements 4.5.2.d.3 and 
4.5.2.d.4. The proposed change specifies granular trisodium phosphate 
dodecahydrate (TSP), increases the minimum required amount of TSP that 
is maintained in containment during power operation, and adjusts the 
TSP sampling requirement accordingly. A change to the TS Basis 3/4.5.2 
has been included to support this change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will operation of the facility in accordance with this 
proposed change involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    Response: No.
    Granular trisodium phosphate dodecahydrate is stored in the 
containment lower level to raise the pH of the sump and spray water 
following a LOCA. As the pH of the water increases, more radioactive 
iodine is kept in solution and the amount of airborne radioactive 
leakage is decreased. This also lessens the potential for boric acid 
solution reacting with galvanized metal in containment to release 
hydrogen. An additional advantage of a higher pH is the beneficial 
reduction in chloride stress corrosion cracking of metal components 
in the containment following an accident.
    This chemical is an accident mitigator, not an accident 
initiator in that it is not used until after an accident has 
occurred. At the time it goes into solution, the accident has 
occurred, containment spray has been activated and water has 
collected in the sump. Therefore, increasing the Technical 
Specification minimum amount verified to be in containment or 
changing the sample solution and sample size will not involve a 
significant increase in the probability of an accident previously 
evaluated.
    At the time TSP goes into solution, the accident has occurred, 
containment spray has been activated and water has collected in the 
containment sump. At Waterford 3, the iodine partition factor is a 
constant 50% and does not vary with pH as allowed in the Standard 
Review Plan (SRP) revision 1. The curve in SRP 6.5.2 revision 1 
allows a partition factor of at least 50% for containment water at a 
pH of 6.5 or less. The partition factor increases as pH rises. But, 
the curve is based on sodium hydroxide which is much more reactive 
than TSP. Therefore, increasing the Technical Specification minimum 
amount verified to be in the containment, and corresponding sample 
size, will not involve any significant increase in the consequences 
probability of an accident because no credit is taken for reducing 
the amount of volatized iodine normally associated with a 7.0 pH 
solution.
    Therefore, the proposed change will not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    2. Will operation of the facility in accordance with this 
proposed change create the possibility of a new or different type of 
accident from any accident previously evaluated?
    Response: No.
    The addition of more TSP does not represent a significant change 
in the configuration or operation of the plant. Trisodium phosphate 
dodecahydrate is currently present in the containment lower level. 
Design Change 3491 which increases the storage capacity of the TSP 
storage baskets was evaluated in accordance with 10 CFR 50.59 and 
found not to involve an unreviewed safety question.

[[Page 17235]]

    Boric acid acts as a buffer to prevent the pH from rising above 
approximately 8.1 as TSP is dissolved. An internal study (EC-S96-013 
revision 0) has shown that given the ``ratio of grams of TSP to 
liters of 3000 ppm boron solution'' stays less than 5.6, TSP cannot 
increase pH above 8.2. As pH increases, components composed of 
aluminum, zinc, or copper become vulnerable to corrosion. Branch 
Technical Position MTEB 6-1 implies that a solution pH greater than 
7.5 enhances the chance for hydrogen generation as a result of 
aluminum corrosion. Waterford 3 administratively limits the amount 
of aluminum in containment to minimize the amount of hydrogen 
expected during a DBA. Zinc is a component of the paint applied to 
surfaces inside containment. The hydrogen recombiner design basis 
includes 464 square feet (1040 pounds) of aluminum and 419,300 
square feet (17,252 pounds) of metallic zinc. Estimates of the 
amount of hydrogen produced by the aluminum assumes that the 
corrosive agent is sodium hydroxide--a much more active chemical 
than is TSP. Thus, the amount of hydrogen expected in the FSAR for 
the hydrogen recombiner bounds what would actually be produced by 
TSP even at a pH of approximately 8.1.
    The 4.5.2.d.3 proposed TSP to boron ratio assures that pH cannot 
rise above 8.1 as long as post accident in-containment boric acid 
solution concentration is no greater than 3011 ppm boron and no less 
than 1504 ppm boron. The main variable in post accident 
concentration (the difference between 1504 and 3011) is the 
concentration in the RCS at the time of the accident.
    Therefore, the proposed change will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Will operation of the facility in accordance with this 
proposed change involve a significant reduction in a margin of 
safety?
    Response: No.
    Trisodium phosphate dodecahydrate is stored in the containment 
lower level to raise the pH of the sump and spray water following a 
LOCA. As the pH of the water increases, more radioactive iodine is 
kept in solution and the amount of airborne radioactive leakage is 
decreased. A neutral pH also reduces the hydrogen generation from 
the corrosion of the galvanized materials in containment. An 
additional advantage of a higher pH is the beneficial reduction in 
chloride stress corrosion cracking of metal components in the 
containment following an accident.
    Technical Specification 4.5.2.d.3 requires verification that a 
minimum volume of TSP is contained in the storage baskets in 
containment. Nine previous runs of surveillance requirement 
4.5.2.d.4 (and similar tests) showed that the TSP actually used in 
the plant properly neutralized a sample of water borated within RWSP 
boron concentration limits. Boron concentrations of eight of the 
sample solutions used in these tests ranged from 1753 ppm to 2217 
ppm and resulted in a pH of 7.02 or greater. (The boron 
concentration of one test performed in 1986 was unavailable.) The 
ratio 4 grams to 4 liters is the amount of TSP needed to bring the 
solution to a pH of at least 7.0 given that the solution is in the 
1753 to 2217 ppm Boron range.
    The amount of TSP in containment currently is adequate assuming 
that RCS boric acid concentration stays below 454 ppm. However, the 
fuel cycle is nearly over and a restart with a refreshed core would 
require substantially more boric acid. We expect that the 
containment water would reach approximately 2400 ppm under ideal 
circumstances during cycle 9. During cycle 10, boron concentration 
in containment could reach 3011 under those same ideal conditions. 
As the maximum boron concentration increases, there is a non-linear 
increase in the amount of TSP needed to raise solution pH to 7.0. 
Thus, we request that the minimum amount of TSP in containment 
required by 4.5.2.d.3 to be increased from 97.5 cubic feet to 380 
cubic feet. This change also proposes to adjust the 4.5.2.d.4 
specified increase that sample solution and the TSP sample size 
accordingly. This change will ensure the safety injection 
containment sump, when filled with water, will have an acceptable pH 
following a LOCA. The test will not only demonstrate that TSP is in 
the baskets but also shows that the amount of TSP in containment can 
neutralize the solution expected in containment during any DBA.
    Therefore, the proposed change will not involve a significant 
reduction in a margin of safety. The amount of iodine kept in 
solution during a DBA is limited to 50%. Note, the pH scale is 
logarithmic so that the amount of TSP needed to raise pH to 7.0 is 
more than three times the amount needed to reach 6.5. Furthermore, 
the amount of hydrogen generated during a DBA is over estimated by 
the analysis when it used sodium hydroxide as the corrosive agent 
rather than TSP.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn, 1400 
L Street N.W., Washington, D.C. 20005-3502.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: May 17, 1996, as supplemented March 17, 
1997.
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Section 3/4.4.5, Steam Generators, 
3/4.4.6, Reactor Coolant System Leakage, and associated Bases to allow 
the installation of tube sleeves as an alternative to plugging to 
repair defective steam generator tubes. The proposed change would also 
specify the Westinghouse topical reports to be used for sleeve design 
and inspection, and identify the inspection sample size for repaired 
tubes. This application was previously published in the Federal 
Register on May 29, 1996, (61 FR 26938).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Listing the specific Westinghouse topical reports in the 
TS binds the South Texas Project (STP) to the sleeve design and 
inspection techniques identified in that revision of the topical 
report. Any changes to sleeve design or inspection technique would 
require a separate TS amendment.
    New TS Table 4.4-3, Steam Generator Repaired Tube Inspection, 
identifies the inspection sample size for steam generator tubes that 
have already been repaired. This table simply identifies inspection 
criteria and associated actions for repaired tubes and does not 
increase 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. Implementation of laser welded sleeving maintains overall 
tube bundle structural and leakage integrity conditions. Providing 
specific Westinghouse topical report references in the TS only 
serves to identify which sleeve design and inspection techniques are 
being employed at STP. Likewise, the addition of Table 4.4-3 
clarifies the expected inspection samples for previously repaired 
tubes. The addition of Table 4.3-3 provides assurance that 
previously repaired tubes will be inspected at regular intervals and 
appropriate action taken if the tube is found defective. Neither of 
these additions to the TS will create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. Both of these changes are being added to 
clarify the STP steam generator tube inspection program and provide 
more specific detail regarding steam generator tube inspection 
samples and inspection techniques. By requiring inspection of 
previously repaired tubes, the margin of safety is increased rather 
than decreased.

[[Page 17236]]

    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: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: William D. Beckner.

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

    Date of amendment request: January 28, 1997.
    Description of amendment request: The proposed amendment would 
relocate the details of Technical Specification (TS) Section 6.2.3 on 
the Independent Safety Engineering Group (ISEG) from the Administration 
Controls section of the TSs and place these details in the Updated 
Final Safety Analysis Report (UFSAR) for South Texas Project, Units 1 
and 2. This relocation is administrative only, and would not render any 
changes to the existing plant philosophy toward the ISEG or any safety 
analysis. Section 6.2.3 would be deleted from the TSs and removed from 
the table of contents for Administrative Controls. Currently UFSAR 
Section 13.4.2.2 describes the ISEG, but not in the detail as the 
current TSs.
    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 changes move details from the Technical 
Specifications [TSs] to the Updated Final Safety Analysis Report 
(UFSAR). The changes do not result in any hardware or operating 
procedure changes. The details being removed from the Technical 
Specifications [TSs] are not assumed to be an initiator of any 
analyzed event. The UFSAR, which will contain the removed Technical 
Specification [TS] details, will be maintained using the provisions 
of 10 CFR 50.59 and is subject to the change control process in the 
Administrative Controls Section of the Technical Specifications 
[TSs]. [In addition] any changes to the UFSAR will be evaluated per 
10 CFR 50.59, no increase in the probability or consequences of an 
accident previously evaluated will be allowed without prior NRC 
[Nuclear Regulatory Commission] approval. Therefore, the 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 proposed changes move details from the technical 
Specifications [TSs] to the Updated Final Safety Analysis Report 
(UFSAR). The changes will not alter the plant configuration (no new 
or different type of equipment will be installed) or make changes in 
methods governing plant operation. The changes will not impose 
different requirements, and adequate control of information will be 
maintained. The changes will not alter assumptions made in the 
safety analysis and licensing basis. Therefore, the changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes move detail from the Technical 
Specifications [TSs] to the Updated Final Safety Analysis Report 
(UFSAR). The changes do not reduce the margin of safety since the 
relocation of details [is an administrative action and] has no 
impact on any safety analysis assumptions. In addition, the detail 
transposed from the Technical Specifications [TSs] to the UFSAR are 
the same as the existing Technical Specification [TS] [6.2.3]. [In 
addition] any future changes to the FSAR will be evaluated per the 
requirements of 10 CFR 50.59, no reduction in a margin of safety 
will be allowed without prior NRC approval. [Therefore, the licensee 
concluded that the changes 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges, Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: William D. Beckner.

