[Federal Register Volume 60, Number 119 (Wednesday, June 21, 1995)]
[Notices]
[Pages 32359-32381]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-50621]



[[Page 32359]]

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 May 26, 1995, through June 9, 1995. The last 
biweekly notice was published on Tuesday, June 6, 1995 (60 FR 29869).

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 Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, 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 July 21, 1995, 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 [[Page 32360]] 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, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, 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 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of amendment request: June 8, 1995, supersedes December 16, 
1994, request in its entirety, supplemented by letters dated November 
30, 1994, April 27, 1995, May 5 and May 11, 1995.
    Description of amendment request: The proposed amendment would 
revise Figure 3.4-4a in the Braidwood Unit 1's technical specifications 
which provides the nominal pressurizer power operated relief valve set 
points for the low-temperature overpressure protection system (LTOPS). 
The proposed revision would extend the applicability of Figure 3.4-4a 
from 5.37 effective full power years (EFPY) to 16 EFPY (Unit 1). In 
addition, the proposed amendment removes the 638 psig administrative 
limit line from the LTOPS curve, because the appropriate instrument 
uncertainties and discharge piping pressure limits are included in the 
proposed LTOPS curve. The amendment request also proposes 
administrative changes to Figure 3.4-4a format and its associated index 
page.
    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 new LTOPS curve will not change any postulated accident 
scenarios. The revised curve was developed using industry standards 
and regulations which are recognized as being inherently 
conservative. Appropriate instrument uncertainties and allowances 
have been included in the development of the LTOPS curves. The PT 
and LTOPS curves provide RCS pressure limits to protect the Reactor 
Pressure Vessel (RPV) from brittle fracture by clearly separating 
the region of normal operations from the region where the RPV is 
subject to brittle fracture.
    Using Regulatory Guide (RG) 1.99, ``Radiation Embrittlement of 
Reactor Vessel Materials,'' Revision 2, Braidwood Unit 1 
Surveillance Capsule U and Capsule X results and the requirements of 
Appendix G to 10 CFR 50, as modified by the guidance in ASME Code 
Case N-514, a new LTOPS curve was prepared. This new curve, in 
conjunction with the PT Limit curves, and the heatup and cooldown 
ranges provides the required assurance that the RPV is protected 
from brittle fracture.
    No changes to the design of the facility have been made, no new 
equipment has been installed, and no existing equipment has been 
removed or modified. This amendment will not change any system 
operating modes. The revised LTOPS curve provides assurance that the 
RPV is protected from brittle fracture.
    The index page and format changes are purely administrative in 
nature and are designed to reflect the change in the duration of 
applicability of Figure 3.4-4a and improve the readability of Figure 
3.4-4a. These administrative changes will have no effect on any 
equipment, system, or operating mode.
    Thus, 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.
    The use of the new LTOPS curve does not change any postulated 
accident scenarios. The new LTOPS curve was generated using 
Braidwood capsule surveillance data and an approved, conservative 
methodology. No new equipment will be installed, and no existing 
equipment will be modified. No new system interfaces are created, 
and no existing system interfaces are modified. The new LTOPS curve 
provides assurance that the RPV is protected from brittle fracture.
    No new accident or malfunction mechanism is introduced by this 
amendment.
    The index page and format changes are purely administrative in 
nature and are designed to reflect the change in the duration of 
applicability of Figure 3.4-4a, and improve the readability of 
Figure 3.4-4a. These administrative changes will have no effect on 
any equipment, system, or operating mode.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The new LTOPS curve was developed using industry standards and 
regulations which are recognized as being inherently conservative. 
Appropriate instrument uncertainties and allowances are included in 
the development of the new LTOPS curve. This amendment will not 
change the operational characteristics or design of any equipment or 
system.
    All accident analysis assumptions and conditions will continue 
to be met. The RPV is adequately protected from non-ductile failure 
by the revised LTOPS curve.
    The index page and format changes are purely administrative in 
nature and are designed to reflect the change in the duration of 
applicability of Figure 3.4-4a, and improve the readability of 
Figure 3.4-4a. These administrative changes will have no effect on 
any equipment, system, or operating mode.
