[Federal Register Volume 63, Number 77 (Wednesday, April 22, 1998)]
[Notices]
[Pages 19964-19989]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 98-10470]


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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 30, 1998, through April 10, 1998. The 
last biweekly notice was published on April 8, 1998 (63 FR 17219).

[[Page 19965]]

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administration 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 22, 1998, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff, or may be delivered to the Commission's

[[Page 19966]]

Public Document Room, the Gelman Building, 2120 L Street, NW., 
Washington, DC, by the above date. A copy of the petition should also 
be sent to the Office of the General Counsel, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, and to the attorney for the 
licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of amendment request: March 14, 1997.
    Description of amendment request: The proposed amendment would 
delete license conditions which have been satisfied, revise others to 
delete parts which are no longer applicable or to revise references, 
and make editorial changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The initial conditions and methodologies used in the accident 
analyses remain unchanged. The proposed changes do not change or 
alter the design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, accident 
analyses results are not impacted.
    The license conditions were one-time commitments that have been 
satisfied. There are no physical changes to the facility, and all 
operating procedures, limiting conditions for operation, limiting 
safety system settings, and safety limits are unchanged. Removal of 
these license conditions is appropriate and safe.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Many of the proposed changes delete references to items that 
have been completed. The NRC required these items as a condition of 
granting the license. Since they have been satisfied as intended, 
deleting them is administrative.
    None of the proposed changes affect the design or operation of 
any system, structure, 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 by the revised surveillance or testing 
requirements. The changes do not affect the manner by which the 
facility is operated and do not change any facility design feature, 
structure, system, or component. 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. 
The remaining changes are editorial in nature and have no impact on 
plant operation or design.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed changes to the Operating License are generally 
administrative in nature and have no impact on the margin of safety 
of any Technical Specification. There is no impact on safety limits 
or limiting safety system settings. The changes do not affect any 
plant safety parameters or setpoints. The operating license 
conditions have been satisfied, as required. There are no changes to 
the conditions themselves. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.
    Therefore, based on the above evaluation, Commonwealth Edison 
has concluded that these changes do not involve significant hazards 
considerations.

    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: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois

    Date of amendment request: October 16, 1997.
    Description of amendment request: The proposed amendment would add 
an exemption from 10 CFR 70.24(a) to the Unit 1 license consistent with 
the Unit 2 license.
    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:

    A. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The initial conditions and methodologies used in the accident 
analyses remain unchanged. The proposed change does not change or 
alter the design assumptions for the systems or components used to 
mitigate the consequences of an accident. Therefore, accidents 
analysis results are not impacted.
    There are no physical changes to the facility, and all operating 
procedures, limiting conditions for operation, limiting safety 
system settings, and safety limits are unchanged.
    The specific requirements for granting an exemption from 10 CFR 
70.24(a) have been met. The request is authorized by law, will not 
endanger life or property or the common defense and security, and is 
in the public interest.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    B. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect the design or operation of 
any system, structure, or component in the plant. The safety 
functions of structures, systems, or components are not changed in 
any manner, nor is the reliability of any structure, system, or 
component reduced by the revised surveillance or testing 
requirements. The change does not affect the manner by which the 
facility is operated and does not change any facility design 
feature, structure, system, or component. 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 change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    C. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change to the Operating License has no impact on 
the margin of safety of any Technical Specification. There is not

[[Page 19967]]

impact on safety limits or limiting safety system settings. The 
change does not affect any plant safety parameters or setpoints. 
Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    Therefore, based on the above evaluation, Commonwealth Edison 
has concluded that the proposed change does not involve significant 
hazards considerations.

    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: Byron Public Library District, 
109 N. Franklin, P.O. Box 434, Byron, Illinois 61010.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Stuart A. Richards.

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

    Date of amendment request: March 13, 1998, as supplemented March 
30, 1998.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to allow any two auxiliary 
feedwater (AFW) flow control valves to be inoperable concurrently for 
up to 72 hours, provided the corresponding redundant flow control 
valves and a pump in the other AFW train are operable.
    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:

    Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change would only alter the allowance for specific 
AFW flow control valves to be inoperable. It would not affect any 
operating limits, any plant operating conditions, or the physical 
capability of any plant equipment. Therefore, it would not affect 
the probability of any accident previously evaluated.
    The proposed change would not reduce the AFW flow capability to 
the steam generators during operation under the affected Action 
Statement. It would allow more operational flexibility in plant 
operation when two AFW flow control valves in the same train were 
concurrently inoperable. The specified AOT [allowed outage time] of 
72 hours would remain unchanged. Current TS allow continued 
operation for 72 hours with one of the three AFW pumps inoperable, 
or with one flow control valve in each train inoperable (provided 
the corresponding redundant flow control valve and a pump in the 
other pipe train are operable), but do not allow continued operation 
with both valves in the same train inoperable. The proposed change 
would allow any two valves to be inoperable, with the same provision 
that the corresponding redundant flow control valve and a pump in 
the other pipe train are operable.
    Since, with the proposed change there would be no reduction in 
the ability to provide AFW flow to either steam generator, operation 
of the Facility in accordance with the proposed changes would not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The changes do not alter the plant configuration (no new or 
different type of equipment will be installed) or make changes in 
the methods governing normal plant operation. The changes do allow 
different sets of AFW flow control valves to be inoperable, however, 
these changes retain a consistent level of AFW capability during 
operation under the Action Statement. Therefore, the changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    Therefore, operation of the Facility in accordance with the 
proposed TS change would not create the possibility of a new or 
different kind of accident from any previously evaluated.
    Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed change would not reduce the AFW flow capability to 
the steam generators during operation under the affected Action 
Statement. It would allow more operational flexibility in plant 
operation when two AFW flow control valves were concurrently 
inoperable. The specified AOT of 72 hours would remain unchanged.
    Therefore, operation of the Facility in accordance with the 
proposed TS change would not involve a significant reduction in a 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Energy 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: March 27, 1998 (NRC-98-0033).
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.5.2, ``ECCS--Shutdown,'' and TS 
3.5.3, ``Suppression Chamber,'' raising the minimum water level 
required in the condensate storage tank (CST) to support the core spray 
system (CSS) when the suppression pool (the normal supply for CSS) is 
unavailable. The amendment would also eliminate incorrect information 
concerning CST inventory reserved for the high pressure coolant 
injection (HPCI) and reactor core isolation cooling (RCIC) systems. The 
associated Bases are also revised.
    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 will not affect the performance or reliability of 
the Condensate Storage System which could lead to an accident 
because the Condensate Storage Tank (CST) is not involved as an 
initiator of any accident previously evaluated. The proposed change 
meets the design standards of the Condensate Storage System by 
providing assurance that sufficient water volume is available for 
the Core Spray System. This change also removes [an] erroneous 
discussion of water inventory for HPCI/RCIC Systems while in 
Operating [Operational] Conditions 4 and 5. The removal of 
information is acceptable since HPCI/RCIC Systems are not operable 
in these modes and will therefore not increase the probability of an 
accident. The increase in volume provides for vortex/air entrainment 
avoidance in the Core Spray System and will not increase 
consequences. Furthermore, the elimination of HPCI/RCIC information 
will not increase consequences of an accident previously evaluated 
because these systems are not credited for accident mitigation in 
Operating [Operational] Conditions 4 and 5.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not add or modify any equipment or 
components related to the Condensate Storage System and will 
therefore not create any new failure modes or common failure modes. 
This proposed change raises the water level within the CST to ensure 
sufficient water volume is maintained and updates the TS by removing 
descriptive information with respect to CST water inventory for 
HPCI/RCIC Systems

[[Page 19968]]

while in Operating [Operational] Conditions 4 and 5. The Condensate 
Storage System will continue to operate as intended and as designed. 
This change will therefore not create the possibility of a new or 
different kind of accident.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change increases the required CST water level to 
provide at least 150,000 gallons of water available for the Core 
Spray System while maintaining adequate submergence of the Core 
Spray standpipe for avoiding vortex and air entrainment. As such, 
the proposed change involves no reduction on any margin of safety. 
Revision to TS Bases concerning discussion of reserve volume in CST 
for HPCI and RCIC, does not alter the requirement for Core Spray or 
Suppression Pool operability and does not involve a reduction in any 
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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: March 27, 1998 (NRC-98-0034).
    Description of amendment request: The proposed amendment would 
clarify a footnote in Technical Specification (TS) 3.5.1, ``ECCS--
Operating,'' and 3.5.2, ``ECCS--Shutdown,'' to indicate that a low 
pressure coolant injection system loop may be considered operable 
during alignment and operation for decay heat removal if it is capable 
of being manually realigned and is not otherwise inoperable. The 
associated Bases would also be revised.
    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 involve actions required to realign the Low Pressure 
Coolant Injection (LPCI) system for LPCI injection if LPCI is 
required when operating in the Shutdown Cooling (SDC) mode. The 
additional actions described involve resetting isolations and trips 
which could occur prior to LPCI initiation. Resetting these logics 
does not initiate any valve operation or pump start; the LPCI 
initiation signals and interlocks remain in control of valve and 
pump logic.
    The equipment interlocks that provide the isolation signal for 
the LPCI injection valves were designed to prevent drain down of the 
Reactor Pressure Vessel (RPV) when in SDC. The injection valve 
closure is the most conservative action in response to an RPV drain 
event. The current TS acknowledges that operator action to realign 
the suction path is necessary. The proposed change acknowledges that 
operator action to reset injection valve logic and pump trips is 
necessary. The time required to realign LPCI is not significantly 
different than the existing actions to realign the suction path.
    No changes in either system design or operating strategies will 
be made as a result of these changes, thus no opportunity exists to 
increase the probability or consequences of a previously analyzed 
accident. Therefore, the proposed change does not involve a 
significant increase in the probability or consequences of a 
previously evaluated accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The manual realignment of the LPCI system from SDC following an 
isolation signal does not affect the accident analysis described in 
Chapter 15 of the UFSAR [updated final safety analysis report]. No 
new limiting single failure has been identified as a result of the 
proposed changes. The possibility of a new or different kind of 
accident from those previously analyzed will not be created by the 
change to the TS footnote or Bases, because the proposed change 
merely clarifies the actions necessary to realign the LPCI system. 
The time required to realign the system is not significantly 
different than the time necessary to realign the suction path. 
Therefore, no new or different types of failures or accident 
initiators are introduced by the proposed changes.
    3. The change does not involve a significant reduction in the 
margin of safety.
    The proposed change described above affects the plant's ability 
to enter Operational Conditions 3, 4, and 5, and to achieve and 
maintain COLD SHUTDOWN conditions when shutting down the plant. The 
proposed change in combination with existing restrictions within the 
TS provide assurance that there is no credible mechanism to inhibit 
running the LPCI system. The minor additional operator action 
required to realign LPCI from SDC requires minimum time and effort 
considering controls for each division are located on their 
respective control panel. As a result of this change, there will be 
no changes in either system design or operating strategies because 
the proposed changes merely clarify existing TS requirements and 
actions necessary to meet TS requirements. 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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: April 2, 1998 (NRC-98-0057).
    Description of amendment request: The proposed amendment would 
permit entering Operational Conditions 1 and 2 prior to completion of 
Surveillance Requirements for the primary containment hydrogen and 
oxygen monitors in order to establish the conditions necessary (inerted 
containment) to properly perform the calibrations. The amendment would 
also increase the frequency of the calibration for the oxygen monitors 
from every 18 months to quarterly in accordance with vendor 
recommendations and correct the nomenclature for the hydrogen and 
oxygen monitors in tables 3.3.7.5-1 and 4.3.7.5-1.
    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 changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change will permit delaying the performance of 
calibrations of the hydrogen and oxygen monitors until after the 
containment is inerted following a plant startup. The proposed 
change will also increase the calibration frequency for the oxygen 
monitors from once per 18 months to once per quarter, and change the 
nomenclature for the hydrogen and oxygen monitors.
    The primary containment hydrogen and oxygen monitors are passive 
instruments that provide indication to control room operators of 
hydrogen and oxygen concentration in the primary containment. 
Because they perform only a passive monitoring function, the