Northeast Nuclear Energy Company, et al., Docket No. 50-245, Millstone 
Nuclear Power Station, Unit No. 1, New London, Connecticut

    Date of amendment request: February 7, 1997.
    Description of amendment request: The proposed Technical 
Specification changes would clarify and/or modify instrument 
calibration, functional, and response time requirements for resistance 
temperature detector and thermocouple testing. Also, certain 
definitions would be clarified and/or modified using applicable wording 
from NRC's NUREG-1433, ``Standard Technical Specifications,'' Revision 
1, and industry recommendations. Additionally, the change would 
relocate the reactor protection system logic response time value 
utilizing the guidance provided by NRC's Generic Letter 93-08, 
``Relocation of Technical Specification Tables of Instrument Response 
Time Limits,'' with the exception of relocating the value to the 
Technical Specifications Bases Section instead of the Updated Final 
Safety Analysis Report. The proposed amendment is intended to clarify 
instrumentation surveillance requirements, thereby helping to ensure 
proper testing of safety-related components.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Pursuant to 10 CFR 50.92, NNECO [Northeast Nuclear Energy 
Company] has reviewed the proposed changes and concludes that the 
changes do not involve a significant hazards consideration (SHC) 
since the proposed changes satisf[y] the criteria in 10 CFR 
50.92(c). That is, the proposed changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment continues to ensure the surveillance 
requirements satisfy the licensing basis. The current TS [technical 
specifications] definition for Instrument Functional Test requires 
injection of a simulated signal into the primary sensor to verify 
proper response. Current TS exempt the sensors of specific 
instrument channels where it is not practical to include them within 
the functional test boundaries. Some examples of these exemptions 
include neutron monitoring system, turbine control valve fast 
closure, and standby gas treatment initiation radiation monitors. In 
these cases, TS permit the performance of the functional test by 
injection of a simulated electrical signal into the measurement 
channel. The proposed definition, which is consistent with the STS 
[standard technical specifications]

[[Page 17237]]

definition, for CHANNEL FUNCTIONAL TEST requires injection of the 
simulated signal ``as close to the sensor as practicable.'' 
Therefore, the proposed definition is consistent with the current TS 
definition and its exemptions. The primary sensor is the transmitter 
or switch or radiation monitor. The definition does not include 
sensing elements such as radiation detectors, flow elements, 
acceleration relays or reference legs.
    This change will allow the channel functional test to be 
performed by means of any series of sequential, overlapping, or 
total channel steps and aligns this methodology with industry 
practice. This change does not affect accident precursors and thus 
does not involve a significant increase in the probability of an 
accident previously evaluated. The proposed change will allow a 
simulated or actual signal to be used to perform an Instrument or 
Channel Functional Test. This change does not impose a requirement 
to create an actual signal, nor does it eliminate any restriction on 
producing an actual signal. While creating an ``actual'' signal 
could increase the probability of an event, existing procedures (and 
the 10 CFR 50.59 control of revisions to them) dictate the 
acceptability of generating this signal. The proposed change does 
not affect the procedures governing plant operations or the 
acceptability of creating these signals; it simply would allow such 
a signal to be utilized in evaluating the acceptance criteria for 
the Instrument or Channel Functional Test requirements. Therefore, 
the change does not involve a significant increase in the 
probability of an accident previously evaluated. Because the method 
of initiation will not affect the acceptance criteria of the 
Instrument or Channel Functional Test, the change does not involve a 
significant increase in the consequences of an accident previously 
evaluated.
    Minor word differences from STS are required to provide 
consistency with current TS wording and support the current 
licensing basis. These minor word differences including Industry/
TSTF [Technical Specification Task Force] Standard Technical 
Specification Change Traveler (TSTF-64) do not alter the meaning of 
instrument testing in the STS or change the current licensing basis.
    Moving the RPS [Reactor Protection System] Logic Response Time 
LCO [Limiting Condition of Operation] description to the TS 
definition section is an administrative change and does not alter 
the original intent or licensing basis.
    Relocation of the RPS Logic Response Time value from the TS to 
the Bases section involves the use of an alternate regulatory 
process for controlling the instrument response time limit. The 
change does not introduce any new modes of plant operation, make any 
physical changes, alter any operational setpoints, or change the 
surveillance requirements. Any change in the RPS logic response time 
value would be evaluated pursuant to the requirements of 10 CFR 
50.59.
    The surveillance section editorial change does not alter the 
meaning of surveillance applicability. Providing RPS Logic Response 
Time surveillance frequency and applicable trip functions ensures 
proper testing of RPS components and is consistent with industry 
practice. An evaluation completed by GE [General Electric] verified 
the applicable RPS trip functions that require a specific logic 
response time using the current accident analysis as the basis. For 
trip functions where no explicit credit is taken in the safety 
analysis, the measurement of logic response time is not important, 
and therefore, not warranted. In addition, we have concluded, that 
instrumentation response time requirements (specified limits) other 
than RPS logic are not important to test, especially considering the 
long delays already accounted for in the accident analyses 
associated with the start of emergency power sources, ECCS 
[Emergency Core Cooling System] components, and containment 
isolations, and that the non-RPS logic response times, including 
response times of other instrumentation such as radiation monitors, 
are not part of the Millstone Unit No. 1 licensing basis. The 
sensors associated with all TS instrumentation are functionally 
tested and calibrated to ensure proper operation.
    No physical change is being made to instrument channels, or to 
any systems or component that interfaces with the instrumentation 
channels, therefore there is no change in the probability or 
consequences of any accident analyzed in the UFSAR [Updated Final 
Safety Analysis Report].
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change does not result in any design or physical 
configuration changes to the instrumentation channels. Operation 
incorporating the proposed change will not impair the 
instrumentation channels from performing as provided in the design 
basis.
    Changing the TS to be consistent with current industry practice 
adopted in STS will help to prevent unnecessary removal and 
potential damage of the temperature detectors (for sensor 
calibration). Clarification of RPS Logic Response Time testing 
requirements consistent with the current licensing basis will ensure 
proper testing of safety-related components.
    Wording changes to Instrument Calibration and Functional Test 
definitions do not involve a physical modification to the plant. The 
injection of an actual or simulated signal as close to the sensor as 
practical minimizes the likelihood of any transients.
    Minor word differences from STS are required to provide 
consistency with current TS wording and support the current 
licensing basis. These minor word differences, including Industry/
TSTF Standard Technical Specification Change Traveler (TSTF-64), do 
not alter the meaning of instrument testing in the STS or change the 
current licensing basis.
    Moving the RPS Logic Response Time LCO description to the TS 
definition section is an administrative change and does not alter 
the current licensing basis.
    Relocation of the RPS Logic Response Time value involves the use 
of an alternate process for controlling the instrument response time 
limits. Therefore, the above change does not introduce any accident 
initiators as it does not involve any new modes of plant operation, 
make any physical changes, alter any operational setpoints, or 
change the surveillance requirements.
    The surveillance section editorial change does not alter the 
meaning of surveillance applicability. Providing RPS Logic Response 
Time surveillance frequency and applicable trip functions ensures 
proper testing of RPS components and is consistent with industry 
practice.
    Since the proposed changes in the Technical Specifications do 
not adversely impact the reliability of the RPS and other automatic 
actuations, no new or different kind of accident is created.
    3. Involve a significant reduction in a margin of safety.
    Because the proposed change does not involve the addition or 
modification of plant equipment, is consistent with the existing 
Technical Specifications, current industry practices as outlined in 
NUREG 1433, ``Standard Technical Specifications GE Plants, BWR/4,'' 
Revision 1, and with the current design and licensing basis of the 
Protective Instrumentation systems including the accident analysis, 
no action will occur that will involve a significant reduction in a 
margin of safety.
    The proposed change to allow the use of an actual signal in 
addition to the existing requirement, which limits use to a 
simulated signal, will not affect functional test acceptance 
criteria. Therefore, the proposed change does not adversely affect 
the reliability of the RPS or other automatic actuation and does not 
involve a significant reduction in a margin of safety.
    Relocation of the RPS Logic Response Time value from the TS to 
the Bases section involves the use of an alternate regulatory 
process for controlling the instrument response time limit. Any 
change in the RPS logic response time value would be evaluated 
pursuant to the requirements of 10 CFR 50.59.
    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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and the Waterford Library, ATTN: Vince Juliano, 49 
Rope Ferry Road, Waterford, CT 06385.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Deputy Director: Phillip F. McKee.