    Thus, 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 [[Page 32361]] 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: Wilmington Public Library, 201 
S. Kankakee Street, Wilmington, Illinois 60481
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: March 4, 1993, as revised April 14, 
1993, as supplemented April 19 and May 31, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) to conform to the wording of 
the revised 10 CFR Part 20, ``Standards for Protection Against 
Radiation,'' and to reflect a separation of chemistry and radiation 
protection responsibilities. The supplemental submittals provided 
additional information on the proposed TS change in response to NRC's 
request for additional information of May 5, 1995. The original 
submittal was noticed on May 12, 1993 (58 FR 28053), as corrected June 
1, 1993 (58 FR 31222).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1.Will the proposed change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change does not affect the probability or 
consequences of an accident. The proposed change is to the 
ADMINISTRATIVE and RADIOLOGICAL EFFLUENT RELEASES sections of the 
facility Technical Specifications, and are administrative in nature.
    - Change ``Chemistry and Radiation Protection Supervisor'' to 
``Radiation Protection Supervisor.''
    - The change from ``mR/h'' to ``mrem/h'' is solely a change in 
terminology since the revised 10 CFR 20 does not recognize or define 
the roentgen as a unit of radiation.
    - The Liquid Effluents Concentration section and the associated 
bases have been revised to conform with 10 CFR 50.36(a) [10 CFR 
50.36a] with effluent concentrations limited to 10 times the limits 
of 10 CFR 20.1001 - 20.2402, Appendix B, Table 2, Column 2.
    - The actual instantaneous dose rate limits of the Gaseous 
Effluents Dose Rate section have not changed. However, the bases 
section has. Under the former 10 CFR 20, these dose rates correspond 
roughly to maximum permissible concentration and dose(s) received by 
the maximum exposed member of the public if allowed to continue for 
an entire year. These limits are used more as instantaneous limits 
(dose rates above which are not allowed to continue for more than 
one hour at a time) so as to provide assurance not to exceed 10 CFR 
50, Appendix I limits.
    2. Will the proposed change(s) create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    This proposed change is required by the implementation of a new 
10 CFR Part 20 requirements (except for the title change) and are 
administrative in nature (sic). Neither the material condition of 
the facility nor the accident analyses are affected by this proposed 
change. Therefore, the proposed change does not create the 
possibility of a different type of accident than previously 
evaluated.
    3. Will the proposed change involve a significant reduction in 
the margin of safety?
    Each limit that was affected increased the margin of safety by 
making the limit more conservative; or remained the same.
    - The change of distance to ``30 centimeters'' (12 inches) is 
more conservative, providing a higher degree of protection for 
occupationally exposed worker.
    - The liquid effluent concentration limits remain essentially 
the same. The bases have changed to [10 CFR 50.36a] reflect 10 times 
10 CFR 20.1001 - 20.2402, Appendix B, Table 2, Column 2 limits as 
controlled by 10 CFR 50.36(a) [10 CFR 50.36a] dose limits.
    - Effluent alarm setpoints were reviewed to determine any 
necessary changes and were found to be set appropriately. No change 
will be necessary.
    - ``The instantaneous release rate limits for airborne releases 
will not be changed because they are imposed on licensees as a 
control to ensure that the licensees meet Appendix I requirements.'' 
Alarm setpoints for these dose rate limits may change slightly due 
to changes in scientific data and will be reviewed and changed as 
appropriate prior to implementation.
    Therefore, the proposed change does not involve a reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location:  North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: Cynthia A. Carpenter, Acting

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

    Date of amendment request: April 12, 1995
    Description of amendment request: The amendments delete Technical 
Specification 3/4.3.4, ``Turbine Overspeed Protection,'' and its 
associated Bases. The deletion of TS 3/4.3.4 and its associated Bases 
provides Duke Power Company the flexibility to implement the 
manufacturer's recommendations for turbine steam valve surveillance 
test requirements. These test requirements will be relocated from the 
TS to the Selected Licensee Commitments (SLC) Manual. The SLC Manual is 
Chapter 16 of the Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Relocation of the affected TS section to the SLC Manual 
will have no effect on the probability of any accident occurring. In 
addition, the consequences of an accident will not be impacted since 
the above system will continue to be utilized in the same manner as 
before. No impact on the plant response to accidents will be 
created.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No new accident causal mechanisms will be created as a 
result of relocating the affected TS requirements to the SLC Manual. 