[[Page 19969]]

hydrogen and oxygen monitors are not associated with the initiation 
of any previously evaluated accident. The indication provided by the 
monitors is used by the control room operators to ensure oxygen 
concentration remains below limits and to make decisions regarding 
the use of the Combustible Gas Control System, if necessary. The 
allowance to permit entry into applicable operational conditions 
before calibration ensures that the conditions (nitrogen 
environment) are appropriate for accurate calibration of the 
instrument. Delaying the calibration does not cause the instrument 
to cease to function. Calibrations verify and adjust, as necessary, 
the accuracy of the instrument to compensate for drift that may 
occur since the last calibration. Thus, even with a delayed 
calibration, the instruments still would provide valuable 
information to the operators. Consequently, this change will not 
involve a significant increase in the consequences of a previously 
evaluated accident because the monitors will still function and 
provide meaningful information until the calibration is completed.
    The change to reduce the interval for calibration of the oxygen 
monitors from once per 18 months to once per quarter provides 
increased assurance of monitor accuracy and is consistent with the 
manufacturer's recommendations. Therefore, because this instrument 
is not associated with the initiation of an accident and the change 
improves the functionality of the instrument, the probability and 
consequences of previously evaluated accidents are not significantly 
affected.
    The change in nomenclature is editorial, and, as such does not 
affect the probability or consequences of a previously evaluated 
accident.
    2. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    As discussed above, the hydrogen and oxygen monitors are 
passive, indication-only instruments which provide information to 
control room operators. The proposed changes do not introduce a new 
mode of operation or involve a physical modification to the plant. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    3. The changes do not involve a significant reduction in the 
margin of safety.
    The proposed changes involve the containment hydrogen and oxygen 
monitors which do not affect any parameters or assumptions used in 
the calculation of any safety margin with regard to Technical 
Specification Safety Limits, Limiting Safety System Settings, 
Limiting Control Settings or Limiting Conditions for Operation, or 
other previously defined margins for any structure, system, or 
component. 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: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

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

    Date of amendment request: March 16, 1998.
    Description of amendment request: The proposed amendment would 
revise technical specification (TS) Table 4.3-1 to add footnote 6 to 
the channel calibration requirement for all instrument channels that 
are provided with an input from neutron flux detectors. Footnote 6 
provides that neutron detectors may be excluded from channel 
calibrations. Additional changes are proposed for BVPS-1 to provide 
consistency between BVPS-1 and BVPS-2. These additional changes would 
add channel calibration requirements to BVPS-1 TS Table 4.3-1 items 
2.b. (Power Range, Neutron Flux, Low Setpoint), 5. (Intermediate Range, 
Neutron Flux), 6. (Source Range, Neutron Flux (Below P-10), and 23. 
(Reactor Trip System Interlocks P-6, P-8, P-9, and P-10). Furthermore, 
changes would be made to correct page numbers in the BVPS-2 Index and 
to add corresponding changes to the bases 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?
    Overall protection system performance will remain within the 
bounds of the previously performed accident analyses since no 
hardware changes are proposed. The protection systems will continue 
to function in a manner consistent with the plant design basis. The 
proposed changes will not affect any of the analysis assumptions for 
any of the accidents previously evaluated. The proposed changes will 
not affect the probability of any event initiators nor will the 
proposed changes affect the ability of any safety-related equipment 
to perform its intended function. There will be no degradation in 
the performance of nor an increase in the challenges imposed on 
safety-related equipment assumed to function during an accident. 
There will be no change to normal plant operating parameters or 
accident mitigation capabilities. Therefore, the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    There are no hardware changes associated with this license 
amendment nor are there any changes in the method by which any 
safety-related plant system performs its safety function. The normal 
manner of plant operation is unchanged.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result 
of these changes. There will be no adverse effect or challenges 
imposed on any safety-related system 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 previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change does not involve a significant reduction in 
a margin of safety.
    The proposed changes do not affect the acceptance criteria for 
any analyzed event nor is there a change to any Safety Analysis 
Limit (SAL). Maintaining the SAL preserves the margin of safety.
    There will be no effect on the manner in which safety limits or 
limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment 
of protection functions. Therefore, the proposed changes do not 
involve a significant reduction in a margin of safety.

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

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 17, 1998.
    Description of amendment request: The proposed amendment would 
revise Action 34 of technical specification (TS)

[[Page 19970]]

Table 3.3-3, ``Engineered Safety Feature Actuation System 
Instrumentation.'' Action 34 applies to Functional Units 6.b., ``Grid 
Degraded Voltage (4.16 kV Bus),'' and 6.c. ``Grid Degraded Voltage (480 
v Bus).'' The proposed revision would require that with one channel 
inoperable, the inoperable channel be placed in the tripped condition 
within one hour; otherwise, the applicable action statement(s) for the 
associated emergency diesel generator made inoperable by the degraded 
voltage start instrumentation be entered immediately. The proposed 
revision would also require that with two channels inoperable, at least 
one of the two channels be restored to operable status and the other 
channel be placed in the tripped condition within one hour; otherwise, 
the associated emergency diesel generator shall be declared inoperable 
and its applicable action statement(s) shall be entered. In addition, 
corresponding changes would be made to the bases for TS 3/4.3.2 and the 
BVPS-2 Index would be revised to reflect changed page numbers.
    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?
    Engineered Safety Feature Actuation System (ESFAS) will continue 
to function in a manner consistent with the plant design basis. The 
proposed change will not affect any of the analysis assumptions for 
any of the accidents previously evaluated. The proposed changes will 
not affect the probability of any event initiators. There will be no 
change to normal plant operating parameters. The emergency bus 
degraded voltage protection system is utilized for accident 
mitigation and is not considered to be the source of accidents 
previously evaluated.
    Implementation of the proposed changes will now provide viable 
corrective actions which do not significantly increase the 
probability of failure of safety related equipment to perform its 
intended function. The proposed Action 34 permits a one hour time 
frame before the affected diesel generator(s) is required to be 
declared inoperable. This one hour period allows for repairs of most 
failures and takes into account the low probability of an event 
which would require the degraded voltage protection system to 
function. If adequate protection is not restored within this one 
hour period, the diesel generator(s) allowable outage time is 
invoked. The diesel generator(s) allowable outage time has been 
previously evaluated and determined to be an acceptable period of 
time during which plant operation may continue without an emergency 
backup power source. The loss of emergency bus degraded voltage 
protection is similar to the loss of the ability of an emergency 
diesel generator to provide electrical power to the safety related 
loads on the emergency buses. In both situations, a loss of offsite 
power, due to a total loss or a degraded condition, will result in 
the safety related loads not being capable of mitigating a design 
basis accident. The proposed changes to the Index page are 
administrative in nature and do not affect plant safety.
    Therefore, the proposed changes do not result in a significant 
increase in probability or consequences of an accident previously 
evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The emergency bus degraded voltage protection system is utilized 
for accident mitigation. The proposed changes will now provide 
viable corrective actions which do not result in a change in the 
manner in which the emergency bus loads are protected from a 
degraded voltage condition. These changes do not alter the function 
of the degraded voltage protection system. The proposed changes will 
continue to require that at least one of the two redundant 4160 volt 
or 480 volt emergency buses is protected from a degraded voltage 
condition assuming a single active failure of the opposite emergency 
bus degraded voltage protection system. This action will ensure that 
at least one train of engineered safety feature (ESF) equipment is 
not damaged due to a sustained bus undervoltage condition. The 
proposed addition of the requirement to enter the action statement 
for the inoperable diesel generator, if the one hour requirements of 
Action 34 cannot be met, will ensure that adequate compensatory 
actions to assure plant safety are taken. These requirements include 
the demonstration of the operability of the A.C. offsite sources by 
performing a specific surveillance within one hour and at least once 
per eight hours thereafter. If both diesel generators are 
inoperable, at least one diesel generator must be restored to 
operable status within two hours or the plant must be placed in cold 
shutdown within the following 36 hours.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of these changes. There 
will be no adverse effect or challenges imposed on any safety-
related system 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. Does the change involve a significant reduction in a margin 
of safety?
    The margin of safety is not significantly reduced because the 
A.C. electrical power sources will continue to provide sufficient 
capability, redundancy, and reliability to ensure availability of 
necessary power to ESF systems. The ESF systems will continue to 
function, as assumed in the safety analyses, to ensure that fuel, 
reactor coolant system and containment design limits are not 
exceeded. The proposed revisions to Action 34 will continue to 
require that at least one of the two redundant 4160 volt or 480 volt 
emergency buses is protected from a degraded voltage condition 
assuming a single active failure of the opposite emergency bus 
degraded voltage protection system. This action will ensure that at 
least one train of ESF equipment is not damaged due to a sustained 
bus undervoltage condition. The emergency loads, which are powered 
from that train of emergency buses, will continue to be available to 
perform their safety related functions. If the one hour requirements 
of Action 34 cannot be met, the affected emergency diesel generator 
will be declared inoperable. This will ensure that adequate 
compensatory actions to ensure plant safety are taken. The loss of 
emergency bus protection from a degraded voltage condition is 
similar to the loss of the ability of an emergency diesel generator 
to provide electrical power to the safety related loads on the 
emergency buses. In both situations, a loss of the offsite power, 
due to a total loss or a degraded condition, will result in the 
safety related loads not being capable of mitigating a design basis 
accident.
    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: 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: Robert A. Capra.

GPU Nuclear Corporation, et al., Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit No. 1, Dauphin County, Pennsylvania

    Date of amendment request: March 23, 1998.
    Description of amendment request: The proposed amendment modifies 
Section 3.1.2 of the Technical Specifications (TS) to incorporate new 
pressure/temperature limits regarding reactor vessel pressurization 
heatup, cooldown, and inservice leak and hydrostatic leak test 
limitations in accordance with 10 CFR 50, Appendix G. These new limits 
would be applicable through the period of 17.7 effective full power 
years (EFPY) of operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards

[[Page 19971]]

consideration, which is presented below:

    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The design basis event related to this change is 
nonductile failure of the reactor coolant pressure boundary. The 
updated pressure/temperature limits have been established in 
accordance with the requirements of 10 CFR 50, Appendix G. Revision 
of these curves for an applicability period of 17.7 EFPY is based on 
maintaining the required design margin. Operation of the facility in 
accordance with the proposed amendment provides assurance of 
protection against nonductile failure of the reactor coolant 
pressure boundary for operation through 17.7 EFPY. Therefore, 
operation in accordance with the proposed amendment does not involve 
a significant increase in the probability of occurrence or 
consequences of an accident previously evaluated.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The design basis 
event related to the change is nonductile failure of the reactor 
coolant boundary. The proposed amendment provides assurance of 
protection against nonductile failure of the reactor coolant 
boundary for operation through 17.7 EFPY and is unrelated to the 
possibility of creating a new or different kind of accident.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve any reduction in a margin of safety 
since the design methodology has maintained the existing margins.

    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: Law/Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Walnut 
Street and Commonwealth Avenue, Box 1601, Harrisburg, PA 17105.
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Cecil O. Thomas.

Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear 
Station, Nemaha County, Nebraska

    Date of amendment request: March 27, 1997.
    Description of amendment request: The proposed amendment, included 
as part of the proposed conversion from the current Technical 
Specifications (TS) to improved TS, would revise the Limiting 
Conditions for Operation in the event that one 250 V DC electrical 
power subsystem is 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. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The DC electrical power sources are used to support mitigation 
of the consequences of an accident; however, they are not considered 
the initiator of any previously analyzed accident. The proposed 
change merely provides direction to the operator to declare 
equipment associated with a 250 V DC electrical power subsystem 
inoperable if the subsystem becomes inoperable. This provides 
assurance that all affected features are immediately recognized as 
incapable of performing their safety functions, and requires 
immediate actions equivalent to those determined appropriate in the 
Technical Specifications for the affected features. Therefore, the 
proposed change does not involve an increase in the probability or 
consequences of any accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not introduce a new mode of plant 
operation and does not involve physical modification to the plant. 
Therefore, the possibility of a new or different kind of accident 
from any accident previously evaluated is not created.
    3. Does this change involve a significant reduction in a margin 
of safety?
    This change does not involve a significant reduction in a margin 
of safety, since the proposed change results in establishing the 
level of safety for the loss of a 250 V DC electrical power 
subsystem equivalent to the level of safety that exists in the 
Technical Specifications for components and systems that are 
supplied by the 250 V DC electrical power system.