[[Page 17238]]

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: November 25, 1996, as supplemented 
December 12, 1996.
    Description of amendment request: The proposed amendment would make 
changes to Section 2.1.A for the Safety Limit Minimum Critical Power 
Ratio (SLMCPR) and to Section 3.11.C for the Operating Limit Minimum 
Critical Power Ratio (OLMCPR). The proposed change to Section 2.1.A 
revises the SLMCPR value from 1.07 to 1.08 for two recirculation pump 
operation and from 1.08 to 1.09 for single loop operation. The proposed 
change to Section 3.11.C deletes the sentence that specifies the OLMCPR 
limit penalty for single recirculation loop operation and adds a 
statement that references the Core Operating Limits Report (COLR) as 
the source for this information.
    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 basis of the MCPR [minimum critical power ratio] Safety 
Limit calculation is to ensure that greater than 99.9% of all fuel 
rods in the core avoid transition boiling if the limit is not 
violated. The new SLMCPRs preserve the existing margin to transition 
boiling and fuel damage in the event of a postulated accident. The 
probability of fuel damage is not increased. The derivation of the 
revised SLMCPRs for Monticello for incorporation into the Technical 
Specification, and its [their] use to determine cycle-specific 
thermal limits, have been performed using NRC-approved methods as 
identified in Technical Specification 6.7.A.7.b. NSP [Northern 
States Power] methodology established OLMCPR such that integrity of 
the SLMCPR is maintained for the bounding analyzed transients. 
Additionally, GENE [General Electric Nuclear Energy] interim 
implementing procedures, which incorporate cycle-specific 
parameters, have been used. Based on the use of these calculations, 
the calculation of the revised SLMCPRs maintains the integrity of 
the safety limits and therefore cannot increase the probability or 
severity of an accident. The single loop OLMCPR evaluation was 
performed using NSP methodology approved by the NRC. Relocating the 
OLMCPR value to the COLR establishes appropriate control on a core 
operating limit which may vary from cycle to cycle because it is 
cycle dependent. Since OLMCPR is developed using procedures approved 
in the Technical Specifications, placing the OLMCPR in the COLR 
cannot result in a change not controlled by the Technical 
Specifications. The change does not affect failure modes of 
equipment, therefore, this amendment will not cause 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 
analyzed.
    The MCPR Safety Limit is a Technical Specification numerical 
value, designed to ensure that fuel damage from transition boiling 
does not occur as a result of the limiting postulated accident. It 
cannot create the possibility of any new type of accident. The new 
SLMCPRs have been calculated using NRC-approved methods and the 
OLMCPR values are more conservative. Additionally, interim 
procedures, which incorporate cycle-specific parameters, have been 
used. Therefore, the proposed Technical Specification change does 
not create the possibility of a new or different kind of accident, 
from any accident previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The MCPR Safety Limit is a Technical Specification numerical 
value, designed to ensure that fuel damage from transition boiling 
does not occur as a result of the limiting postulated accident. 
Increasing the SLMCPR and OLMCPR values results in an increase in 
the margin of safety to fuel failure, and does not affect other 
plant systems. Therefore, the proposed Technical Specification 
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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: November 20, 1996, as supplemented by 
letter dated February 20, 1997.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to allow the Vice President to 
designate the Safety Audit and Review Committee (SARC) Chairperson, to 
change the work hours limitation in accordance with guidance in GL 82-
12, ``Nuclear Power Plant Staff Working Hours;'' to change radioactive 
shipments record retention requirements to comply with recent 10 CFR 
Part 20 changes; to revise position titles to reflect organizational 
changes; and other editorial changes. The February 20, 1997, 
supplemental letter differs from the November 20, 1996, application 
which was noticed in the Federal Register on January 2, 1997 (62 FR 
131), in that the previous application did not propose changes to TS 
5.3, 5.5, 5.6, 5.7, and 5.11 reflecting recent organizational changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The changes requested are administrative in nature. Paragraph 
3.D was placed in the License by Amendment No. 155 to authorize 
Omaha Public Power District (OPPD) to increase the storage capacity 
of the FCS spent fuel pool. Amendment No. 155 stated that the TS as 
issued would be effective when the last new rack was installed. 
Since the last new rack was installed on August 8, 1994, Paragraph 
3.D is no longer necessary and should be deleted from the License.
    Table of Contents, Section 6.0, ``Interim Special Technical 
Specifications,'' Subsections 6.1 through 6.4 are proposed for 
deletion because all of the Specifications referred to have been 
deleted by previous Amendments.
    The revision proposed for TS 2.15 (Item 2C of Table 2-3 & Item 
1C of Table 2-4) will insert the correct terminology (Pressurizer 
Low/Low Pressure) into the Functional Unit description.
    The revision proposed for TS 5.2 will delete the specific 
working hours as stated and relocate these requirements to the 
Updated Safety Analysis Report (USAR). Overtime will remain 
controlled by plant administrative procedures with the USAR 
generally following the guidance of the NRC's Policy Statement on 
working hours contained in Generic Letter 82-12, ``Nuclear Power 
Plant Staff Working Hours.'' Specifying personnel working hours in 
TS does not meet any of the four criteria contained in 10 CFR 50.36 
for inclusion in the TS. Revisions to plant procedures containing 
these requirements are required to be evaluated in accordance with 
10 CFR 50.59. The proposed relocation is similar to recent 
Amendments issued to the Davis-Besse Nuclear Power Station and the 
San Onofre Nuclear Generating Station.
    The revision proposed for TS 5.5.2.2 will replace the specific 
title of the Chairperson of the Safety Audit and Review Committee

[[Page 17239]]

and replace it with ``Member as appointed by the Vice President.'' 
This will allow the flexibility to change chairmanship of the 
committee amongst the members.
    The revisions proposed to TS 5.3, 5.5, 5.6, 5.7, and 5.11 revise 
position titles and reporting responsibilities to reflect 
organizational changes. Qualifications for individuals in these 
positions meet or exceed the previous requirements.
    The revision to TS 5.10 concerning retention of records of 
radioactive shipments will update the TS to current 10 CFR 20 
requirements. Plant procedures already comply with current 10 CFR 20 
record retention requirements. The addition of the Section 5.0 title 
corrects a minor format discrepancy.
    These proposed revisions are administrative in nature. The 
proposed revisions have no effect on any initial assumptions or 
operating restrictions assumed in any accident, nor do these changes 
have any effect on equipment required to mitigate the consequences 
of an accident. Therefore the proposed revisions 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 proposed revisions correct minor errors, remove outdated 
information, are consistent with changes in organizational 
structure, 10 CFR Part 20, or the criteria contained in 10 CFR 
50.36. These changes will not result in any physical alterations to 
the plant configuration, changes to setpoint values, or changes to 
the application of setpoints or limits. No new operating modes are 
proposed as a result of these changes. 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 revisions listed above correct minor errors, remove outdated 
information, or are consistent with changes in organizational 
structure, 10 CFR Part 20, or the criteria contained in 10 CFR 
50.36. These changes will not result in any physical alterations to 
the plant configuration, changes to setpoint values, or changes to 
the application of setpoint or limits. Therefore the proposed 
changes do not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: Perry D. Robinson, Winston & Strawn, 1400 L 
Street, N.W., Washington, DC 20005-3502.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: February 26, 1997.
    Description of amendment request: The proposed amendment would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to revise TS 3/4.4.5 and 3.4.6.2, 
including associated Bases 3/4.4.5 and 3/4.4.6.2, to allow the 
implementation of steam generator (SG) tube voltage based repair 
criteria for outside diameter stress corrosion cracking (ODSCC) 
indications at tube-to-tube support plate (TSP) intersections. The 
allowed primary-to-secondary operational leakage from any one SG would 
be reduced from 500 gpd to 150 gpd.
    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.

Structural Integrity Considerations

    The structural criteria ensure that all indications subjected to 
voltage-based repair limits will be able to withstand pressure 
loading consistent with the criteria of NRC Regulatory Guide (RG) 
1.121.
    Tube burst criteria are inherently satisfied during normal 
operating conditions because of the proximity of the tube support 
plate (TSP). It is conservatively assumed that the entire crevice 
region is uncovered during the secondary side blowdown of a main 
steam line break (MSLB). Therefore, during a postulated MSLB 
accident, tube burst capability must exceed the RG 1.121 criterion 
requiring a margin of 1.43 times the steam line break pressure 
differential on tube burst.
    Based on the latest industry database, the RG 1.121 criterion is 
satisfied by bobbin coil indications of outside diameter stress 
corrosion cracking (ODSCC) with signal amplitudes less than 8.7 
volts. The latest NRC-approved database will be used for repair and 
analysis applications.
    Industry testing of model boiler and operating plant tube 
specimens for free-span tubing (no tube support plate (TSP) 
restraint) at room temperature conditions show typical burst 
pressures in excess of 5,000 psi for ODSCC indications with voltage 
measurements at or below 8.7 volts. This tube burst capability 
exceeds the RG 1.121 criterion.
    The lower voltage repair limit is conservatively defined to be 
2.0 volts in accordance with NRC Generic Letter (GL) 95-05, 
``Voltage-Based Repair Criteria for Westinghouse Steam Generator 
Tubes Affected by Outside Diameter Stress Corrosion Cracking,'' 
August 3, 1995. This 2.0 volt repair limit is very conservative 
because it contains a large safety margin, based on a structural 
limit of 8.7 volts. A maximum allowable upper repair limit (URL) is 
also established using the guidance of GL 95-05. The URL is 
calculated before each inspection by subtracting the NDE uncertainty 
and growth rate allowances from the current structural limit. The 
URL for near term inspections at DCPP Units 1 and 2 is expected to 
be about 5.0 volts. Bobbin indications greater than 2.0 volts and 
less than or equal to 5.0 volts that are confirmed by RPC will be 
repaired. Bobbin indications greater than 5.0 volts will be 
repaired.
    Following each inspection, burst probability analyses are 
performed for the end of cycle (EOC) distribution. In accordance 
with GL 95-05, the projected MSLB burst probability must be less 
than the threshold value of 1 x 10 x 2. Based on the relatively 
small number and voltages of ODSCC indications identified to date at 
DCPP Units 1 and 2, it is expected that the near term EOC 
conditional burst probability for a faulted SG will be much less 
than this threshold value, providing further assurance of acceptable 
structural integrity.