Plant operation will not be affected by the proposed amendments and 
no new failure modes will be created.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. No impact upon any plant safety 
margins will be created. Relocation of the affected TS requirements 
to the SLC Manual in consistent with the content of the Westinghouse 
RSTS [Revised Standard Technical Specifications], as the NRC did not 
require technical specification controls for the turbine overspeed 
protection system in the RSTS. The proposed amendments are 
consistent with the NRC philosophy of encouraging utilities to 
propose amendments that are consistent with the content of the RSTS.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this [[Page 32362]] 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: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 18, 1995, as supplemented by letter 
dated May 31, 1995.
    Description of amendment request: The proposed amendment would 
change Tecnical Specification (TS) 3.6.1.2 to defer the next scheduled 
containment integrated leak rate test (ILRT) at Catawba, Unit 2, for 
one outage, from the end-of-cycle (EOC) 7 refueling outage (scheduled 
for October 1995) to EOC-8 (scheduled for March 1997). Title 10 of the 
Code of Federal Regulations, Part 50, Appendix J, requires that three 
ILRTs be performed at approximately equal intervals during each 10-year 
service period at a nuclear station. ``Approximately equal intervals'' 
is defined in Catawba's TS as 40 plus or minus 10 months. The proposed 
one-time change would allow Catawba to extend that interval to less 
than or equal to 70 months.
    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 will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Containment leak rate testing is not an initiator of any 
accident; the proposed interval extension does not affect reactor 
operations or accident analysis, and has no perceptible radiological 
consequences. Therefore, this proposed change will not involve a 
significant increase in the probability or consequences of any 
previously[]evaluated accident.
    2. The proposed change will not create the possibility of any 
new accident not previously evaluated.
    The proposed change does not affect normal plant operations or 
configuration, nor does it affect leak rate test methods. The test 
history at Catawba (no ILRT [intergrated leak rate test] failures) 
provides continued assurance of the leak tightness of the 
containment structure.
    3. There is no significant reduction in a margin of safety.
    It has been documented in draft NUREG-1493 that an increase in 
the ILRT interval from 1 test every 3 years to 1 test every 10 years 
would result in an increase in population exposure risk in the 
vicinity of 5 representative plants from .02% to .14%. The proposed 
change included herein, an increase from 40 [plus or minus] 10 
months to [less than or equal to] 70 months, represents a small 
fraction of that already very small increase in risk. Therefore, it 
may be concluded that no significant reduction in a margin of safety 
will occur.
    Based on the above, no significant hazards consideration is 
created by the proposed change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: April 12, 1995
    Description of amendment request: The amendments delete Technical 
Specification 3/4.3.4, ``Turbine Overspeed Protection,'' and its 
associated Bases. The deletion of TS 3/4.3.4 and its associated Bases 
provides Duke Power Company the flexibility to implement the 
manufacturer's recommendations for turbine steam valve surveillance 
test requirements. These test requirements will be relocated from the 
TS to the Selected Licensee Commitments (SLC) Manual. The SLC Manual is 
Chapter 16 of the Updated Final Safety Analysis Report.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. Relocation of the affected TS section to the SLC Manual 
will have no effect on the probability of any accident occurring. In 
addition, the consequences of an accident will not be impacted since 
the above system will continue to be utilized in the same manner as 
before. No impact on the plant response to accidents will be 
created.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No new accident causal mechanisms will be created as a 
result of relocating the affected TS requirements to the SLC Manual. 
Plant operation will not be affected by the proposed amendments and 
no new failure modes will be created.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. No impact upon any plant safety 
margins will be created. Relocation of the affected TS requirements 
to the SLC Manual in consistent with the content of the Westinghouse 
RSTS [Revised Standard Technical Specifications], as the NRC did not 
require technical specification controls for the turbine overspeed 
protection system in the RSTS. The proposed amendments are 
consistent with the NRC philosophy of encouraging utilities to 
propose amendments that are consistent with the content of the RSTS.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Florida Power and Light Company, et al., Docket No. 50-335, St. 