    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: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.
    Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power 
District, Post Office Box 499, Columbus, NE 68602-0499.
    NRC Project Director: John N. Hannon.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 2, 1998.
    Description of amendment request: The proposed changes would revise 
the frequency for the performance of specific surveillances associated 
with the emergency diesel generators (EDGs) and delete the requirements 
contained in the current Technical Specifications for accelerated 
testing whenever the number of valid test failures associated with the 
EDGs is met or exceeded. In addition, the special requirements for 
reporting valid or invalid EDG failures would be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect accident initiators or 
precursors and do not alter the design assumptions affecting the 
ability of the EDGs to mitigate the consequences of an accident.
    Industry experience has indicated that excessive testing 
requirements have proven to be a contributor to increased EDG 
unavailability and equipment degradation. Removing inappropriate 
testing requirements increases EDG reliability and enhances the 
ability of EDGs to mitigate the consequences of an accident. 
Implementing the maintenance rule in accordance with 10 CFR 50.65, 
Regulatory Guide 1.160, and NUMARC 93-01 for the EDGs provides 
additional assurance that high EDG performance and availability will 
be maintained.
    Deleting the special reporting requirements from the Technical 
Specifications is an administrative change that does not affect the 
ability of the EDGs to perform their specified safety function. 
North Atlantic will continue to notify the NRC of significant EDG 
failures in accordance with the provisions of 10 CFR 50.72 and 
50.73.
    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 previously analyzed.
    The proposed changes do not alter the ability of the EDGs to 
perform their intended function to mitigate the consequences of an 
initiating event within the acceptance limits assumed in the Updated 
Final Safety Analysis Report (UFSAR). The proposed changes have no 
impact on component or system interactions, or the plant design 
basis. Instrumentation setpoints, starting,

[[Page 19972]]

sequencing and loading functions associated with the EDGs are not 
affected by the proposed changes. Furthermore, combining the 
implementation of the maintenance rule program with the proposed 
amendment will enhance both the availability and the performance of 
the EDGs.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There is no impact on equipment design or operation and there 
are no changes being made to the Technical Specification required 
safety limits or safety system settings that would adversely affect 
plant safety. The proposed changes do not affect the EDG's ability 
to ensure that sufficient power is available to supply the safety 
related equipment required for: 1) the safe shutdown of the 
facility, and 2) the mitigation and control of accident conditions 
within the facility. In addition, the proposed changes do not affect 
the EDG's ability to ensure that: 1) the facility can be maintained 
in a shutdown or refueling condition for extended periods of time, 
and 2) sufficient instrumentation and control capability is 
available for monitoring and maintaining the unit status.
    EDG reliability and availability are expected to be improved by 
the proposed changes. Eliminating excessive testing requirements can 
improve safety by reducing challenges to plant systems and reducing 
equipment wear and degradation. While the proposed changes affect 
surveillance intervals there are no changes to the methods used to 
perform the surveillances. The surveillances will continue to 
demonstrate the ability of the EDGs to perform their intended 
function of providing electrical power to the emergency safety 
systems needed to mitigate design basis transients and accidents. 
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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 5, 1998.
    Description of amendment request: The proposed changes would revise 
the Seabrook Station Radiological Effluent Technical Specifications 
(TS) and Administrative Controls section of the Technical 
Specifications, as authorized by NRC Generic Letter (GL) 89-01, 
``Implementation Of Programmatic Controls For Radiological Effluent 
Technical Specifications (RETS) In The Administrative Controls Section 
Of The Technical Specifications And The Relocation Of Procedural 
Details of RETS To The Offsite Dose Calculation Manual Or To The 
Process Control Program.'' The proposed amendment would incorporate 
programmatic controls in the TSs for radioactive effluents and for 
environmental monitoring conforming to the applicable regulatory 
requirements and would relocate the existing procedural details of the 
current RETS to the Offsite Dose Calculation Manual (ODCM). Procedural 
details associated with solid radioactive wastes would be relocated to 
the Process Control Program (PCP).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not affect accident initiators or 
precursors and do not alter the design assumptions, conditions, 
configuration of the facility or the manner in which the plant is 
operated. The proposed changes do not alter or prevent the ability 
of structures, systems, or components (SSCs) to perform their 
intended function to mitigate the consequences of an initiating 
event within the acceptance limits assumed in the Updated Final 
Safety Analysis Report (UFSAR). The proposed changes are 
administrative in nature and do not change the level of programmatic 
controls and procedural details relative to radiological effluents.
    Incorporation of programmatic controls for RETS in TSs will 
assure that the applicable regulatory requirements pertaining to the 
control of radioactive effluents will continue to be maintained. 
Since there are no changes to previous accident analyses, the 
radiological consequences associated with these analyses remain 
unchanged, 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 previously analyzed.
    The proposed changes do not alter the design assumptions, 
conditions, configuration of the facility or the manner in which the 
plant is operated. The proposed changes have no impact on component 
or system interactions. The proposed changes are administrative in 
nature and do not change the level of programmatic controls and 
procedural details relative to radiological effluents. Therefore, 
the proposed changes do not create the possibility of a new or 
different kind of accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There is no impact on equipment design or operation and there 
are no changes being made to the Technical Specification required 
safety limits or safety system settings that would adversely affect 
plant safety. The proposed changes are administrative in nature and 
do not change the level of programmatic controls and procedural 
details relative to radiological effluents. A comparable level of 
administrative control will continue to be applied to those design 
conditions and associated surveillances being relocated to the ODCM 
or PCP. Therefore, the proposed changes do not involve a significant 
reduction in any 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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 23, 1998.
    Description of amendment request: The proposed change would revise 
the Seabrook Station Technical Specifications (TSs) to add a new TS 
3.0.5 that would provide an exception to TSs 3.0.1 and 3.0.2 to allow 
the performance of required testing to demonstrate the operability of 
the equipment being returned to service or the operability of other 
equipment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The design basis accidents are not affected by the proposed 
administrative changes.

[[Page 19973]]

    Specification 3.0.5 provides the administrative controls to 
ensure the time the equipment is returned to service in conflict 
with the requirements of the ACTIONS is limited to the time 
absolutely necessary to perform the allowed required testing. 
Specification 3.0.5 was incorporated in NUREG-1431, ``Standard 
Technical Specifications--Westinghouse Plants,'' (as modified by 
approved Technical Specification Task Force (TSTF) generic change 
Traveler TSTF-165), to address these, and other similar situations, 
that conflict with the requirements with the ACTIONS when equipment 
is returned to service. Specification 3.0.5 does not provide time to 
perform other preventative or corrective maintenance.
    Inclusion of Specification 3.0.5 into the Seabrook Station 
Technical Specifications will provide operational flexibility with 
the restrictive compliance requirements of the other Applicability 
Specifications (3.0.1 and 3.0.2) and allow the performance of post-
maintenance/surveillance activities to facilitate returning 
equipment to service or to allow other equipment to be tested. 
Therefore, inclusion of Specification 3.0.5 into the Seabrook 
Station Technical Specifications enhances plant safety by minimizing 
the potential for plant trip and/or transients. A qualitative risk 
assessment concerning returning components to service for post-
maintenance testing was performed and concluded that the 
configurations allowed by Specification 3.0.5 have a negligible 
effect on the Seabrook Station risk profile. The components involved 
will have either completed calibration or maintenance, and can 
reasonably be expected to be able to perform their required safety 
function when returned to service for testing purposes. 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 previously analyzed.
    The proposed changes do not introduce new features or modify 
plant structures, systems and components or procedures that could 
possibly affect station operations under normal or abnormal 
conditions, thus, the potential for an unanalyzed accident is not 
created. The proposed administrative changes have no adverse affect 
on the safety limits or design basis accidents. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There are no changes being made to the Technical Specification 
safety limits or safety system settings that would adversely affect 
plant safety. The changes do not affect the operation of structures, 
systems or components (SSCs) nor do they introduce administrative 
changes to plant procedures that could affect operator response 
during normal, abnormal or emergency situations. Inclusion of 
Specification 3.0.5 into the Seabrook Station Technical 
Specifications enhances plant safety by minimizing the potential for 
plant trip and/or transients by allowing equipment to be returned to 
service. A qualitative risk assessment concerning the return of 
components to service for post-maintenance testing was performed and 
concluded that the configurations allowed by Specification 3.0.5 
have a negligible effect on the Seabrook Station risk profile. The 
components involved will have either completed calibration or 
maintenance, and can reasonably be expected to be able to perform 
their required safety function when returned to service for testing 
purposes. 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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: March 27, 1998.
    Description of amendment request: The proposed change would revise 
Technical Specification (TS) 3.7.6, ``Control Room Emergency Makeup Air 
and Filtration (CREMAFS).'' The proposed change would modify the 
existing required action when both trains of CREMAFS are inoperable in 
Modes 5 and 6 by eliminating the restriction of suspending positive 
reactivity 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 proposed changes have no impact on the probability or 
consequences of an accident previously evaluated in the UFSAR. The 
control room ventilation systems are support systems which have a 
role in the detection and mitigation of accidents but do not 
contribute to the initiation of any accident previously evaluated. 
The removal of the positive reactivity addition restriction in Mode 
5 and 6 has no impact on the course of any accidents previously 
evaluated. There are no presently evaluated positive reactivity or 
boron dilution accidents that credit the CREMAFS to mitigate its 
consequences or provide radiological protection. The positive 
reactivity restriction is overly restrictive in that it does not 
allow cooldown below 200 deg. F when Mode 5 is entered as a result 
of both trains of CREMAFS being inoperable nor does it allow Reactor 
Coolant System temperature to vary.
    The restriction is also redundant to Technical Specification 
3.1.1.2 ``Reactivity Control Systems Shutdown Margin-Tavg 
less than or equal to 200 deg. F'' in Mode 5 and Technical 
Specification 3.9.1 ``Refueling Operations Boron Concentration'' in 
Mode 6. Technical Specification 3.1.1.2 action, with shutdown margin 
less than the limit specified in the Core Operating Limits Report or 
with the Reactor Coolant System boron concentration less than 2000 
ppm boron, requires immediate and continued boration until the 
restoration of the required shutdown margin or boron concentration. 
Similarly, Technical Specification 3.9.1 actions require suspension 
of core alterations or positive reactivity changes in addition to 
immediate and continued boration until the restoration of the 
required shutdown margin (Keff) or boron concentration 
while in Mode 6. Sufficient shutdown margin ensures that (1) the 
reactor can be made subcritical from all operating conditions, (2) 
the reactivity transients associated with the postulated accident 
conditions are controllable within acceptable limits and (3) the 
reactor will be maintained sufficiently subcritical to preclude 
inadvertent criticality in the shutdown condition. The above 
referenced reactivity control system specifications provide the 
necessary protection for postulated reactivity addition accident 
conditions. Therefore, modifying the Technical Specification action 
that requires the suspension of positive reactivity changes and core 
alterations with both trains of the CREMAFS inoperable does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed change that removes the positive reactivity 
addition restriction in Mode 5 and 6 does not create the possibility 
of a new accident nor does it create the possibility of a different 
kind of accident previously evaluated. There are no presently 
evaluated positive reactivity or boron dilution accidents that 
credit the CREMAFS to mitigate its consequences or provide 
radiological protection. The addition of positive reactivity during 
the above described situation is overly restrictive and furthermore 
redundant to Technical Specification 3.1.1.2 ``Reactivity Control 
Systems Shutdown Margin-Tavg less than or equal to 
200 deg. F'' in Mode 5 and Technical Specification 3.9.1 ``Refueling 
Operations Boron Concentration'' in Mode 6. The above referenced 
reactivity control system specifications provide the necessary 
protection for postulated reactivity addition accident conditions. 
Therefore, modifying the Technical Specification action that 
requires the suspension of positive