Leakage Considerations

    PG&E will implement reduced operational leakage limits as 
recommended in GL 95-05. PG&E will revise the TS to implement a 
maximum leakage rate of 150 gpd for any one SG to help preclude the 
potential for excessive leakage during power operation in Modes 1 
and 2. The TS has also been changed to specify that the 150 gpd leak 
limit is not necessarily a limiting condition for operation in Modes 
3 and 4. The 150 gpd leak rate per steam generator has been 
established for normal operation. This leakage rate provides added 
assurance against tube rupture at normal and faulted conditions. In 
Modes 3 and 4, there is less differential pressure across the tube 
and the potential source term from a tube failure is much less than 
in Modes 1 and 2. The operational leak rate monitoring program is a 
defense-in-depth measure that provides a means for identifying leaks 
during power operation to allow for repair before such leaks can 
result in tube failure. The leakage criteria ensure that for 
indications subjected to voltage-based repair criteria, induced 
leakage under worst-case MSLB conditions will not result in offsite 
and control room dose releases that exceed the applicable guideline 
values of 10 CFR 100 and GDC 19.
    Relative to the expected leakage during accident condition 
loadings, a postulated MSLB outside of containment, but upstream of 
the main steam isolation valve (MSIV), represents the most limiting 
radiological condition for implementation of voltage-based repair 
criteria. The steam generator tubes are subjected to an increase in 
differential pressure following a MSLB, resulting in a postulated 
increase in leakage

[[Page 17240]]

and associated offsite doses. Leakage following a MSLB bypasses 
containment.
    PG&E will calculate the primary-to-secondary leakage for 
degradation subjected to the voltage repair criteria under worst-
case postulated MSLB conditions. The leak rate will be compared to 
the maximum allowable leak rate limit of 12.8 gpm to ensure that a 
postulated MSLB occurring at EOC would not result in radiological 
consequences that are in excess of the applicable offsite and 
control room dose guidelines of 10 CFR 100 and GDC 19. Based on the 
relatively small number of ODSCC indications identified to date at 
DCPP Units 1 and 2, it is expected that the near term EOC predicted 
leak rates for a faulted SG will be much less than the maximum 
allowable leak rate limit.
    Therefore, based on the structural integrity and leakage 
considerations discussed above, 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.
    Implementation of the proposed voltage-based repair criteria for 
ODSCC at TSP intersections does not introduce any significant change 
to the plant design basis. Use of the criteria does not create a 
mechanism which could result in an accident in the free span because 
the repair criteria do not apply to tubes containing ODSCC located 
outside the thickness of the TSPs. Based on the burst probability 
acceptance limit of 1 x 10-2, it is expected that for all plant 
conditions, neither a single nor multiple tube rupture event would 
likely occur in a steam generator where voltage-based repair 
criteria have been applied.
    Steam generator tube integrity is continually maintained through 
inservice inspection and primary-to-secondary leakage monitoring. 
Any tubes with ODSCC degradation in excess of the URL are repaired.
    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 use of the bobbin probe to disposition ODSCC degraded tubes 
within TSP intersections by voltage-based repair criteria is 
demonstrated to maintain SG tube integrity in accordance with the 
requirements of RG 1.121. RG 1.121 describes a method acceptable to 
the NRC Staff for meeting GDCs 14, 15, 31, and 32 by reducing the 
probability or the consequences of SG tube rupture. This is 
accomplished by determining the limiting conditions of degradation 
of SG tubing, as established by inservice inspection, for which 
tubes with unacceptable degradation are removed from service. Upon 
implementation of the voltage-based repair criteria, even under the 
worst case conditions, the occurrence of ODSCC at TSP intersections 
is not expected to lead to a SG tube rupture during normal or 
faulted plant conditions, nor is it expected to lead to unacceptable 
primary-to-secondary leakage.
    In addressing the combined effects of a loss of coolant accident 
(LOCA) and safe shutdown earthquake (SSE) on the SGs, as required by 
GDC 2, it has been determined that tube collapse may occur based on 
analysis for a large break LOCA plus SSE. The analysis identifies a 
maximum of 7.5 percent of tubes per SG located adjacent to wedge 
regions that are subject to potential collapse during combined LOCA 
and SSE. Tubes located in the wedge region exclusion zone will be 
excluded from application of voltage-based repair criteria. Thus, 
existing tube integrity requirements apply to these tubes and the 
margin of safety is not reduced.
    Implementation practices using voltage-based repair criteria 
bounds RG 1.83 considerations. Specifically, GL 95-05 requires the 
following: (1) enhanced eddy current inspection guidelines are 
implemented to provide consistency in voltage normalization; (2) 100 
percent bobbin coil inspections are performed each cycle for all hot 
leg TSP intersections and all cold leg TSP intersections down to the 
lowest cold leg TSP with known ODSCC indications; and (3) rotating 
pancake coil (RPC) inspection of indications greater than 2 volts 
are performed to characterize the principal degradation as ODSCC. 
DCPP's proposed voltage-based repair criteria implementation 
practices meet the above requirements, and in some areas exceed them 
(for example, 100 percent bobbin coil inspections are routinely 
performed each cycle on every TSP intersection).
    Implementation of voltage-based repair criteria at TSP 
intersections will decrease the number of tubes which must be 
repaired. Since the installation of tube plugs to remove ODSCC 
degraded tubes from service reduces RCS flow margin, voltage-based 
repair criteria implementation will help preserve the margin of RCS 
flow.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Project Director: William H. Bateman.

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: February 27, 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 by revising Technical Specifications 
(TS) 3/4.8.1.1, ``A.C. Sources--Operating,'' to clarify that emergency 
diesel generator (EDG) testing is initiated from standby conditions 
rather than ``ambient'' conditions. The associated TS Bases will be 
revised to discuss the temperature range that satisfies EDG standby 
conditions. This amendment also proposes to revise TS 3/4.3.2, 
``Instrumentation--Engineering Safety Features Actuation System 
Instrumentation.'' This revision clarifies that when one or both of the 
first level load shed relays, or one or both of the second level 
undervoltage relays are inoperable, the associated EDG for that bus 
shall be declared 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:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to the technical specifications (TS) do not 
change the function or operation of any plant equipment or affect 
the response of that equipment if it is called upon to operate.
    The proposed change to TS 4.8.1.1.2a.2 and the Bases will 
clarify the term ``ambient conditions'' as used in the emergency 
diesel generator (EDG) surveillance requirement. EDG testing will 
still be completed on a frequency commensurate with the current TS.
    The proposed change to TS 3.3.2, Table 3.3-3, will permit time 
to restore the load shed first level undervoltage relays (FLURs) and 
second level undervoltage relays (SLURs) to operable status that is 
consistent with times allowed for outage of other safety-related 
equipment affecting one train of vital equipment. This proposed 
change maintains a high degree of equipment availability without 
requiring unnecessary initiation of a plant shutdown for partial 
equipment outages.
    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

[[Page 17241]]

accident from any accident previously evaluated.
    The proposed change to TS 4.8.1.1.2a.2 and the Bases will 
clarify the term ``ambient conditions'' as used in the EDG 
surveillance requirement. EDG testing will still be completed on a 
frequency commensurate with the current TS, and will be more 
representative of the conditions under which the EDGs would be 
required to start in an accident condition.
    The proposed change to TS 3.3.2, Table 3.3-3, will provide time 
to restore the load shed FLURs and SLURs to operable status that is 
consistent with times allowed for outage of other safety-related 
equipment affecting one train of vital equipment. The load shed FLUR 
and SLUR sets for one 4 kV bus only affect one train of vital 
equipment. If an accident occurred while the relays were inoperable, 
the redundant trains (two remaining EDGs and vital buses) would 
complete the safety function. The proposed allowed outage time (AOT) 
for the load shed FLURs and SLURs is bounded by the time allowed for 
an EDG supporting the vital 4 kV bus and is consistent with AOTs for 
other safety-related components.
    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 proposed change to TS 4.8.1.1.2a.2 and its Bases, clarifies 
the term ``ambient conditions'' as used in the EDG surveillance 
requirement. EDG testing will still be completed on a frequency 
commensurate with the current TS. Use of temperatures in the standby 
range result in no significant variation in EDG start times as 
indicated by the diesel vendor and by PG&E test results. Standby 
conditions are representative of actual starting conditions that 
would be in effect if the EDGs started in an accident.
    The proposed change to TS 3.3.2, Table 3.3-3, will provide time 
to restore the load shed FLURs and SLURs to operable status that is 
consistent with times allowed for outage of other safety-related 
equipment affecting one train of vital equipment. If an accident 
occurred while the relays were inoperable, the redundant trains (two 
remaining EDGs and vital buses) would complete the safety function. 
The proposed change eliminates an unneccessary plant shutdown and 
associated risk due to shutdown transient. It prevents a transient 
that could require the EDGs at a time when less than all three EDGs 
would be operable.
    Therefore, neither of the proposed changes involves a 
significant reduction in a margin of safety.

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

Portland General Electric Company, et al., Docket No. 50-344, 
Trojan Nuclear Plant, Columbia County, Oregon

    Date of amendment request: January 28, 1997.
    Description of amendment request: The proposed amendment by 
Portland General Electric (PGE or the licensee) clarifies the 
administrative controls that are used for the revision and maintenance 
of the Certified Fuel Handler Training Program. The change allows the 
licensee to make changes to the certified fuel handlers program without 
prior NRC staff approval. The text of the proposed change is taken from 
the improved standard technical specifications, NUREG-1431, ``Standard 
Technical Specifications, Westinghouse Plants.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis of the issue of no significant hazards 
consideration, which is presented below:

    In accordance with the requirements of 10 CFR 50.92, ``Issuance 
of amendment,'' this license amendment request is judged to involve 
no significant hazards consideration based upon the following:
    1. The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change is a clarification of the method of control 
that will be used for the Certified Fuel Handler Training Program, 
and as such, is administrative in nature and has no impact on the 
probability or consequences of accidents previously evaluated. The 
physical structures, systems, and components of the facility and the 
operating procedures for their use are unaffected by this proposed 
clarification. The proposed administrative controls provide adequate 
confidence that personnel that perform the certified fuel handler 
functions will have been adequately trained for the changing 
conditions of the facility. Since the training program will prepare 
the operations personnel for fuel handling operations, including 
responses to abnormal events/accidents, there will be no increase in 
the probability of occurrence or in the consequences of an accident 
previously evaluated.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    This change ensures the qualifications of the operations 
personnel are commensurate with the tasks to be performed and the 
conditions to which they may be required to respond. This change 
does not affect plant equipment or the procedures for operating 
plant equipment and, therefore, does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    This change ensures the qualification of the operations 
personnel are commensurate with the tasks to be performed and the 
conditions to which they may be required to respond. The assumptions 
for a fuel handling accident in the Fuel Building are not affected 
by the proposed change. The proposed amendment does not, therefore, 
involve a reduction in a margin of safety.