Lucie Plant, Unit No. 1, St. Lucie County, Florida

    Date of amendment request: May 17, 1995
    Description of amendment request: The amendment will extend the 
applicability of the current Reactor Coolant System (RCS) Pressure/
Temperature Limits and maximum allowed RCS heatup and cooldown rates to 
23.6 Effective Full Power Years (EFPY) of operation. In addition, 
administrative changes are proposed for [[Page 32363]] TS 3.1.2.1 
(Boration Systems Flow Paths-Shutdown) and TS 3.1.2.3 (Charging Pump-
Shutdown) to clarify the conditions for which a High Pressure Safety 
Injection pump may be used.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if operation of the facility in accordance with the 
proposed amendment would not: (1) involve a significant increase in 
the probability or consequences of an accident previously evaluated; 
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. Each standard is discussed as 
follows:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The pressure-temperature (P/T) limit curves in the Technical 
Specifications are conservatively generated in accordance with the 
fracture toughness requirements of 10 CFR 50 Appendix G as 
supplemented by the ASME Code Section XI, Appendix G 
recommendations. The RT<INF>NDT values are based on Regulatory Guide 
1.99, Revision 2, shift prediction and attenuation formula. Analyses 
of reactor vessel material irradiation surveillance specimens are 
used to verify the validity of the fluence predictions and the P/T 
limit curves. Use of these curves in conjunction with the 
surveillance specimen program ensures that the reactor coolant 
pressure boundary will behave in a non-brittle manner and that the 
possibility of rapidly propagating fracture is minimized. Based on 
the use of plant specific material data, analysis has demonstrated 
that the current P/T limit curves will remain conservative for up to 
23.6 EFPY.
    In conjunction with extending the applicability of the existing 
P/T limit curves, the low temperature overpressure protection (LTOP) 
analysis for 15 EFPY is also extended. The LTOP analysis confirms 
that the current setpoints for the power-operated relief valves 
(PORVs) will provide the appropriate overpressure protection at low 
Reactor Coolant System (RCS) temperatures. Because the P/T limit 
curves have not changed, the existing LTOP values have not changed, 
which include the PORV setpoints, heatup and cooldown rates, and 
disabling of non-essential components.
    The proposed amendment does not change the configuration or 
operation of the plant, and assurance is provided that reactor 
vessel integrity will be maintained. Therefore, operation of the 
facility in accordance with the proposed amendment will not involve 
a significant increase in the probability 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 accident previously evaluated.
    By applying plant specific data in the determination of critical 
vessel material limits, the applicability of the existing pressure 
temperature limits and LTOP requirements can be extended. There is 
no change in the configuration or operation of the facility as a 
result of the proposed amendment. The amendment does not involve the 
addition of new equipment or the modification of existing equipment, 
nor does it alter the design of St. Lucie plant systems. Therefore, 
operation of the facility in accordance with the proposed amendment 
would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety.
    Analysis has demonstrated that the fracture toughness 
requirements of 10 CFR 50 Appendix G are satisfied and that 
conservative operating restrictions are maintained for the purpose 
of low temperature overpressure protection. The P/T limit curves 
will provide assurance that the RCS pressure boundary will behave in 
a ductile manner and that the probability of a rapidly propagating 
fracture is minimized. Therefore, operation of the facility in 
accordance with the proposed amendment would not involve a 
significant reduction in a margin of safety.