[[Page 19974]]

reactivity changes and core alterations with both trains of the 
CREMAFS inoperable does not create the possibility of a new or 
different accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The changes being proposed do not revise equipment design or 
operation nor do they make changes to Technical Specification 
required safety limits or safety system settings. In addition, they 
do not alter the environmental conditions which are to be maintained 
in the control room during normal operation and following an 
accident and they do not revise the accident analyses. Therefore, 
the proposed changes do not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis, and based on 
this review, it appears that the three standards of 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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: April 3, 1998.
    Description of amendment request: The proposed change would revise 
the Seabrook Station Technical Specifications (TSs) with administrative 
changes to support phased implementation of 24-month fuel cycle 
surveillance interval extensions. Specifically, the proposed change 
would: (1) provide wording changes in the Bases Section of TS 4.0.2 
necessary to support 24-month surveillance interval extensions, (2) 
revise TS 4.0.5.b to provide revised terminology for inservice 
inspection and testing activities and their associated frequencies, (3) 
revise TS Table 1.1 to clarify current and future refueling intervals 
and their associated surveillance requirements and frequencies, and (4) 
delete the ``during shutdown'' restriction from the performance 
requirements of certain surveillance requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The design basis accidents are not affected by the proposed 
editorial and administrative changes. The proposed changes do not 
change the level of programmatic controls or the procedural details 
currently in place. Furthermore, these changes have no adverse 
affect to the safe operation of the station. Performance of certain 
maintenance and testing activities during conditions or modes other 
than shutdown will be evaluated by North Atlantic to ensure proper 
regard to their effect on safe operation of the plant is given prior 
to conduct of a particular surveillance, or portion thereof. 
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 previously analyzed.
    The proposed changes do not introduce new features or modify 
plant structures, systems and components or procedures that could 
possibly affect station operations under normal or abnormal 
conditions, thus, the potential for an unanalyzed accident is not 
created. Performance of maintenance and testing activities on-line, 
as well as shutdown, are controlled by North Atlantic's procedures 
and policies to perform reviews and assessments of these activities 
to determine the affect on safe operation of the facility. The 
proposed editorial and administrative changes have no adverse affect 
on the safety limits or design basis accidents. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously analyzed.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    There are no changes being made to the Technical Specification 
safety limits or safety system settings that would adversely affect 
plant safety. The changes do not affect the operation of structures, 
systems or components nor do they introduce administrative changes 
to plant procedures that could affect operator response during 
normal, abnormal or emergency situations. Performance of certain 
maintenance and testing activities during conditions or modes other 
than shutdown will be evaluated by North Atlantic to ensure proper 
regard to their effect on safe operation of the plant is given prior 
to conduct of a particular surveillance, or portion thereof. 
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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Cecil O. Thomas.

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

    Date of amendment request: April 1, 1998.
    Description of amendment request: The proposed revision to the 
Millstone Unit 3 licensing basis would add a new sump pump subsystem to 
address groundwater inleakage through the containment basemat.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    NNECO has reviewed the proposed revision in accordance with 10 
CFR 50.92 and has concluded that the revision does not involve a 
significant hazards consideration (SHC). The bases for this 
conclusion is that the three criteria of 10 CFR 50.92(c) are not 
satisfied. The proposed revision does not involve an SHC because the 
revision would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The current FSAR [Final Safety Analysis Report] credits the 
waterproof membrane for assuring that groundwater inleakage is not 
significant and would have no impact on safety related structures 
and components. However, degradation of the waterproof membrane has 
been detected, and it is now concluded that groundwater inleakage 
can be significant in that it could affect the operability of the 
RSS [recirculation spray system] pumps. The original plant design 
had only nonsafety-related RSS sump pumps available for pumping the 
groundwater from the RSS sumps. These pumps are not powered from the 
emergency busses and would not be accessible during a design basis 
LOCA [loss-of-coolant accident].
    Thus, it is assumed that they would not be available to mitigate 
a design basis accident. Two independent safety-related air-driven 
sump pumps have been installed to eliminate the potential for 
groundwater inleakage that would affect the RSS pumps.
    Air-driven sump pumps have been installed with the air supply 
line routed to a connection outside the ESF [engineered safety 
features] building. This allows the installation of an air 
compressor in an area that is accessible during a design basis 
accident such as a LOCA. Two air compressors have been staged in 
designated locations, and will be maintained and

[[Page 19975]]

periodically tested to ensure their availability. Periodic testing 
of the sump pumps will also be performed. The surveillance 
requirements have been incorporated into the Technical Requirements 
Manual.
    EOP [Emergency Operating Procedure] 35-ES1.3 has been modified 
to add a step to install the compressors and start the sump pumps. 
It is estimated that these sump pumps would be needed approximately 
ten hours after a design basis accident. Thus, there is sufficient 
time for the operators to perform this action. Since sufficient time 
is available, the action has been incorporated into procedures and 
the environmental conditions allow access to the area, it is 
concluded that credit for operator action can be taken.
    Thus, the new system is single failure proof and meets the 
requirements of Standard Review Plan 3.4.1 which stated the 
following:
    ``If safety-related structures are protected from below-grade 
groundwater seepage by means of a permanent dewatering system, then 
the system should be designed as a safety-related system and meet 
the single failure proof criterion.''
    This provides assurance that the RSS pumps and other safety-
related structures and components will perform the required safety 
function as assumed in the accident analysis.
    The current nonsafety-related RSS sump pump system will continue 
to provide protection from groundwater inleakage during normal 
operation. Thus, there is no impact on the probability of occurrence 
of a transient because of equipment or structural failure due to 
groundwater inleakage. In addition, the new safety-related RSS sump 
pump system provides additional assurance that groundwater inleakage 
would not affect structures or equipment during an extended loss of 
offsite power or a design basis accident. Thus, it is concluded that 
there is no impact on the probability of occurrence of any 
previously evaluated accident.
    The change results in the use of the new air-driven sump pumps 
to remove groundwater in-leakage from the RSS cubicles. To preclude 
the possibility for radiological contamination of the groundwater, 
all sources of liquid radiological contamination to the sumps have 
been eliminated. The RSS cubicle floor drains leading to Sumps 7A/7B 
have been plugged. Drains from equipment determined not to be a 
potential source of radiological contamination continue to drain to 
Sumps 7A/7B (sources include CCP [component cooling water] and 
Service Water relief valves) and are covered with splash guards to 
prevent the entrance of contaminated spray. The Hydrogen Recombiner 
area floor drains and the drain from the PASS [post accident 
sampling system] sample sink, all of which are nonsafety-related, 
have been isolated from the indirect waste receptor which drains to 
Sump 7B. Sumps 7A and 7B have been cleaned and the existing 
nonsafety-related sump pumps replaced to remove any existing 
residual contamination. The nonsafety-related pumps (3DAS-P8A/B) 
discharge to ESF Building sump 3DAS-SUMP 10. To preclude any 
potential siphoning from the potentially contaminated Sump 10 back 
to Sumps 7A/7B, the lines of the existing nonsafety-related pumps 
have been shortened to discharge above the water level in Sump 10.
    The walls of Sumps 7A/7B have been extended to protect from a 
Limited Passive Failure and Pipe Break in the RSS cubicles. The 
expected flooding height is 6.6 inches [ ]. The sump cubicle height 
was extended to 3 ft. above the cubicle floor, well above this 
height. The sumps are covered with a vented hood to protect from 
pipe break spray and miscellaneous overhead leaks to further assure 
the sumps remain isolated from potentially contaminated RSS system 
fluids.
    The existing SLCRS [supplementary leak collection and release 
system] boundary has been extended to the isolation valves located 
outside of the ESF building. Additionally, when the sump level is 
reduced while using the air driven pump, the pumps are designed to 
prevent air from being discharged through the pump discharge outside 
of the ESF building.
    Thus, use of the new sump pumps would not affect the offsite 
doses following a design basis accident.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequence of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The current nonsafety-related RSS sump pump system will continue 
to provide protection from groundwater inleakage during normal 
operation. This will continue to provide assurance there is no 
potential for a transient because of equipment or structural failure 
due to groundwater inleakage. In addition, the new safety-related 
RSS sump pump system provides additional assurance that groundwater 
inleakage would not affect structures or equipment during an 
extended loss of offsite power or a design basis accident.
    Therefore, the proposed revision does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety.
    The current FSAR credits the waterproof membrane for assuring 
that groundwater inleakage is not significant and would have no 
impact on safety related structures and components. However, 
degradation of the waterproof membrane has been detected and it is 
now concluded that groundwater inleakage can be significant in that 
it could affect the operability of the RSS pumps. Original design 
had only nonsafety-related RSS sump pumps available for pumping the 
groundwater from the RSS sumps. These pumps are not powered from the 
emergency busses and would not be accessible during a design basis 
LOCA. Thus, it is assumed that they would not be available to 
mitigate a design basis accident. Two independent safety-related 
air-driven sump pumps have been installed to eliminate the potential 
for groundwater inleakage that would affect the RSS pumps. The new 
system is single failure proof and meets the requirements of 
Standard Review Plan 3.4.1.
    Use of the new system requires operator action to install pre-
staged air compressors to provide power for the new air-driven sump 
pumps. It is estimated that these sump pumps would be needed 
approximately ten hours after a design basis accident. Thus, there 
is sufficient time for the operators to perform this action. Since 
sufficient time is available, the action has been incorporated into 
procedures and the environmental conditions allow access to the 
area, it is concluded that credit for operator action can be taken.
    With credit for the new single failure proof air-driven sump 
pumps and operator action to install pre-staged compressors to 
provide power for the pumps, the new subsystem provides the required 
assurance that the RSS pumps will not be affected by groundwater 
inleakage. Thus, it is concluded that the RSS pumps would be 
operable for long term accident mitigation and there is no impact on 
the margin of safety as defined in the basis of the Emergency Core 
Cooling Technical Specifications or any other Technical 
Specification.
    Therefore, the proposed revision does not involve a significant 
reduction in a margin of safety.

    In conclusion, based on the information provided, it is determined 
that the proposed revision does not involve an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
Connecticut.
    NRC Deputy Director: Phillip F. McKee.

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: December 23, 1997.
    Description of amendment request: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Power Plant, Unit Nos. 1 and 2 to revise Technical Specification (TS) 
3/

[[Page 19976]]

4.7.11, Table 3.7-1, ``Maximum Allowable Power Range Neutron Flux High 
Setpoint With Inoperable Steam Line Safety Valves.'' The power range 
(PR) neutron flux high setpoints would be changed based on revised 
calculational methodologies for 1, 2, or 3 inoperable MSSVs per steam 
generator (SG). The proposed TS change would lower the PR neutron flux 
high setpoints when 2 or 3 MSSVs are inoperable per loop such that the 
maximum power level allowed would be within the heat removing 
capability of the remaining operable MSSVs. Although the method for 
calculating the maximum power level allowed when one MSSV per loop is 
inoperable has been revised, the results have not and the limit remains 
the same. The associated Bases would also be revised.
    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 reduction in the power range (PR) neutron flux high setpoint 
Technical Specification (TS) values does not initiate an accident. 
Technician adjustments to lower the PR neutron flux high setpoints 
could cause a reactor trip (RT). However, this action is already a 
TS requirement. Thus, reducing the TS setpoint values from their 
current values will not change the requirement for a technician to 
adjust the setpoints downward when main steam safety valves (MSSVs) 
become inoperable, and therefore, will not increase the probability 
of a RT.
    The reduction of the setpoints assures that the consequences of 
an accident when the MSSVs are inoperable are not affected by 
assuring that the MSSVs will continue to prevent overpressure of the 
main steam leads and steam generators (SGs) and remove adequate heat 
for the reactor coolant system.
    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.
    Reduction of the PR neutron flux high setpoints does not change 
the method by which any safety-related system performs the 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 change does not involve a significant reduction 
in a margin of safety.
    With the reduction in the PR neutron flux high setpoints for 
inoperable MSSVs, the MSSVs will still prevent SG pressure from 
exceeding 110 percent of SG design pressure in accordance with the 
ASME code. The change is conservative. The conclusions for the Final 
Safety Analysis Report Update accident analyses are unaffected by 
the change, remain valid, and provide margin.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