    The NRC staff has reviewed the analysis of the licensee 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: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for the Licensees: Leonard A. Girard, Esq., Portland 
General Electric Company, 121 S.W. Salmon Street, Portland, Oregon 
97204.
    NRR Project Director: Seymour H. Weiss.

Vermont Yankee Nuclear Power Corporation, Docket No. 50-271, 
Vermont Yankee Nuclear Power Station, Vernon, Vermont

    Date of amendment request: September 11, 1996.
    Description of amendment request: The proposed amendment would 
permit operation with increased safety relief valve (SRV) and safety 
valve (SV) setpoint tolerance and permit operation up to 100% of rated 
power with a single inoperable SRV.
    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 changes will permit operation with increased SRV 
and SV setpoint tolerance and permit operation up to 100% of rated 
power with a single inoperable SRV.

[[Page 17242]]

    The valves are not related to the control rod system. The valves 
are not involved in the initiation of a Control Rod Drop Accident. 
The valves are part of the Reactor Vessel (RV) pressure boundary and 
their failure could initiate a LOCA [loss-of-coolant accident]. 
However, the proposed changes do not constitute a change in the 
design of the valves from a pressure boundary perspective. The 
proposed changes do not affect the probability of a LOCA initiated 
by valve failure. The valves are not a component, system, or 
structure involved in refueling operations. The valves and their as-
found setpoint tolerance are not involved in the initiation of a 
Refueling Accident.
    The design basis Main Steam Line Break is a complete severance 
of one main steam line outside the secondary containment. The SRVs 
and SVs are located inside primary containment and cannot cause a 
main steam line rupture outside secondary containment. The valves 
are not involved in the initiation of a design basis Main Steam Line 
Break. The probability or consequences of these accidents are not 
affected.
    Attachment C [see application dated September 11, 1996] includes 
an analysis to demonstrate that margin exists to SV challenges 
during an Abnormal Operational Transient (AOT). For this purpose a 
Generator Load Rejection without Bypass (GLRWOBP) was identified as 
the limiting AOT. The results confirm that SV challenges would not 
occur with an inoperable SRV at rated power.
    The current Technical Specification limit of 95% rated power or 
less with an inoperable SRV is therefore not required to prevent SV 
challenges during an AOT.
    As discussed in Attachment C [see application], the impact of 
the proposed as-found SRV setpoint tolerance increase on SRV piping/
supports and discharge loads to the Torus was evaluated. A 
mechanical loads analysis confirmed the integrity of these 
components, systems, and structures during SRV discharge with the 
proposed changes.
    Attachment C [see application] provides an evaluation of the 
impact of the proposed changes on the consequences of the Loss of 
Coolant Accident and the Main Steam Line Break. The limiting LOCA 
event is a break in the recirculation loop, with a break area of 0.6 
ft\2\, at the pump discharge location, with a loss of one train of 
DC power as the single failure. For breaks in the recirculation line 
larger than 0.4 ft\2\, the SRVs would not be challenged. Therefore, 
in assessing the impact of the proposed changes on 10CFR50.46 
acceptance criteria, only recirculation line breaks less than 0.4 
ft\2\ were reevaluated. Results show that the 0.6 ft\2\ 
recirculation line break remains the limiting LOCA event and it is 
not affected. The consequences of the limiting design basis LOCA are 
not increased by the proposed changes. The design basis accident for 
containment performance is a double-ended break in the recirculation 
pump suction. For this size break, the SRVs are not challenged. 
Therefore, the proposed changes do not have any effect on the design 
basis accident for containment performance. The design basis 
accident for radioactive material releases and radiological effects 
is a complete severance of one main steam line outside the secondary 
containment. For steam line breaks outside the containment, MSIVs 
[main steam isolation valves] close and terminate radiological 
releases outside the containment, SRVs are not challenged until 
after MSIV closure and isolation. Therefore, the proposed changes do 
not increase the radiological consequences of the design basis Main 
Steam Line Break.
    The SRVs and SVs are designed to mitigate the consequences of 
malfunctions of equipment which result in a Nuclear System pressure 
increase. These abnormal operational transients are defined and 
analyzed in Section 14.5.1 of the VY [Vermont Yankee] FSAR [final 
safety analysis report]. The impact of the proposed changes on these 
abnormal operational transients was evaluated. Results are 
documented in Attachment C [see application] and show that 
applicable acceptance criteria are met provided operating MCPR 
[minimum critical power ratio] limits as specified in the COLR [core 
operating limit report] are adjusted to reflect the effects of the 
proposed changes. A hot channel analysis of the limiting delta CPR 
overpressure transient confirmed that a 0.02 increase in the 
operating MCPR limits bounds the combined effects of implementing 
the proposed changes in the current cycle. The operating MCPR limits 
in COLR have already been increased for the current cycle. 
Appropriate operating MCPR limits for future cycles will be 
determined from cycle-specific safety analyses performed with the 
approved changes.
    Current practice regarding SRV setpoints is to assure plus or 
minus 1% tolerance is met as required by the ASME [American Society 
of Mechanical Engineers] Boiler & Pressure Vessel Code referenced in 
Technical Specification Surveillance Requirement 4.6.E.2. As-left 
setpoints always meet the plus or minus 1% tolerance. The safety 
analysis in Attachment C [see application] demonstrates that as-
found setpoints within plus or minus 3% are acceptable. However, 
valves re-installed after testing will continue, as previously, to 
meet plus or minus 1% tolerance as required by the ASME Boiler & 
Pressure Vessel Code. Thus, the probability of SRV actuation (and 
the associated risk of failure to reseat properly) is not increased 
by the proposed change.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from an accident previously 
evaluated.
    The proposed changes will permit operation with increased Safety 
Relief Valve (SRV) and Safety Valve (SV) setpoint tolerance and 
permit operation up to 100% of rated power with a single inoperable 
SRV. The proposed changes:
    (1) do not constitute a change in the design of the valves;
    (2) will not cause the valve or associated systems and 
structures to be operated beyond their original design envelopes; 
and,
    (3) do not involve new plant equipment.
    Therefore, this amendment does not create the possibility of a 
new or different kind of accident.
    3. The proposed amendment will not involve a significant 
reduction in a margin of safety.
    Technical Specification Basis 3.6 and 4.6D identifies the 
minimum critical power ratio (MCPR) safety limit. Operational 
restraints on MCPR are placed in the COLR to assure no violation of 
the MCPR safety limit during AOTs. The impact of the proposed 
changes on MCPR limits was determined by performing a hot channel 
analysis for the overpressure transient which yields the largest 
transient drop in CPR [critical power ratio] (delta CPR). Results 
are documented in Attachment C [see application], and show that a 
0.02 increase in the operating MCPR limits bounds the combined 
effects of the proposed changes and assures the MCPR safety limit is 
not violated during AOTs. The margin of safety defined by the MCPR 
safety limit is not reduced.
    Technical Specification Basis 3.6 and 4.6D also identifies the 
ASME Boiler and Pressure Vessel Code Section III-A limit which 
permits pressure transients up to 10% over design pressure (110% x 
1250 = 1375 psig). This margin of safety is not reduced by the 
proposed changes. Attachment C [see application] documents new 
overpressure transient analysis with results that demonstrate the 
ASME overpressure limit of 110% of design is met. This license 
amendment request does not propose to reduce the margin of safety 
defined by the ASME Boiler & Pressure Vessel Code limit.

    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: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.
    Attorney for licensee: R. K. Gad, III, Ropes and Gray, One 
International Place, Boston, MA 02110-2624.
    NRC Project Director: Patrick D. Milano, Acting.

Virginia 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: February 3, 1997 as supplemented March 
18, 1997.
    Description of amendment request: The proposed change to Technical 
Specification 4.15.B.1 is administrative in nature in that it revises 
the Technical Specifications (TS) to be consistent with the NRC-
approved inservice inspection program. In addition, three TS pages 
which were previously approved by NRC, and which were inadvertently 
omitted in an earlier amendment (amendments 40 and 39 for units 1 and 
2, respectively), are being reissued.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 17243]]

issue of no significant hazards consideration, which is presented 
below:

    1. Operation of Surry Units 1 and 2 in accordance with the 
proposed Technical Specifications change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed change is administrative in nature, and station 
operations are not being affected. The ASME Section XI Code 
requirements are thoroughly established, reviewed and approved by 
ASME, the industry and ultimately endorsed by the NRC for inclusion 
into 10 CFR 50.55a. Updates to the Code reflect advances in 
technology and consider information obtained from plant operating 
experience to provide enhanced inspection and examination techniques 
for pipe welds. Therefore, performing weld examinations for the pipe 
in our augmented inspection program to the requirements of the 1989 
edition of the ASME Section XI Code provides a regulatory acceptable 
and adequate level of assurance that the integrity of the pipe will 
be maintained. By not referencing a specific Code edition in the 
Technical Specifications, our examinations for pipe in the augmented 
inspection program will consistently be performed to the Code of 
record, consistent with the requirements [of] 10 CFR 50.55a. 
Consequently, the probability or consequences of an accident 
previously evaluated are not increased.
    2. The proposed Technical Specifications change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    As noted above, the proposed change is administrative in nature, 
and no new accident precursors are being introduced. Since the 
augmented inspection program will continue to be performed to NRC 
approved ASME Section XI Code requirements, adequate assurance is 
provided to ensure the integrity of the pipe. Consequently, the 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed Technical Specifications change does not involve 
a significant reduction in a margin of safety.
    Performing weld examinations to the Code of record is prudent, 
consistent with accepted industry and regulatory requirements, and 
provides adequate assurance that piping integrity will be 
maintained. The use of a general ASME Section XI Code reference in 
Technical Specification 4.15.B.1 is consistent with the existing 
wording in Technical Specifications 4.15.A and C, and ensures that 
weld examinations are being consistently performed to the currently 
approved edition of the ASME Section XI Code. This is an 
administrative change and as such does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. The staff notes that the reissuance of three TS pages is a 
purely administrative matter which involves no significant hazards 
consideration and which has been considered previously. Therefore, the 
NRC staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Local Public Document Room location: Swem 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: Mark Reinhart, Acting.

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

    Dates of amendment requests: June 4, 1996, as supplemented August 
5, September 26, October 21, November 13, November 20, and December 2, 
1996, and January 16, March 5, and March 20, 1997 (TSCR 188 and 189).
    Description of amendment requests: The proposed amendments would 
revise License Nos. DPR-24 and DPR-27 to add commitments for control 
room habitability and revise Technical Specification (TS) Sections 
15.1, ``Definitions,'' 15.2.1, ``Safety Limit, Reactor Core,'' 15.2.3, 
``Limiting Safety System Settings and Protective Instrumentation,'' 
Section 15.3.1, ``Reactor Coolant System,'' 15.3.4, ``Steam and Power 
Conversion System,'' 15.3.5, ``Instrumentation System, 15.4.1, 
``Operational Safety Review,'' 15.5.3, ``Design Features--Reactor,'' 
and 15.6.9, ``Plant Reporting Requirements,'' and modify the bases for 
Section 15.2.2, ``Safety Limit, Reactor Coolant System Pressure,'' and 
Section 15.3.1.C, ``Maximum Coolant Activity,'' to incorporate changes 
associated with the operation of Point Beach Nuclear Plant (PBNP), Unit 
2, with replacement steam generators. The new analyses performed for 
replacing Unit 2 steam generators resulted in changes to the reactor 
core safety limits and protective instrumentation setpoints for Unit 1 
as well as Unit 2. Calculations are based on operation at either 2000 
psia or 2250 psia and an average temperature limit of greater than or 
equal to 557 degrees Fahrenheit and less than or equal to 573.9 degrees 
Fahrenheit. New dose calculations were performed based on new setpoints 
for low-low steam generator water level, new values of primary and 
secondary steam generator volumes, and revised accident analyses for 
steam generator tube rupture, main steam line break, locked rotor, and 
control rod ejection. Additional license conditions are proposed to 
document the commitments made to improve habitability of the control 
room so that dose limits do not exceed 10 CFR Part 50, Appendix A, 
General Design Criterion 19, without relying on the use of potassium 
iodide pills and/or self-contained breathing apparatus. The original 
applications were previously noticed in the Federal Register on July 3, 
1996 (61 FR 34903 and 34904).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed TS changes reflect the replacement of steam 
generators at PBNP, including new analyses and setpoints, and a 
different standard and acceptance criteria for Dose Equivalent I-
131. The proposed setpoints maintain the margin to safe operation of 
Unit 2 with the replacement steam generators. In order to maintain 
one set of safety analyses for both units, the analyses for 
operation of Unit 2 with the replacement steam generators were 
performed to encompass the operation of Unit 1. Therefore, the 
proposed changes apply to the operation of both units and maintain 
the margin of safety for each. The staff independently performed an 
evaluation of the dose consequences for steam generator tube 
rupture, main steam line break, locked rotor accident, and a rod 
ejection accident. The staff determined there are no significant 
increases in dose for the low population zone or the exclusion area 
boundary. The licensee had not previously analyzed these accidents 
for control room habitability. As a result of the proposed changes, 
limiting control room doses will require compensatory measures, use 
of potassium iodide and self-contained breathing apparatus, which 
have been previously approved, until such time that the control room 
ventilation design is improved. The commitments to improve control 
design/operation are included as license conditions. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Installation of new steam generators, with a small increase in 
primary side volume and new setpoints for instrumentation, does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The proposed setpoints

[[Page 17244]]

maintain the margin to safe operation of Unit 2 with the replacement 
steam generators. In order to maintain one set of safety analyses 
for both units, the analyses for operation of Unit 2 with the 
replacement steam generators were performed to encompass the 
operation of Unit 1. Therefore, the proposed changes apply to the 
operation of both units and maintain the margin of safety for each. 
These changes do not affect any of the parameters or conditions that 
contribute to initiation of any accidents. Therefore, the proposed 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed setpoints maintain the margin to safe operation of 
Unit 2 with the replacement steam generators. In order to maintain 
one set of safety analyses for both units, the analyses for 
operation of Unit 2 with replacement steam generators were performed 
to encompass the operation of Unit 1. Therefore, the proposed 
changes apply to the operation of both units and maintain the margin 
of safety for each. Compensatory measures will ensure control room 
doses remain within the dose guidelines in 10 CFR Part 50, Appendix 
A, General Design Criterion 19, until such time as the control 
ventilation system design/operation is revised. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.
    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: September 30, 1996, as supplemented 
November 26, and December 12, 1996, February 13, and March 5, 1997 
(TSCR 192).
    Description of amendment request: The proposed amendments would 
revise License Nos. DPR-24 and DPR-27 to add commitments for control 
room habitability and revise Technical Specification (TS) Sections 
15.3.3, ``Emergency Core Cooling System, Auxiliary Cooling Systems, Air 
Recirculation Fan Coolers, and Containment Spray,'' TS 15.3.7, 
``Auxiliary Electrical Systems,'' 15.5.2, ``Design Features-
Containment,'' and associated TS Bases to reflect proposed changes to 
the limiting conditions for operation, action statements, allowable 
outage times, and design specifications for the Point Beach Nuclear 
Plant (PBNP) TS associated with the containment accident fan coolers, 
service water equipment (pumps and piping), component cooling water 
pumps, and normal and emergency power supplies. Specifically, these 
proposed changes increase the number of service water pumps and 
component cooling water pumps required to be operable, change the 
description of the service water system to define three separate loops, 
modify the limiting conditions for operation of the containment cooling 
and iodine removal systems and the component cooling water and service 
water systems, modify the auxiliary electrical system requirements, 
modify the associated TS Bases, and change the design value for each 
containment ventilation/air coolers from 55,600 Btu/sec to 41,700 Btu/
sec. The original application was previously noticed in the Federal 
Register on November 19, 1996 (61 FR 58905).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.

    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes involve components currently installed in 
the facilities and reflect current capabilities of this equipment. 
Increasing the number of service water and component cooling water 
pumps required to be operable, changing the service water header 
definitions and modifying the limiting conditions for operation for 
service water and component cooling water, and modifying the 
requirements for the 4160/480-volt safeguards buses does not 
increase the probabilities of any accidents currently evaluated in 
the final safety analysis report (FSAR). The probabilities of 
accidents previously evaluated in the FSAR are based on the 
probability of initiating events for these accidents. Initiating 
events for accidents previously evaluated for Point Beach include: 
Control rod withdrawal and drop, CVCS [chemical volume and control 
system] malfunction (boron dilution), startup of an inactive reactor 
coolant loop, reduction in feedwater enthalpy, excessive load 
increase, losses of reactor coolant flow, loss of external 
electrical load, loss of normal feedwater, loss of all AC power to 
the auxiliaries, turbine overspeed, fuel handling accidents, 
accidental releases of waste liquid or gas, steam generator tube 
rupture, steam pipe rupture, control rod ejection, and primary 
coolant system ruptures. The change to the heat removal capability 
of the containment ventilation/air coolers from 55,600 Btu/sec to 
41,700 Btu/sec was evaluated to ensure that containment design is 
not challenged. Therefore, the proposed changes do not affect the 
probability of occurrence or the consequences of any accident 
previously evaluated in the FSAR. During review of the proposed 
changes, the staff determined that other changes made to the 
operation of the containment spray system and the control room 
ventilation design and operation could affect the doses associated 
with a loss-of-coolant accident. The staff has determined that there 
is no significant increase in offsite doses. As a result of the 
proposed changes and current plant design, limiting control room 
doses will require compensatory measures, use of potassium iodide 
and self-contained breathing apparatus, which have been previously 
approved, until such time that the control room ventilation design/
operation is improved. The commitments to improve control design/
operation are included as license conditions.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not introduce any new accidents from any 
previously evaluated. Failures for the systems affected by the 
proposed changes, service water system, component cooling water 
system, containment ventilation/air cooling units, and the 4160/480-
volt safeguards buses are factored into the accident analyses 
included in the FSAR. No new or different kinds of accidents are 
created since no new or different accident initiators or sequences 
are involved. Therefore, these proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated in the Point Beach FSAR.
    (3) The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes provide the appropriate limiting conditions 
for operation, action statements, allowable outage times, and design 
specifications for service water, component cooling water, 
containment cooling, and normal and emergency power supplies. This 
ensures that the safety systems that protect the reactor and 
containment will operate as required. The impact of changes to 
design and operation of affected systems do not affect the reactor 
and containment design. Therefore, the margins of safety for Point 
Beach are not being reduced because the design and operation of the 
reactor and containment are not being changed and the safety systems 
and limiting conditions of operation for these safety systems that 
provide their protection that are being changed will continue to 
meet the requirements for accident mitigation for PBNP. Compensatory 
measures will ensure control room doses remain within the dose 
guidelines in 10 CFR Part 50, Appendix A, General Design Criterion 
19, until such time as the control ventilation system design/

[[Page 17245]]

 operation is revised. Therefore, the proposed changes will not 
create a significant reduction in a margin of safety.

    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: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

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

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

    Date of amendment request: March 7, 1997.
    Description of amendment request: The proposed amendments would 
revise Technical Specification 3/4.7.1.6 and Section 15.6.3 of the 
Updated Final Safety Analysis Report to require four instead of three 
steam generator pressure operated relief valves operable.
    Date of publication of individual notice in Federal Register: March 
13, 1997 (62 FR 11931).
    Expiration date of individual notice: April 14, 1997.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina.

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

    Date of amendment request: March 17, 1997.
    Brief description of amendment request: The proposed amendment 
would modify the Design Features Section 5.3.1 of the Technical 
Specifications to reflect the Atrium-10 design and would include a 
Siemens Power Corporation topical report reference in Section 6.9.3.2 
to reflect mechanical design criteria for this fuel. This change would 
allow this fuel to be loaded and maintained in the core only under 
Condition 5, (refueling).
    Date of publication of individual notice in Federal Register: March 
25, 1996 (62 FR 14167).
    Expiration date of individual notice: April 24, 1997.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Notice of Issuance of Amendments to Facility Operating Licenses

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

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2, 
and 3, Maricopa County, Arizona

    Date of application for amendments: May 2, 1995, as supplemented by 
letter dated March 7, 1996.
    Brief description of amendments: These amendments modify the 
licenses to authorize incorporation in the Updated Final Safety 
Analysis Report (UFSAR) of certain changes to the description of the 
facilities involving a revised large-break loss of coolant accident 
(LOCA) analysis that addresses a previously unanalyzed release path 
through the steam generators to the atmosphere.
    Date of issuance: March 17, 1997.
    Effective date: March 17, 1997, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 1--111; Unit 2--103; Unit 3--83.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Operating Licenses and Updated Final Safety 
Analysis Report.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62487). The March 7, 1996, supplemental 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 
March 17, 1997. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.