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

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

    Date of amendment request: February 6, 1998.
    Description of amendment request: The proposed changes would revise 
the Reactor Protection System (RPS) Normal Supply Electrical Protection 
Assembly (EPA) Undervoltage Trip setpoint to reflect a reanalysis of 
the most limiting applied load minimum voltage requirements.
    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 proposed setpoint change evaluated in Section III does not 
involve any physical changes to the plant, does not alter the way 
these systems function, and will not degrade the performance of the 
plant safety systems. The proposed instrument setpoint changes 
ensures that plant safety limits are not exceeded for the most 
limiting voltage requirements. The type of testing and the 
corrective actions required if the subject surveillances fail 
remains the same. The proposed changes do not adversely affect the 
reliability of these systems or affect the ability of the systems to 
meet their design objectives. A historical review of surveillance 
test results supports these conclusions.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed setpoint change evaluated in Section III does not 
modify the design or operation of the plant, therefore, no new 
failure modes are introduced. The proposed instrument setpoint 
change ensures that plant safety limits are not exceeded for the 
most limiting voltage requirements. No changes are proposed to the 
type and method of testing performed. A historical review of 
surveillance test results supports these conclusions.
    3. Involve a significant reduction in a margin of safety.
    The proposed setpoint change evaluated in Section III results in 
minimal impact on system reliability in the interval between 
surveillance tests. This is based on the redundant design of the 
evaluated systems. A review of past surveillance history has shown 
no evidence of failures which would significantly impact the 
reliability of these systems. Operation of the plant remains 
unchanged by this proposed setpoint change. The assumptions in the 
Plant Licensing Basis are not adversely impacted. Therefore, the 
proposed changes do not result in a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: February 6, 1998.
    Description of amendment request: The proposed changes would allow 
reactor coolant system pressure tests to be conducted in Cold Shutdown 
Mode. Primary containment integrity is not required in this mode, 
facilitating containment access for inspections. The proposed changes 
also allow some outage activities on other systems to continue during 
the pressure testing. The licensee claims the proposed changes are 
consistent with the Boiling Water Reactor Standard Technical 
Specifications given in NUREG-1433, Revision 1.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 19977]]

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 a leak in the reactor coolant pressure 
boundary during reactor coolant system pressure testing is not 
increased by considering the reactor to be in Cold Shutdown. Since 
the pressure tests are performed nearly water solid, at low decay 
heat values, and near Cold Shutdown conditions, the stored energy in 
the reactor core will be low. Under these conditions, the potential 
for failed fuel and a subsequent increase in coolant activity is 
minimized. In addition, Special Operations LCO [Limiting Condition 
for Operation] 3.12.A requires supporting LCOs for ECCS [emergency 
core cooling system]-Cold Condition, Standby Gas Treatment, 
Secondary Containment isolation and Standby Gas Treatment initiation 
instrumentation, and Auxiliary Electrical Systems to be met to 
ensure secondary containment integrity is maintained and capable of 
handling any airborne radioactivity or steam leaks that could occur 
during the performance of hydrostatic or leak testing. A listing of 
secondary containment isolation valves required to maintain 
Secondary Containment Integrity is included in plant controlled 
procedures. The required pressure testing conditions provide 
adequate assurance that the consequences of a steam leak will be 
conservatively bounded by the consequences of the postulated main 
steam line break outside of primary containment. In the event of a 
large primary system leak, the reactor vessel would rapidly 
depressurize, allowing the low pressure core cooling systems to 
operate. The capability of these systems would be adequate to keep 
the core flooded under this low decay heat load condition. Small 
system leaks would be detected by leakage inspections before 
significant inventory loss occurred. Therefore, the consequences of 
an accident previously evaluated are not significantly increased.
    2. Create the possibility of a new or different kind of accident 
from those previously evaluated.
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms since the changes do not involve 
any changes to structures, systems, or components, do not involve 
any change to the operation of systems, and alter procedures only to 
the extent that the 212 deg. F limit may be exceeded during reactor 
coolant system pressure testing with certain systems inoperable. 
There are no alterations to plant systems designed to mitigate the 
consequences of accidents. The only difference is that a different 
subset of plant systems would be utilized for accident mitigation 
than those utilized during the Hot Shutdown Mode. Therefore, the 
proposed changes do 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.
    Since pressure tests are performed nearly water solid, at low 
decay heat values, and near Cold Shutdown conditions, the stored 
energy in the reactor core will be low. Under these conditions, the 
potential for failed fuel and a subsequent increase in coolant 
activity is minimized. Since secondary containment integrity will be 
maintained, in accordance with the Special Operations LCO, the 
secondary containment will be capable of handling any airborne 
radioactivity or steam leaks that could occur during the performance 
of hydrostatic or leak testing. 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: February 26, 1998.
    Description of amendment request: The proposed changes would change 
the allowed containment leakage rate to 1.5 percent per day, changes 
the assumed standby gas treatment system (SBGT) filter efficiency, and 
revises reactor coolant sampling requirements for low Iodine-131 
concentrations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The Authority [the licensee] has evaluated the proposed TS 
[technical specification] Amendment and determined that it does not 
represent a significant hazards consideration. Based on the criteria 
for defining a significant hazards consideration established in 10 CFR 
50.92, operation of the James A. FitzPatrick Nuclear Power Plant in 
accordance with the proposed amendment will not:

    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    The proposed changes do not involve a change to the design or 
operation of the plant. The systems affected by this proposed TS 
change are not assumed in any safety analyses to initiate any 
accident sequence. Therefore, the probability of any accident 
previously evaluated is not increased by this proposed TS change. 
The change in the allowable containment leakage rate (La) 
is consistent with the accident analyses. The assumption of only 90% 
SBGT filter efficiency is conservative with actual system 
performance and is consistent with Regulatory Guide 1.52. There is 
no significant change to the consequences of an accident previously 
evaluated because maintaining containment leakage within limits 
assumed in the accident analyses ensures that the dose consequences 
resulting from an accident are not increased. The calculated doses 
with the decreased SBGT system charcoal efficiency for design basis 
accidents are marginally increased but still meet, and are well 
below, the dose acceptance criteria of 10 CFR 100, the SRP [Standard 
Review Plan, NUREG-0800], and GDC [General Design Criterion] 19 of 
Appendix A to 10 CFR 50. The proposed TS changes maintain an 
equivalent level of reliability and availability for all affected 
systems. The ability of the affected systems associated with 
maintaining leak rate integrity to perform their intended function 
is unaffected by the proposed TS changes. Implementation of these 
changes will provide continued assurance that specified parameters 
associated with containment integrity will remain within acceptance 
limits, and as such, will not significantly increase the 
consequences of a previously evaluated accident. The change in the 
value of .007 [microcurie]/ml to .002 [microcurie]/ml in section of 
4.6.C. ``Coolant Chemistry'' is a minor editorial change, is more 
conservative, and will correct the inconsistency between the 
technical specification and its basis and as such, will not 
significantly increase the consequences of a previously evaluated 
accident.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated because:
    The proposed amendment changes the allowed containment leakage 
rate to 1.5%, changes the assumed value for SBGT system charcoal 
filter efficiency, and changes a specification in section of 4.6.C. 
``Coolant Chemistry'' from the value of .007 [microcurie]/ml to .002 
[microcurie]/ml. No new accident modes are created by clarifying the 
numerical value of the allowable containment leakage rate 
(La) or changing the assumed value for the SBGT system 
charcoal filter efficiency. No safety-related equipment or safety 
functions are altered, or adversely affected, as a result of these 
changes. The proposed changes will not introduce failure mechanisms 
beyond those already considered in the current plant safety 
analyses. Changing the allowable leakage rate, the assumed value for 
the efficiency of the SBGT system charcoal filter, and the 
specification in the bases section of 4.6.C. ``Coolant Chemistry'' 
does not contribute to the possibility of a new or different kind of 
accident or malfunction from those previously analyzed.

[[Page 19978]]

    (3) Involve a significant reduction in the margin of safety 
because:
    The proposed amendment changes the allowed containment leakage 
rate to 1.5%, changes the assumed value for SBGT system charcoal 
filter efficiency, and changes a specification in section of 4.6.C. 
``Coolant Chemistry'' from the value of .007 [microcurie]/ml to .002 
[microcurie]/ml. The design of the FitzPatrick plant is not changed. 
The methodology for test performance is unchanged and Type A, B and 
C tests will continue to be performed at [greater than or equal to] 
Pa. The value of La specified in proposed 
specification 6.20 is consistent with the accident analyses, 
therefore, the dose consequences of any analyzed accidents are not 
increased as a result of this change. The calculated doses as a 
result of the decrease in the assumed efficiency of the SBGT system 
charcoal filters for design basis accidents are marginally increased 
but still meet, and are well below, the dose acceptance criteria of 
10 CFR 100, the SRP, and GDC 19 of Appendix A to 10 CFR 50. The 
change in the specification in section 4.6.C. ``Coolant Chemistry'' 
from .007 [microcurie]/ml to .002 [microcurie]/ml is a minor 
editorial change, is more conservative, and will correct the 
inconsistency between the technical specification and its basis. 
Therefore, the proposed changes provide continued assurance of the 
leak tightness of the containment and conservatively assume SBGT 
system charcoal filter efficiency for the purpose of dose 
calculations for design basis accidents without adversely affecting 
the public health and safety and, as such, will not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

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

    Date of amendment request: March 30, 1998.
    Description of amendment request: The proposed changes would change 
the interval of selected Logic System Functional Tests (LSFT) from 
semiannually to once per 24 months. The definition of LSFT is also 
revised to be consistent with the Boiling Water Reactor Standard 
Technical Specifications.
    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 only significant change proposed by this application 
involves the extension of the surveillance test interval for the 
LSFTs required by the TS. The other changes involve editorial, 
format, and clarification changes, which by their nature are of no 
safety significance.
    Extending the LSFT interval from semiannually to once per 24 
months does not involve plant physical changes, change any TS 
setpoints, or introduce any new mode of plant operation. Therefore, 
the change does not degrade the performance of any safety system 
assumed to function in the accident analysis, and therefore, will 
not increase the consequences of an accident.
    Extending the LSFT interval from semiannually to 24 months 
results in no significant change in the logic system unavailability 
due to equipment failure. The reliability of safety systems subject 
to the LSFT are dominated by that of the mechanical components, and 
the logic system circuit relay coils which are subject to the more 
frequent functional test requirements. These factors are confirmed 
by the availability record of the affected safety system based on 
the past surveillance test history. Furthermore, the longer test 
intervals reduce the unavailability due to testing for the 
applicable safety system while the plant is operating. For these 
reasons, there is not a significant increase in the probability of 
an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not introduce any new accident 
initiators or failure mechanisms since the changes do not introduce 
any new modes of plant operation, make any physical changes, or 
change any TS setpoints. The changes reduce the probability of 
accidents initiated by test-induced plant transients by reducing the 
number of times the tests must be performed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety system settings, or limiting conditions for 
operation are determined. In several aspects, the proposed changes 
may actually enhance the margin of safety by reducing the potential 
for test-induced plant transients, reducing the unavailability due 
to test of the applicable safety system, and reducing any potential 
incremental logic system component wear. For these reasons, the 
changes do not involve a significant reduction in the margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mr. David E. Blabey, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: S. Singh Bajwa, Director.