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

    Date of application for amendment: January 29, 1997, as 
supplemented February 6, and February 21, 1997.
    Brief description of amendment: The amendment adds a new Technical

[[Page 17246]]

Specification 3.0.5 to provide guidance for returning equipment to 
service under administrative controls for the sole purpose of 
performing testing to demonstrate operability.
    Date of issuance: March 17, 1997.
    Effective date: March 17, 1997.
    Amendment No.: 69.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 
(62FR6569).
    The February 6, and February 21, 1997 letters 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 March 17, 1997.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendment: January 10, 1997, as 
supplemented January 31, February 20, and March 3, 1997.
    Brief description of amendment: The amendment revises Technical 
Specification 4.8.1.1.2 to clarify pressure testing requirements for 
the isolable and non-isolable portions of the diesel fuel oil piping.
    Date of issuance: March 19, 1997.
    Effective date: March 19, 1997.
    Amendment No.: 70.
    Facility Operating License No. NPF-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 5, 1997 (62 FR 
5490). The January 31, February 20, and March 3, 1997, letters 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 March 19, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: November 4, 1996 and 
supplemented February 5, 1997.
    Brief description of amendments: The amendments revise Technical 
Specification Section 4.7.13.1.c to eliminate the requirement that the 
18-month Standby Shutdown System diesel generator inspection be 
performed only during shutdown of both reactors.
    Date of issuance: March 13, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--157--Unit 2--149.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64383) The supplemental letter dated February 5, 1997, provided 
additional information that did not change the scope of the November 4, 
1996, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 13, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730.

Duke Power Company, Docket Nos. 50-369 and 50-370, McGuire Nuclear 
Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: January 13, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications so that the containment integrated leak rate 
Type A testing will now be performed consistent with the revised 10 CFR 
Part 50, Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.'' No changes to 
implement Option B for the Type B and Type C tests were requested by 
the licensee at this time.
    Date of issuance: March 21, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--173--Unit 2--155.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6575) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: J. Murrey Atkins Library, 
University of North Carolina at Charlotte, 9201 University City 
Boulevard, North Carolina 28223-0001.

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

    Date of application of amendments: February 15, 1996, as 
supplemented by letter dated February 18, 1997.
    Brief description of amendments: The amendments add operability and 
surveillance requirements regarding operation and testing of the Keowee 
Hydro Station to the Oconee Technical Specifications.
    Date of Issuance: March 20, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days. Implementation shall include revision of the Selected Licensee 
Commitment manual to incorporate the Keowee Hydro units' commercial 
power operating restrictions curves in accordance with the application 
for the amendments.
    Amendment Nos.: Unit 1--222; Unit 2--222; Unit 3--219.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13523) The February 18, 1997, letter provided clarifying information 
that did not change the scope of the February 15, 1996, application and 
the initial proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 20, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.

[[Page 17247]]

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

    Date of application for amendments: September 18, 1992, as 
supplemented October 6, 8, 15, 23, and November 13 and 20, 1992, March 
5, May 24, June 10, and December 20, 1993, April 6 and July 28, 1995, 
and September 11, October 1, December 13, 19 and 23, 1996.
    Brief description of amendments: The amendments modify the Facility 
Operating Licenses, Technical Specifications, Environmental Protection 
Plan, and Antitrust conditions to add Southern Nuclear Operating 
Company, Inc., as operator of the facilities, with exclusive 
responsibility and control over its physical construction, operation, 
and maintenance. The antitrust license conditions divorce Southern 
Nuclear from marketing or brokering power or energy from the Hatch 
Plant and holds Georgia Power Company accountable for the actions of 
its agent, Southern Nuclear, to the extent Southern Nuclear's actions 
contravene the Hatch antitrust license conditions. An Order Approving 
Southern Nuclear Operating Company, Incorporated, As Exclusive Operator 
was included along with the issuance of the amendments.
    Date of issuance: March 17, 1997.
    Effective date: To be implemented within 60 days of the date of 
issuance.
    Amendment Nos.: 203 and 144.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications and Operating Licenses.
    Date of initial notice in Federal Register: October 14, 1992 (57 FR 
47131). The October 6, 8, 15, 23, and November 13 and 20, 1992, March 
5, May 24, June 10, and December 20, 1993, April 6 and July 28, 1995, 
and September 11, October 1, December 13, 19 and 23, 1996, letters, did 
not change the scope of the September 18, 1992, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 17, 1997, and an Environmental Assessment 
dated October 27, 1992.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

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

    Date of application for amendments: October 7, 1996.
    Brief description of amendments: The amendments revise Surveillance 
Requirements (SRs) 3.1.7.7 and 3.4.3.1, and Limiting Conditions for 
Operation 3.4.3, 3.5.1, and 3.6.1.6 to increase the nominal mechanical 
pressure relief setpoints for all of the 11 safety/relief valves (SRVs) 
to 1150 psig and allow operation with one SRV and its associated 
functions inoperable. The change will reduce the potential for SRV 
pilot leakage and the potential for forced outages due to an inoperable 
SRV during a fuel cycle.
    Date of issuance: March 21, 1997.
    Effective date: As of the date of issuance to be implemented for 
Unit 1 prior to startup from its refueling outage scheduled for fall 
1997; and for Unit 2 prior to startup from its refueling outage 
currently scheduled for March 15, 1997.
    Amendment Nos.: 204 and 145.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
129). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

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

    Date of application for amendments: October 29, 1996, as 
supplemented February 19, 1997.
    Brief description of amendments: The amendments revise the 
Technical Specifications associated with the installation of a digital 
Power Range Neutron Monitoring system.
    Date of issuance: March 21, 1997.
    Effective date: As of the date of issuance to be implemented for 
Unit 1 prior to its startup from the fall of 1997 refueling outage; and 
implemented for Unit 2 prior to its startup from the spring of 1997 
refueling outage.
    Amendment Nos.: 205 and 146.
    Facility Operating License Nos. DPR-57 and NPF-5: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
130). The February 19, 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 March 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513.

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear

    Date of application for amendment: November 27, 1996 (TSCR 232).
    Brief description of amendment: The amendment changes the 
acceptance criteria for the individual cell voltage from 2.0v to 2.09v, 
the frequency for battery specific gravities to implement the 
recommendations of IEEE 450-1995, deletes surveillance 4.7.B.4.d, and 
adds a clarifing phrase ``while on a float charge . . .'' where 
appropriate.
    Date of Issuance: March 24, 1997.
    Effective date: March 24, 1997.
    Amendment No.: 189.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 12, 1997 (62 
FR 6576) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated March 24, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of amendment request: November 20, 1996, as supplemented by 
letters dated February 20, 1997, and March 25, 1997.
    Brief description of amendment: The amendment revises Section 5.2 
of the Fort Calhoun Station technical specifications to relocate 
controls for working hours to the Updated Safety Analysis Report.
    Date of issuance: March 27, 1997.
    Effective date: March 27, 1997.
    Amendment No.: 181.

[[Page 17248]]

    Facility Operating License No. DPR-40: Amendment revised the 
Technical Specifications and operating license.
    Date of initial notice in Federal Register: January 2, 1997 (62 FR 
131) The February 20, 1997, and March 25, 1997, supplemental letters 
provided additional clarifying information that did not change the 
portion of the initial no significant hazards consideration 
determination that addressed this proposed change.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 27, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102.

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

    Date of application for amendment: January 11, 1996, as 
supplemented by letters dated February 26, May 22, June 27, July 12, 
December 23, 1996, and March 17, 1997
    Brief description of amendment: The amendments revise Section 6.0 
(Administrative Controls) of the Hope Creek TS to: (1) Relocate the 
requirements of Section 6.5 (Station Operations Review Committee, 
Nuclear Safety Review and Audit, and Technical Review and Control) to 
the Quality Assurance Program, (2) replace specific management titles 
with generic management functional positions, (3) change Operating 
Engineer to Assistant Operations Manager, (4) require a Senior Reactor 
Operator license be held by either the Operations Manager or one of the 
Assistant Operations Managers, and (5) correct some typographical 
errors in Section 6.0.
    Date of issuance: March 21, 1997.
    Effective date: As of date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 97.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications and the license.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5817).
    The supplemental letters provided clarifying information that did 
not change the initial proposed no significant hazards consideration 
determination nor the original notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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

    Date of application for amendment: October 25, 1996, as 
supplemented by letters dated December 4, 1996, and January 24, 1997.
    Brief description of amendment: This amendment changes Hope Creek 
Technical Specification (TS) 3/4.1.3.5, ``Control Rod Scram 
Accumulator,'' in order to: 1) Permit a separate entry into a TS action 
statement for each inoperable control rod; 2) provide more specific 
applicability for required actions in Operational Condition 1 or 2 with 
one inoperable control rod scram accumulator (reactor pressure of 
 900 psig would be specified); 3) provide more specific 
actions (verify charging water pressure) for two or more inoperable 
control rod scram accumulators when reactor pressure is  900 
psig; 4) provide more specific actions when reactor pressure is < 900 
psig and one or more control rod scram accumulators are inoperable 
(verify insertion of control rods associated with inoperable 
accumulators and verify that charging water header pressure is 
 940 psig); 5) provide specific actions in Operational 
Condition 5 with one or more withdrawn control rods inoperable; and 6) 
eliminate the requirements to perform an 18-month channel functional 
test of the leak detectors and the 18-month channel calibration of the 
pressure detectors.
    Date of issuance: March 26, 1997.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment No.: 98.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1996 (61 FR 
64394) The December 4, 1996, and January 24, 1997, supplements 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 March 26, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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

    Date of application for amendments: January 11, 1996, as 
supplemented by letters dated February 26, May 22, June 27, July 12, 
December 23, 1996, and March 17, 1997.
    Brief description of amendments: The amendments revise Section 6.0 
(Administrative Controls) of the Salem TS to: 1) relocate the 
requirements of Section 6.5 (Station Operations Review Committee, 
Nuclear Safety Review and Audit, and Technical Review and Control) to 
the Quality Assurance Program, 2) replace specific management titles 
with generic management functional positions, 3) change Operating 
Engineer to Assistant Operations Manager, 4) require a Senior Reactor 
Operator license be held by either the Operations Manager or one of the 
Assistant Operations Managers, and 5) correct some typographical errors 
in Section 6.0.
    Date of issuance: March 21, 1997.
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.
    Amendment Nos.: 192 and 175.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the Technical Specifications and the license.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5818) The supplemental letters provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination nor the original notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 21, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of amendment request: December 26, 1997, as supplemented by 
letter dated February 6, March 7, and March 21, 1997.
    Brief Description of amendment: The amendment changes Technical 
Specification 3/4.4.6, ``Steam Generators'' and associated Bases to 
implement the voltage-based alternate repair criteria for steam 
generator tubes in Farley Unit 1 in accordance with

[[Page 17249]]

Generic Letter 95-05, ``Voltage-Based Repair Criteria for Westinghouse 
Steam Generator Tubes Affected by Outside Diameter Stress Corrosion 
Cracking.''
    Date of issuance: March 24, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 124.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4353) By letter dated February 6, 1997, the licensee submitted 
additional information to clarify the changes to the proposed repair 
criteria, which did not change the scope of the December 26, 1996, 
application and the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated March 24, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

    Date of amendments request: January 10, 1997, as supplemented by 
letter dated February 24, 1997.
    Brief Description of amendments: The amendments revise the 
Technical Specifications (TS) to incorporate the latest revised topical 
reports governing the installation of laser welded steam generator tube 
sleeves. In addition, the reference to a one-cycle implementation of 
L*, which expired at the last Unit 2 outage was deleted from the Unit 2 
TS.
    Date of issuance: March 24, 1997.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 125 and 119.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: January 29, 1997 (62 FR 
4355) The February 24, 1997, letter provided clarifying information 
that did not change the original application and the initial proposed 
no significant hazards consideration determination published in the 
Federal Register on January 29, 1997 (62 FR 4355).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 24, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

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

    Date of amendments request: September 30, 1996.
    Brief description of amendments: The amendments revise Technical 
Specifications (TS) 3/4.1.1.1, 3/4.1.1.2, 3/4.1.1.3, 3/4.1.3.5, 
3.1.3.6, 3.2.1, 3.2.2 and 3.2.3 and associated Bases to remove certain 
cycle-specific parameter limits from the TS and relocate them to the 
Core Operating Limits Report.
    Date of issuance: March 25, 1997.
    Effective date: As of the date of issuance to be implemented for 
Unit 1 prior to entry into Mode 5 following the next scheduled 
refueling outage, which should begin in March 1997; for Unit 2 prior to 
entry into Mode 5 following the refueling outage scheduled to begin in 
March 1998.
    Amendment Nos.: 126 and 120.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications and License Conditions.
    Date of initial notice in Federal Register: November 6, 1996 (61 FR 
57491) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 25, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Southern California Edison Company, et al., Docket No. 50-362, San 
Onofre Nuclear Generating Station, Unit No. 3, San Diego County, 
California

    Date of application for amendment: February 18, 1997, as 
supplemented by letter dated February 21, 1997.
    Brief description of amendment: The amendment defers implementation 
of Surveillance Requirement 3.3.5.6 of Technical Specifcation 3.3.5, 
``Engineered Safety Features Actuation System (ESFAS) 
Instrumentation,'' until the next SONGS Unit 3 shutdown, which will be 
no later than the upcoming Cycle 9 refueling outage (currently 
scheduled for April 12, 1997).
    Date of issuance: March 17, 1997.
    Effective date: March 17, 1997.
    Amendment No.: 127
    Facility Operating License No. NPF-15: The amendments revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: Yes (62 FR 9001 dated February 27, 1997). The notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by March 31, 1997, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of the amendment. The February 
21, 1997, letter provided additional clarifying information and did not 
change the original no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 17, 1997.
    Attorney for licensee: Douglas K. Porter, Esquire, Southern 
California Edison Company, P. O. Box 800, Rosemead, California 91770.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse Nuclear 
Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: February 14, 1997.
    Brief description of amendment: This amendment revises Technical 
Specification (TS) Section 3/4.5.2, ``Emergency Core Cooling Systems, 
ECCS Subsystems--Tavg  280 deg.F.'' Surveillance 
requirement 4.5.2.f would be modified to state that opening and closing 
of the inspection port on the watertight enclosure for the decay heat 
valve pit would not require this surveillance procedure to be 
performed. This amendment also revises the applicable TS bases.
    Date of issuance: March 24, 1997.
    Effective date: Immediately, and shall be implemented no later than 
120 days after issuance.
    Amendment No.: 215.
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (62 FR 8783 dated February 26, 1997). The 
notice

[[Page 17250]]

provided an opportunity to submit comments on the Commission's proposed 
NSHC determination. No comments have been received. The notice also 
provided for an opportunity to request a hearing by March 30, 1997, but 
indicated that if the Commission makes a final NSHC determination, any 
such hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of NSHC are contained in 
a Safety Evaluation dated March 24, 1997.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606.

United States Department of Commerce, National Institute of Standards 
and Technology, Docket No. 50-184, NIST Test Reactor

    Date of application for amendment: January 17, 1997.
    Brief description of amendment: This amendment revises the 
Technical Specifications to change the name of the Reactor Radiation 
Division to the NIST Center for Neutron Research and the Chief, 
Radiation Division to Director, NIST Center for Neutron Research.

    Date of issuance: March 31, 1997.
    Effective date: March 31, 1997.
    Amendment No.: 8.
    Amended Facility License No. TR-5: This amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1997 (62 
FR 8801). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 31, 1997.
    No significant hazards consideration comments received: No.
    Local Public Document Room Location: N/A.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By May 9, 1997, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the

[[Page 17251]]

designated Atomic Safety and Licensing Board will issue a notice of a 
hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington, DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 
50-529, and STN 50-530, Palo Verde Nuclear Generating Station, 
Units 1, 2, and 3, Maricopa County, Arizona

    Date of application for amendments: December 27, 1996, as 
supplemented by letter dated March 18, 1997.
    Brief description of amendments: The amendments modify the licenses 
to authorize incorporation of certain changes to the description of the 
facilities involving offsite power sources into the Updated Final 
Safety Analysis Report (UFSAR) for the Palo Verde Nuclear Generating 
Station (PVNGS).
    Date of issuance: March 26, 1997.
    Effective date: March 26, 1997, to be implemented within 60 days of 
the date of issuance.
    Amendment Nos.: Unit 1--112; Unit 2--104; Unit 3--84.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the operating licenses and the Updated Final Safety 
Analysis Report.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated March 
26, 1997.
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004.
    NRC Project Director: William H. Bateman.

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

    Date of application for amendments: March 26, 1997, as supplemented 
on March 27, 1997.
    Brief description of amendments: The proposed amendments provided 
(1) An evaluation of the Unreviewed Safety Question (USQ) involving the 
control room operator dose resulting from error in the secondary 
containment volume, (2) a change in Technical Specification (TS) 
4.7.P.2.b and 4.7.P.3 values for the allowed methyl iodide penetration 
for the standby gas treatment charcoal adsorbers, and (3) change of TS 
5.2.C to reflect the new calculated free volume of the secondary 
containment.
    Date of Issuance: March 27, 1997.
    Effective date: March 27, 1997.
    Amendment Nos.: 175, 171.
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated March 
27, 1997.

[[Page 17252]]

    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.
    NRC Project Director: Robert A. Capra.

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

    Date of application for amendments: January 29, 1997, as 
supplemented February 11, 12, March 7, 10, 11, 19, and 20, 1997.
    Brief description of amendments: The amendments authorize Northern 
States Power Company to continue operation of Prairie Island Units 1 
and 2 on an interim basis, through the incorporation of three license 
conditions into its licenses, until a seismically qualified emergency 
cooling water source is provided that will provide the basis to extend 
the time for operator post-seismic cooling water load management. This 
could be done either through a seismic evaluation of the intake canal, 
physical modifications to the intake canal or plant, or some 
combination of the two.
    Date of issuance: March 25, 1997.
    Effective date: March 25, 1997, with implementation of License 
Condition 1 prior to Unit 2 entering Mode 2, with implementation of the 
requirements of License Condition 2 by July 1, 1997, and December 31, 
1998, and with implementation of License Condition 3 at the next 
updated safety analysis report update following completion of License 
Condition 2, but no later than June 1, 1999.
    Amendment Nos.: 128 and 120.
    Facility Operating License Nos. DPR-42 and DPR-60: Amendments 
revised the licenses.
    Public comments requested as to proposed no significant hazards 
consideration (NSHC): Yes (62 FR 5857 dated February 7, 1997). This 
notice provided an opportunity to submit comments on the Commission's 
proposed NSHC determination. The notice also provided for an 
opportunity to request a hearing by March 10, 1997, but indicated that 
if the Commission makes a final NSHC determination, any such hearing 
would take place after issuance of the amendments. Because of the 
significant revisions to the licensee's original application, NRC also 
published a public notice of the proposed amendments, issued a proposed 
finding of no significant hazards consideration, and requested that any 
comments on the proposed finding be provided to the staff by close of 
business on March 20, 1997. The notice was published in the St. Paul 
Pioneer Press on March 15, 1997, the Minneapolis Star Tribune on March 
16, 1997, and the Red Wing Republican Eagle on March 17, 1997. No 
comments have been received. The Commission's related evaluation of the 
amendments, finding of exigent circumstances, and final determination 
of NSHC are contained in a Safety Evaluation dated March 25, 1997.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    NRC Project Director: John N. Hannon.

    Dated at Rockville, Maryland, this 2nd day of April, 1997.

    For the Nuclear Regulatory Commission.
Jack W. Roe,
Director, Division of Reactor Projects III/IV, Office of Nuclear 
Reactor Regulation.
[FR Doc. 97-8916 Filed 4-8-97; 8:45 am]
BILLING CODE 7590-01-P