Public Service Electric & Gas Company, Docket No. 50-272, Salem Nuclear 
Generating Station, Unit No. 1, Salem County, New Jersey

    Date of amendment request: March 26, 1998.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.1.3.3, ``Rod Drop Time,'' to change 
the applicability from Mode 3 (hot shutdown) to Modes 1 and 2 (startup 
and power operation).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to the Technical Specification Mode 
applicability provides consistency between the testing requirements 
as stated in the surveillance requirement of the Technical 
Specifications and intended by the initial conditions specified in 
the limiting condition for operations. The proposed change does not 
introduce any physical changes to the plant or equipment already in 
place in the plant, the proposed change ensures that testing of the 
rod drop times is performed in a manner that is consistent with the 
Technical Specifications and the assumptions made in the Salem 
accident analysis.
    Therefore, the proposed amendment does not increase the 
probability or consequences of any accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not introduce a new component or 
changes the manner in which the facility is operated, maintained or 
tested. Thus no new accident scenarios, failure mechanisms or 
limiting single failures are introduced as a result of the proposed 
change to the facility.
    Therefore the proposed amendment does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

[[Page 19979]]

    3. The proposed changes does not involve a significant reduction 
in a margin of safety.
    As stated in question number 2, the proposed change does not 
introduce a new component or changes the manner in which the 
facility is operated. Operation of the facility in accordance with 
the proposed amendment would not involve a significant reduction in 
the margin of safety. The Technical Specifications remain the same, 
as the input, or initial conditions, of the safety analysis have not 
changed. Therefore, there is no reduction in the margin to safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, NJ 08079.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Project Director: Robert A. Capra.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2, and 3, Limestone County, Alabama

    Date of amendment request: March 3, 1998.
    Description of amendment request: The proposed amendment would 
change the Browns Ferry Nuclear Plant Unit 2 and Unit 3 Technical 
Specification Figure 3.6-1 which contains the reactor vessel pressure-
temperature (PT) limits. The change would extend the validity of the 
curves to 32 effective full-power years (EFPY). The current PT curves 
are effective up to 12 EFPY. In addition to revised PT curves, several 
changes to the notes applicable to the curves are also proposed to be 
consistent with the supporting analysis.
    The proposed PT curves also would support a planned 5% power 
increase for each unit. Approval of the proposed power increase is 
pending and is the subject of a separate action before the Commission.
    The Tennessee Valley Authority has submitted the proposed change in 
current technical specification (CTS) format and in the improved 
standard technical specification (ISTS) format. Conversion to the ISTS 
format is pending.
    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.

    A. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 
CFR 50.92(c)(1)) because the proposed changes merely adjust the 
reference temperature for the limiting reactor vessel beltline 
material to account for accumulated and projected irradiation 
effects. The adjusted reference temperature analyses were performed 
in accordance with the requirements of Appendix G of 10 CFR part 50 
and the guidance contained in Regulatory Guide 1.99, Revision 2. The 
changes do not otherwise affect the manner by which the facility is 
operated and do not change any facility design feature or equipment. 
Since the protection previously provided will continue to be 
provided and there is no change to the facility or operating 
procedures, there is no effect upon the probability or consequences 
of any accident previously analyzed.
    B. The changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
(10 CFR 50.92(c)(2)) because no new failure modes are introduced. 
The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
(10 CFR 50.92(c)(2)) because they do not affect the function of any 
facility structure, system or component, or affect the manner by 
which the facility is operated.
    C. The changes do not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)) because the proposed changes 
assure that the reactor vessel PT limits will be valid for operation 
up to 32 EFPY and that the safety margins specified in Appendix G of 
10 CFR part 50 will be maintained.

    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: Athens Public Library, 405 E. 
South Street, Athens, Alabama 35611.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
ET 10H 400 West Summit Hill Drive, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: February 13, 1998 (TS 97-03).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specification (TS) by adding a new 
limiting condition for operation (LCO) that addresses requirements for 
the main feedwater isolation, regulating, and bypass valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    TVA has concluded that operation of SQN Units 1 and 2 in 
accordance with the proposed change to the TSs [or operating 
license(s)], does not involve a significant hazards consideration. 
TVA's conclusion is based on its evaluation, in accordance with 10 
CFR 50.91(a)(1), of the three standards set forth in 10 CFR 
50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    TVA will not change plant components, functions, or operating 
practices by implementing a change that adds a TS requirement for 
the main feedwater isolation, regulating, and bypass valves. TVA 
will maintain and verify operability of these valves through the 
proposed surveillance and actions to ensure the accident mitigation 
functions are available when applicable. These valves are not 
considered to be the source of an accident and the conservative 
addition of a requirement to maintain their safety function will not 
increase the probability of an accident. TVA will not increase the 
consequences of an accident by implementing this change because this 
addition ensures that the isolation of main feedwater is available 
to mitigate the consequences of an accident.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    TVA will not alter plant equipment or operating activities in 
the implementation of the proposed TS change. The valves used for 
the isolation of main feedwater are not a potential source for 
accidents and are designed for accident mitigation purposes. 
Therefore, TVA will not create the possibility of an accident of a 
different kind.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    TVA maintains and ensures the availability of the isolation 
function for the main feedwater system as assumed in the SQN 
accident analysis. TVA proposes this TS change to further assure 
this capability and to meet the requirements of 10 CFR 50.36. TVA 
will not change the methods of operating the plant or setpoints 
associated with safety-related equipment in the implementation of 
this request. Therefore, TVA will not reduce the margin of safety by 
implementing a TS LCO for the isolation functions of the main 
feedwater system.

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

[[Page 19980]]

    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah 
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendments: February 13, 1998 (TS 97-07).
    Brief description of amendments: The proposed amendments would 
change the Sequoyah (SQN) Technical Specification (TS) requirements for 
main steam isolation valves (MSIVs) to incorporate MSIV requirements 
consistent with the Westinghouse Standard TS (NUREG-1431) and would add 
testing requirements for the MSIVs that ensure the valves close on an 
actual or simulated automatic actuation signal.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), Tennessee Valley 
Authority (TVA), the licensee, has provided its analysis of the issue 
of no significant hazards consideration, which is presented below:

    TVA has concluded that operation of SQN Units 1 and 2, in 
accordance with the proposed change to the TSs, does not involve a 
significant hazards consideration. TVA's conclusion is based on its 
evaluation, in accordance with 10 CFR 50.91(a)(1), of the three 
standards set forth in 10 CFR 50.92(c).
    A. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes provide enhancements and clarifications of 
the requirements for inoperable MSIVs and periodic testing 
provisions. These changes do not alter the safety functions of the 
MSIVs or the operating practices that govern their application to 
plant conditions. The actions for Modes 2 and 3 are revised such 
that a longer time could occur before an inoperable MSIV is closed 
or the unit is placed in a mode that does not apply. However, this 
increase will not significantly impact the ability of the valves to 
mitigate an accident or affect the accident generation possibility. 
This is based on the low probability of an accident occurring that 
would require closure of the MSIVs and reasonable time intervals to 
transition to lower modes based on operating experience to reach the 
required modes in an orderly manner without challenging unit 
systems.
    The MSIVs provide accident mitigation functions but do not 
contribute to accident generation. The MSIV functions have not been 
altered by the proposed changes. Therefore, the proposed changes 
will not increase the probability of a previously evaluated 
accident. Based on the above discussions, the proposed changes will 
not significantly increase the consequences of an accident and in 
some instances they will enhance the safety functions.
    B. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The primary function of the MSIVs is to support accident 
mitigation and are not a significant contributor to events that 
could generate accidents. The main impact that could result from an 
inoperable MSIV is an inadvertent closure that results in a unit 
trip. This event is bounded by the accidents that are currently 
evaluated for SQN. Since the proposed change does not alter MSIV 
functions and the new surveillance will be performed in modes that 
will not challenge unit systems, the possibility of a new or 
different kind of accident is not created.
    C. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed changes clarify and enhance the current SQN 
requirements for the MSIVs with one exception. This exception is the 
completion time added to the Modes 2 and 3 action that could be a 
negative impact to the margin of safety. This change could allow the 
MSIV safety function to be inoperable for a longer period of time. 
The overall effect of the proposed changes considering the 
additional end-device testing, periodic verification of inoperable 
MSIV closure, and removal of the action to allow MSIV closure in 
Mode 1, is considered a positive impact to the margin of safety. 
Therefore, there is not a significant reduction in the margin of 
safety.

    The NRC has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1001 Broad Street, Chattanooga, Tennessee 37402.
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 10H, Knoxville, Tennessee 37902.
    NRC Project Director: Frederick J. Hebdon.

Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339, 
North Anna Power Station, Units No. 1 and No. 2, Louisa County, 
Virginia

    Date of amendment request: March 25, 1998.
    Brief description of amendments: The amendments revise the 
Technical Specifications (TS) Sections 6.1.1; 6.2.1.b; 6.5.1.1; 
6.5.1.6. a, d, h, and m; 6.5.1.7.c; 6.5.1.8; 6.14.1.2; 6.15.b; 6.2.3.5; 
6.5.1.2; 6.5.1.7.a for Unit 1 and 6.1.1; 6.2.1.b; 6.5.1.1; 6.5.1.6. a, 
d, h, and m; 6.5.1.7.c; 6.5.1.8; 6.13.b; 6.14.b; 6.2.3.5; 6.5.1.2; and 
6.5.1.7.a for Unit 2, changing the title of Station Manager to Site 
Vice President, and the titles of the Assistant Station Managers to 
Manager-Station Operation and Maintenance and Manager-Station Safety 
and Licensing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Virginia Electric and Power Company has reviewed the proposed 
Technical Specifications changes against the criteria of 10 CFR 
50.92 and has concluded that the changes do not pose a significant 
hazards consideration. Specifically, station operations in 
accordance with the proposed Technical Specifications changes will 
not:
    a. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are administrative in nature. The overall 
responsibility for safe operation and review of plant operations is 
not being changed. There are no changes to the operation of any 
plant system or its design as a result of these changes. Therefore, 
neither the probability of occurrence nor the consequences of an 
accident or malfunction of equipment important to safety previously 
evaluated in the safety analysis report are increased.
    b. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes are administrative in nature. The overall 
responsibility for safe operation and review of plant operations is 
not being changed. There are no changes to the operation of any 
plant system or its design that could create any new modes of 
operation or accident precursors. Therefore, it is concluded that no 
new or different kind of accident or malfunction from any previously 
evaluated has been created.
    c. The proposed changes do not result in a significant reduction 
in margin of safety as defined in the basis for any Technical 
Specifications.
    The proposed changes are administrative in nature. The overall 
responsibility for safe operation and review is not being changed. 
There are no changes to the operation of any plant system or its 
design as a result of these changes. Safety systems are maintained 
operable as required by Technical Specifications. Therefore, the 
margin of safety is not changed.

    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: The Alderman Library, Special

[[Page 19981]]

Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: P. T. Kuo, Acting Project Director.

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: November 5, 1997.
    Description of amendment request: The proposed Operating License 
change and changes to the technical specifications (TS) would permit 
the use of a temporary alternate supply line (jumper) to provide 
service water (SW) to the component cooling heat exchangers. The 
temporary jumper will permit maintenance to be performed on the 
existing supply line.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Virginia Electric and Power Company has reviewed the proposed 
changes against the criteria of 10 CFR 50.92 and has concluded that 
the changes do not pose a significant safety hazards consideration 
as defined therein. The proposed Operating License and Technical 
Specifications and Bases changes are necessary to allow the use of a 
temporary, seismic, non-missile protected jumper to provide service 
water (SW) to the Component Cooling Heat Exchangers (CCHXs) while 
maintenance work is performed on the existing SW supply line to the 
CCHXs. Since there is only one SW supply line to the CCHXs, an 
alternate SW supply must be provided whenever the line is removed 
from service. The temporary jumper provides this function.
    The use of the temporary jumper has been thoroughly evaluated, 
and appropriate constraints and compensatory measures (including a 
Contingency Action Plan) have been developed to ensure that the 
temporary jumper is reliable, safe, and suitable for its intended 
purpose. A complete and immediate loss of SW supply to the operating 
CCHXs is not considered credible, given the project constraints and 
the unlikely probability of a generated missile. Existing station 
abnormal procedures already address a loss of component cooling, and 
the use of alternate cooling for a loss of decay heat removal, in 
the unlikely event that they are required. Furthermore, appropriate 
mitigative measures have been identified to address potential 
flooding concerns. The minor administrative changes merely correct a 
table format inconsistency and update Basis section references.
    Consequently, the operation of Surry Power Station with the 
proposed amendment and license condition will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The SW and CC Systems will function as designed under the Unit 
operating constraints specified by this project (i.e., Unit 2 in 
operation and Unit 1 in a refueling outage), and the potential for a 
loss of component cooling is already addressed by Station Abnormal 
Procedures. Therefore, there is no increase in the probability of an 
accident previously evaluated. The possibility of flooding due to 
failure of the temporary SW supply jumper in the Turbine Building 
basement has been evaluated and dispositioned by the implementation 
of appropriate precautions and compensatory measures to preclude 
damage to the temporary jumper and to respond to a postulated 
flooding event. A flood watch will be present around-the-clock with 
authority and procedural guidance to isolate the jumper, if 
required. Furthermore, the CCHXs serve no design basis accident 
mitigating function. Therefore, the consequences of an accident 
previously evaluated are not increased.

    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The SW and CC Systems' design functions and basic configurations 
are not being altered as a result of using a temporary SW supply 
jumper. The temporary jumper is designed to be safety-related and 
seismic with all of the design attributes of the normal SW supply line, 
except for the automatic isolation function and complete missile 
protection. The design functions of the SW and CC systems are unchanged 
as a result of the proposed changes due to (1) required plant 
conditions, (2) compensatory measures, (3) a Contingency Action Plan 
for restoration of the normal SW supply if required, and (4) strict 
administrative control of the temporary SW valve to preclude flooding 
or to isolate non-essential SW within the design basis assumed time 
limits. Unit 1 will be in a plant condition which will provide adequate 
time to restore the normal SW supply, if required. Therefore, since the 
SW and CC systems will basically function as designed and will be 
operated in their basic configuration, the possibility of a new or 
different type of accident than previously evaluated in the UFSAR 
[Updated Final Safety Analysis Report] is not created.
    3. Involve a significant reduction in a margin of safety.
    The margin of safety as defined in the Technical Specifications is 
not reduced since an operable SW flowpath to the required number of 
CCHXs is provided, and Unit operating constraints, compensatory 
measures and contingencies will be implemented as required to ensure 
the integrity and the capability of the SW flowpath. The use of the 
temporary jumper will be limited to the time period when missile 
producing weather is not expected, and Unit 1 meets specified unit 
conditions. Therefore, the temporary SW jumper, under the imposed 
project constraints and compensatory measures, provides the same 
reliability as the normal SW supply line. Furthermore, the 
Probabilistic Safety Assessment for Surry Power Station has been 
reviewed relative to potential flooding when the temporary SW jumper is 
in use. It has been determined that due to the SW restoration project's 
compensatory and contingency measures, as well as the constraints 
imposed by the Maintenance Rule online risk matrix, the impact on core 
damage frequency due to flooding is negligible.
    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: 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: P. T. Kuo, Acting.

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

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

[[Page 19982]]

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

    Date of amendment request: March 6, 1998.
    Description of amendment request: The proposed amendment would (1) 
update the Technical Specification heatup and cooldown rate curves and 
extend their reactor fluence limit from the current 20 effective full 
power years (EFPY) to a new value of 35 EFPY, (2) incorporate into 
Technical Specifications the use of a Pressure and Temperature Limits 
Report (PTLR), and (3) change the power-operated relief valves (PORVs) 
temperature requirement for operability.
    Date of individual notice in the Federal Register: March 27, 1998 
(63 FR 14972).
    Expiration date of individual notice: April 27, 1998.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401.

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.

Commonwealth Edison Company, Docket No. 50-237, Dresden Nuclear Power 
Station, Unit 2, Grundy County, Illinois

    Date of application for amendment: March 19, 1998, as supplemented 
by letters dated March 28, 1998, and April 3, 1998.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TS) by revising the Dresden, Unit 2, Minimum Critical 
Power Ratio (MCPR) in TS Section 2.1.B and footnotes in TS Section 
5.3.A, to allow the use of Siemens Power Corporation ATRIUM-9B fuel for 
all operating Modes at Dresden, Unit 2, Cycle 16.
    Date of issuance: April 10, 1998.
    Effective date: Immediately, to be implemented within 30 days.
    Amendment No.: 168.
    Facility Operating License No. DPR-19: The amendment revised the 
TS. Public comments requested as to proposed no significant hazards 
consideration: Yes (63 FR 14735 dated March 26, 1998). 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 April 27, 1998, 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 March 28, 1998, and April 3, 1998, letters provided additional 
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 April 10, 1998.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    Local Public Document Room location: Morris Area Public Library 
District, 604 Liberty Street, Morris, Illinois 60450.

Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck 
Plant Middlesex County, Connecticut

    Date of application for amendment: Two applications, both dated May 
30, 1997.
    Brief description of amendment: Changes Administrative Controls 
Section of the Technical Specifications to implement new Certified Fuel 
Handler position and to implement revised management responsibilities 
and titles that reflect the permanently shut down status of the plant. 
In addition, minor typographical errors were corrected.
    Date of issuance: March 27, 1998.
    Effective date: Date of issuance, but to be implemented within 60 
days of issuance.
    Amendment No.: 192.
    Operating License No. DPR-61: Amendment revised the Technical 
Specifications.
    Date of initial notice in Federal Register: July 16, 1997 (62 FR 
38132 and 62 FR 38133). The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated March 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.

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

    Date of application for amendment: March 26, 1997.
    Brief description of amendment: The amendment revises the 
Containment Systems technical specifications (TS) to incorporate a note 
to allow opening an operable airlock door to perform repairs on 
inoperable airlock components when the other airlock door is 
inoperable. This amendment is in partial response to Consumers Energy's 
March 26, 1997, application. The Consumers Energy request also proposed 
revising the requirements contained in TS sections 3.6 and 4.5 to 
closely emulate the format and content of NUREG-1432, ``Standard 
Technical Specifications, Combustion Engineering Plants,'' (STS). That 
portion of the Consumers Energy request remains under staff review and 
will be addressed in a separate evaluation.
    Date of issuance: April 8, 1998.
    Effective date: April 8, 1998.
    Amendment No.: 179.
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.

[[Page 19983]]

    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66136).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendment: January 28, 1998 (NRC-98-0008), 
as supplemented on March 10, 1998 (NRC-98-0036).
    Brief description of amendment: The amendment revises the technical 
specifications (TSs) by modifying the ``#'' footnote to Table 1.2 and 
the ``*'' footnote to surveillance requirements 4.9.1.2 and 4.9.1.3 to 
permit the Reactor Mode Switch to be placed in the Run or Startup/Hot 
Standby positions to test switch interlock functions provided that all 
control rods are verified to remain fully inserted in core cells 
containing one or more fuel assemblies.
    Date of issuance: March 31, 1998.
    Effective date: March 31, 1998, with full implementation within 90 
days.
    Amendment No.: 116.
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9599) The March 10, 1998, supplement requested a change in the 
implementation period and was not outside the scope of the initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

Duke Energy Corporation (DEC), Docket Nos. 50-369 and 50-370, McGuire 
Nuclear Station, Units 1 and 2, Mecklenburg County, North Carolina

    Date of application for amendments: December 17, 1997.
    Brief description of amendments: The amendments revise Technical 
Specification Section 6.9.1.9 to reference updated or recently approved 
topical reports, which contain methodologies used to calculate cycle-
specific limits contained in the Core Operating Limits Report. For 
several reports DEC indicated staff approval, but neglected to provide 
an ``A'' designation for the report number. Upon agreement by DEC, the 
staff has made these appropriate editorial corrections. These topical 
reports have all been previously approved by the staff under licensing 
actions separate from the current amendment request.
    Date of issuance: April 8, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1--178; Unit 2--160.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4311).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 8, 1998.
    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, Charlotte, North Carolina.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: February 9, 1998.
    Brief description of amendment: The amendment approves the use of 
the repair roll technology (reroll) for the upper tubesheet region of 
the ANO-1 steam generators. The reroll technology is an alternative to 
the either sleeving or plugging steam generator tubes found during 
inservice inspections to have defects that exceed the stated repair 
criteria. The reroll methodology works by creating a new mechanical 
tube to tubesheet structural joint below the tube defect indication.
    Date of issuance: April 10, 1998.
    Effective date: April 10, 1998.
    Amendment No.: 190.
    Facility Operating License No. DPR-51: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 24, 1998 (63 
FR 9268).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 10, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit 
No. 1, Pope County, Arkansas

    Date of amendment request: April 1, 1998, as supplemented by letter 
dated April 8, 1998.
    Brief description of amendment: The amendment allows approximately 
440 steam generator tubes with confirmed volumetric indications within 
the upper tube sheet to remain in service during Cycle 15. The 
amendment revises TS 4.18.5.b to incorporate five criteria which need 
to be satisfied to allow steam generator tubes to remain in service 
during Cycle 15 with indications of outer diameter intergranular attack 
(ODIGA) in the upper tube sheet region of the steam generators.
    Date of issuance: April 10, 1998.
    Effective date: April 10, 1998.
    Amendment No.: 191.
    Facility Operating License No. DPR-51: The amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated April 
10, 1998.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801.

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

    Date of amendment request: November 18, 1996, as supplemented by 
letter dated January 21, 1998.
    Brief description of amendment: The amendment changes Technical 
Specification (TS) Surveillance Requirement 4.4.8.3.1.b to test the 
Shutdown Cooling System suction line relief valves in accordance with 
TS 4.0.5. Editorial changes to 4.4.8.3.1 and 4.4.8.3.1.a have also been 
made.
    Date of issuance: April 1, 1998.
    Effective date: April 1, 1998.
    Amendment No.: 140.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6985).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans

[[Page 19984]]

Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

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

    Date of amendment request: May 24, 1997, as supplemented by letter 
dated January 21, 1998.
    Brief description of amendment: The amendment modifies Technical 
Specifications (TS) 3.1.1.1, 3.1.1.2, 3.10.1 and Figure 3.1-1 by 
removing cycle dependent boron concentration and boration flow rate 
from the Action Statements and removing the ``RWSP at 1720 ppm'' curve 
from the figure. A change to TS Bases 3/4.1.1.1 and 3/4.1.1.2 has been 
included to support this change.
    Date of issuance: April 8, 1998.
    Effective date: April 8, 1998, to be implemented within 60 days.
    Amendment No.: 141.
    Facility Operating License No. NPF-38: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 18, 1997 (62 FR 
33123).
    The January 21, 1998, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Entergy Mississippi, Inc.

Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne 
County, Mississippi

    Date of application for amendment: October 28, 1997, as 
supplemented by letter dated January 9, 1998.
    Brief description of amendment: The amendment revised the Technical 
Specifications (TSs) to implement the containment leak rate testing 
provisions of 10 CFR Part 50, Appendix J, Option B.
    Date of issuance: April 6, 1998.
    Effective date: April 6, 1998.
    Amendment No: 135.
    Facility Operating License No. NPF-29: Amendment revises the TSs.
    Date of initial notice in Federal Register: December 3, 1997 ( 62 
FR 63976).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120.

IES Utilities Inc, Central Iowa Power Cooperative, and Corn Belt Power 
Cooperative, Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of application for amendment: February 3, 1998.
    Brief description of amendment: The amendment changes the 
operability requirement for the Standby Liquid Control system to Run/
Power Operations and Startup.
    Date of issuance: March 31, 1998.
    Effective date: March 31, 1998.
    Amendment No.: 221.
    Facility Operating License No. DPR-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 26, 1998 (63 
FR 9874).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

IES Utilities Inc, Central Iowa Power Cooperative, and Corn Belt Power 
Cooperative, Docket No. 50-331, Duane Arnold Energy Center, Linn 
County, Iowa

    Date of application for amendment: October 3, 1997 as supplemented 
on December 10, 1997.
    Brief description of amendment: The amendment revises the Operating 
License to allow the start of core offload as soon as 60 hours after 
shutdown.
    Date of issuance: April 2, 1998.
    Effective date: April 2, 1998.
    Amendment No.: 222.
    Facility Operating License No. DPR-49: Amendment revised the 
Operating License.
    Date of initial notice in Federal Register: January 28, 1998 (63 FR 
4314).
    The December 10, 1997 submittal provided clarifying information 
that did not change the initial no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, SE., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: October 20, 1997, as 
supplemented February 10, and March 10, 1998.
    Brief description of amendment: The amendment replaces in their 
entirety the existing Technical Specifications incorporated in Facility 
Operating License No. DPR-36 as Appendix A. Maine Yankee developed the 
revised Technical Specifications, titled Permanently Defueled Technical 
Specifications, to reflect the permanently shutdown and defueled status 
of the plant. Changes were made to the definitions, limiting conditions 
for operation, surveillance, and administrative control sections.
    Date of issuance: March 30, 1998.
    Effective date: March 30, 1998.
    Amendment No.: 161.
    Facility Operating License No. DPR-36: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63978). The February 10, and March 10, 1998, submittals added 
additional programs to the Section 5.5 Procedures and Section 5.6 
Programs and Manuals did not change the proposed no significant hazards 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Wiscasset Public Library, High 
Street, P.O. Box 367, Wiscasset, ME 04578.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point 
Nuclear Station, Unit 2, Oswego County, New York

    Date of application for amendment: October 31, 1997, as 
supplemented by letter dated February 3, 1998.
    Brief description of amendment: This amendment changes Technical 
Specifications to support design changes to upgrade the analog-based 
average power range monitor system with General Electric's Nuclear 
Measurement Analysis and Control Power Range Neutron Monitor System, 
including an Oscillation Power Range Monitor function.
    Date of issuance: March 31, 1998.
    Effective date: As of the date of issuance to be implemented upon

[[Page 19985]]

completion and acceptance of design modifications resulting from the 
installation of the Nuclear Measurement Analysis and Control Power 
Range Neutron Monitor System.
    Amendment No.: 80.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68310).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

North Atlantic Energy Service Corporation, Docket No. 50-443, Seabrook 
Station, Unit No. 1, Rockingham County, New Hampshire

    Date of amendment request: September 26, 1997, as supplemented by 
letter dated March 12, 1998.
    Description of amendment request: The amendment revises Technical 
Specification 3.7.6, ``Control Room Emergency Makeup Air and 
Filtration,'' and its associated Bases to separate the requirements for 
the control room air conditioning subsystem from the requirements for 
control room makeup air and filtration subsystem based on system 
function. The amendment also increases the allowed outage time for the 
Control Room Air Conditioning Subsystem.
    Date of issuance: April 9, 1998.
    Effective date: As of the date of issuance, with full 
implementation within 60 days.
    Amendment No.: 56.
    Facility Operating License No. NPF-86: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54874). The March 12, 1998, supplemental letter 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 April 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of application for amendment: December 1, 1997.
    Brief description of amendment: The amendment changes the Technical 
Specifications (TSs) by adding a 2.0 second plus or minus 0.1 second 
time delay to the 4.16 kV Emergency Bus Undervoltage Loss of Power, 
Level One, trip setpoint and allowable values in TS Table 3.3-4.
    Date of issuance: April 1, 1998.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 214.
    Facility Operating License No. DPR-65: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14,1998 (63 FR 
2280).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 1, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

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

    Date of application for amendment: November 11, 1997.
    Brief description of amendment: The amendment allows NNECO to 
credit soluble boron for maintaining k-effective at less than or equal 
to 0.95 within the spent fuel pool rack matrix following a seismic 
event of a magnitude greater than or equal to an operating basis 
earthquake.
    Date of issuance: April 9, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 158.
    Facility Operating License No. NPF-49: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 3, 1997 (62 FR 
63980).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 9, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, Connecticut, and the Waterford Library, ATTN: Vince Juliano, 
49 Rope Ferry Road, Waterford, Connecticut.

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay Power 
Plant, Unit 3, Humboldt County, California

    Date of application for amendment: December 9, 1996, as 
supplemented on June 12, 1997, and March 13, 1998.
    Brief description of amendment: This amendment revised the 
Technical Specification to incorporate the requirements of appendix I 
of 10 CFR Part 50, into the Radiological Effluent Technical 
Specification (RETS) and to relocate the controls and limitations on 
RETS and radiological environmental monitoring (Currently in the 
Technical Specifications) to the Offsite Dose Calculation Manual and 
the Process Control Program. The amendment also revised the Technical 
Specifications to implement Generic Letter 89-10 (GL 89-10) and to 
incorporate the requirements of the revised 10 CFR Part 20.
    Date of issuance: April 8, 1998.
    Effective date: As of the date of issuance and shall be implemented 
no later than 30 days from the date of issuance.
    Amendment No.: 32.
    Facility Operating License No. DPR-7: Amendment revised the TS.
    Date of initial notice in Federal Register: April 24, 1996 (61 FR 
18174).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Humboldt County Library, 131 
3rd Street, Eureka, California 95501.

Pennsylvania Power and Light Company, Docket No. 50-387, Susquehanna 
Steam Electric Station, Unit 1, Luzerne County, Pennsylvania

    Date of application for amendment: August 26, 1997, as supplemented 
by letters dated December 4, 1997, February 2, March 19, and April 2, 
1998.
    Brief description of amendment: This amendment changes the 
Susquehanna Unit 1 Technical Specifications to support the use of the 
Siemens Power Corporation ATRIUM-10 fuel design in the upcoming Cycle 
11 refueling outage.
    Date of issuance: April 6, 1998.
    Effective date: As of date of issuance, to be implemented within 30 
days.

[[Page 19986]]

    Amendment No.: 174.
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 31, 1997 (62 
FR 68314).
    The December 4, 1997, February 2, March 19, and April 2, 1998, 
submittals provided clarifying information that did not change the 
initial no significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated April 6, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, PA 18701.

Philadelphia Electric Company, Docket No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania

    Date of application for amendment: September 2, 1997.
    Brief description of amendment: These amendments revise LGS, Units 
1 and 2, TS Section 4.0.5, and Bases Sections B 4.0.5 and B 3/4.4.8 
regarding the surveillance requirements associated with Inservice 
Testing and Inservice Inspection Programs of the American Society of 
Mechanical Engineers Code Class 1, 2, and 3 components.
    Date of issuance: March 31, 1998.
    Effective date: March 31, 1998.
    Amendment Nos.: 125 and 89.
    Facility Operating License No. NPF-39: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 11, 1998 (63 
FR 6990).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, PA 19464.

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: July 23, 1997, as supplemented by 
letters dated September 30, October 27, and December 18, 1997, and 
February 12, 1998.
    Brief Description of amendments: The amendments revise the 
Pressure-Temperature Limit Heatup, Cooldown, and Hydrostatic Testing 
curves for Farley Units 1 and 2 and relocate the curves from the 
Technical Specifications to a Pressure and Temperature Limits Report 
for each unit.
    Date of issuance: April 9, 1998.
    Effective date: As of the date of issuance to be implemented for 
Unit 1 prior to entering Mode 4 for Cycle 16 refueling outage (fall 
1998); for Unit 2 prior to entering Mode 4 for Cycle 13 refueling 
outage (spring 1998).
    Amendment Nos.: Unit 1-136; Unit 2-128.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: September 10, 1997 (62 
FR 47699); January 14, 1998 (63 FR 2281); February 23, 1998 (63 FR 
9020).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated April 9, 1998.
    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.

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

    Date of amendment request: October 24, 1997 (TXX-97228).
    Brief description of amendments: The amendments revise core safety 
limit curves and Overtemperature N-16 reactor trip setpoints based on 
analyses of the core configuration and expected operation for CPSES 
Unit 1, Cycle 7. The changes apply equally to CPSES Units 1 and 2 
licenses since the Technical Specifications are combined.
    Date of issuance: March 27, 1998.
    Effective date: March 27, 1998, to be implemented within 30 days.
    Amendment Nos.: Unit 1--Amendment No. 57; Unit 2--Amendment No. 43.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications/operating licenses.
    Date of initial notice in Federal Register: November 19, 1997 (62 
FR 61847).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 27, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: August 8, 1997, as supplemented 
by letter dated January 16, 1998.
    Brief description of amendment: The amendment revises the Callaway 
Plant, Unit 1 Technical Specification Table 3.3-3 Functional Units 
4.b.2 and 5.a.2 to make the number of main steam and feedwater 
isolation system (MSFIS) channels consistent with the solid state 
protection system, adds a clarifying note and changes Table 4.3-2 
Functional Units 4.b.2 and 5.a.2 slave relay quarterly test to a 
monthly staggered actuation logic test.
    Date of issuance: March 25, 1998.
    Effective date: March 25, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 123.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 17, 1997 (62 
FR 66143) The January 16, 1998, supplemental letter provided additional 
clarifying information that did not change the staff's original no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 25, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, Missouri

    Date of application for amendment: October 17, 1997, as 
supplemented by letters dated March 3, 1998, and March 17, 1998.
    Brief description of amendment: The amendment revises the technical 
specifications to modify the heatup and cooldown curves and the maximum 
allowable power operated relief valve setpoint curves for cold 
overpressure protection.
    Date of issuance: April 2, 1998.
    Effective date: April 2, 1998, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 124.
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 14, 1998 (63 FR 
2282).
    The March 3, 1998, and March 17, 1998, supplemental letters 
provided

[[Page 19987]]

additional clarifying information and did not change the initial no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated April 2, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Missouri-
Columbia, Elmer Ellis Library, Columbia, Missouri 65201-5149.

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

    Date of application for amendment: October 11, 1996.
    Brief description of amendment: The amendment revises the Technical 
Specifications regarding the amount of foam concentrate required to 
support operability of the reactor recirculation motor generator set 
foam fire suppression system.
    Date of Issuance: March 31, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 156.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 22, 1997 (62 FR 
54877).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated March 31, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: June 9, 1997.
    Brief description of amendment: The amendment revises Technical 
Specification Section 6.0 to add and revise reference to NRC-approved 
methodologies which will be used to validate or generate the cycle-
specific thermal hydraulic stability based operating limits in the 
Vermont Yankee Core Operating Limits Report.
    Date of Issuance: April 7, 1998.
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 157.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 13, 1997 (62 FR 
43377).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 7, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

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

    Date of application for amendment: August 22, 1996.
    Brief description of amendment: The amendment revises the Technical 
Specifications to change the action statement for the high range stack 
noble gas monitor based on the guidance of Generic Letter 83-36, NUREG-
0737 Technical Specifications.
    Date of Issuance: April 8, 1998.
    Effective date: April 8, 1998, to be implemented within 30 days.
    Amendment No.: 158.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 4, 1997 (62 FR 
30647).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated April 8, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, VT 05301.

Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek 
Generating Station, Coffey County, Kansas

    Date of amendment request: January 28, 1998.
    Brief description of amendment: The amendment revises Technical 
Specification Secs. 6.3 and 6.12 to reflect a merger for the positions 
of Superintendent Radiation Protection and Superintendent Chemistry 
into one new position, Manager Chemistry/Radiation Protection.
    Date of issuance: March 30, 1998.
    Effective date: March 30, 1998.
    Amendment No.: 115.
    Facility Operating License No. NPF-42: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 25, 1998 (63 
FR 9614).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 30, 1998.
    No significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621.

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

[[Page 19988]]

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 22, 1998, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings and 
Adjudications Staff or may be delivered to the Commission's Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC, 
by the above date. A copy of the petition should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and to the attorney for the licensee. 
Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

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

    Date of amendment request: April 2, 1998 (NRC-98-0062).

[[Page 19989]]

    Description of amendment request: The amendment revised the action 
specified in Technical Specification Table 3.3.7.5-1 if one channel of 
drywell oxygen monitoring is inoperable.
    Date of issuance: April 3, 1998.
    Effective date: April 3, 1998, with full implementation by April 6, 
1998.
    Amendment No.: 117.
    Facility Operating License No. NPF-43: Amendment revises the 
License and the Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No. The Commission's related evaluation of the 
amendment, finding of emergency circumstances, and final determination 
of no significant hazards consideration are contained in a Safety 
Evaluation dated April 3, 1998.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: Cynthia A. Carpenter.

    Dated at Rockville, Maryland, this 15th day of April 1998.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 98-10470 Filed 4-21-98; 8:45 am]
BILLING CODE 7590-01-P