    Based on the discussion presented above and on the supporting 
Evaluation of Proposed TS Changes, FPL has concluded that this 
proposed license amendment involves no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: May 17, 1995
    Description of amendment request: The proposed amendments will 
improve consistency between the Technical Specifications and the 
improved Combustion Engineering Standard Technical Specifications 
(NUREG-1432, dated September 1992) by incorporating changes in text and 
resolving other inconsistencies identified by the NRC and plant 
operations staff.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Pursuant to 10 CFR 50.92, a determination may be made that a 
proposed license amendment involves no significant hazards 
consideration if operation of the facility in accordance with the 
proposed amendment would not: (1) involve a significant increase in 
the probability or consequences of an accident previously evaluated; 
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. Each standard is discussed as 
follows:
    (1) Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments consist of administrative changes to the 
Technical Specifications (TS) for St. Lucie Units 1 and 2. The 
amendments will implement changes in text to improve consistency 
within the TS for each unit, the improved Combustion Engineering 
Standard Technical Specifications (NUREG-1432, dated September 
1992), and the regulations. The proposed amendments do not involve 
changes to the configuration or method of operation of plant 
equipment that is used to mitigate the consequences of an accident, 
nor do the changes otherwise affect the initial conditions or 
conservatism assumed in any of the plant accident analyses. 
Therefore, operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed administrative revisions will not change the 
physical plant or the modes of plant operation defined in the 
Facility License for each unit. The changes do not involve the 
addition or modification of equipment nor do they alter the design 
or operation of plant systems. Therefore, operation of the facility 
in accordance with the proposed amendments would not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (3) Operation of the facility in accordance with the proposed 
amendment would not [[Page 32364]] involve a significant reduction 
in a margin of safety.
    The proposed amendments are administrative in nature and do not 
change the basis for any technical specification that is related to 
the establishment of, or the preservation of, a nuclear safety 
margin. Therefore, operation of the facility in accordance with the 
proposed amendments would not involve a significant reduction in a 
margin of safety.
    Based on the above discussion and the supporting Evaluation of 
Technical Specification changes, FPL has determined that the 
proposed license amendment involves no significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003
    Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: May 23, 1995
    Description of amendment request: The licensee proposes to change 
Turkey Point Units 3 and 4 Technical Specifications (TS) by changing 
the setpoint presentation format for the Reactor Protection System 
(RPS) and Engineered Safety Features Actuation System (ESFAS) 
instrumentation setpoints contained in Technical Specification Tables 
2.2-1 and 3.3-3. The approved Westinghouse five-column instrument 
setpoint methodology currently being used to establishing those 
setpoints would be retained. The intent of the amendments is to 
eliminate the need for minor administrative license amendments to these 
tables that do not impact either the Trip Setpoints or the Safety 
Analysis Limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    No changes to the Reactor Trip System instrumentation setpoints, 
ESFAS instrumentation setpoints, or the Turkey Point Plant licensing 
basis (NRC-approved, Westinghouse five-column setpoint methodology, 
as documented in Westinghouse topical report WCAP-12745P), is being 
made. The changes proposed reduce the level of detail in the 
Technical Specifications and place that detailed information in 
controlled procedures, drawings and the Final Safety Analysis 
Report. Since the setpoints and methodology remain the same, the 
changes proposed by this submittal will not increase the probability 
or consequences of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    These proposed changes remove from the Technical Specifications 
a level of detail which will be maintained in controlled procedures 
and drawings. The Turkey Point Plant licensing basis (NRC-approved, 
Westinghouse five column setpoint methodology, as documented in 
Westinghouse topical report WCAP-12745P), continues to be used to 
calculate the Reactor Trip System and ESFAS setpoints. No changes to 
Reactor Trip System or ESFAS instrumentation setpoints are proposed. 
Since the same methodology will be used to determine the setpoints 
and no setpoints are changed, the possibility that a new or 
different kind of accident from any previously evaluated will not be 
created.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The Turkey Point Plant licensing basis (NRC-approved, 
Westinghouse five column setpoint methodology, as documented in 
Westinghouse topical report WCAP-12745P), continues to be used to 
calculate the Reactor Trip System and ESFAS setpoints. No changes to 
the Reactor Trip System or ESFAS instrumentation setpoints are 
proposed. Since the same methodology will be used to determine the 
setpoints, and no setpoints are changed by this submittal, this 
change does not involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied.Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: J.R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: David B. Matthews

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

    Date of amendment request: June 6, 1995
    Description of amendment request: The proposed change would revise 
Plant Hatch Units 1 and 2 Technical Specification (TS) Surveillance 
Requirements (SR) 3.6.4.1.3 and 3.6.4.1.4 for the secondary containment 
drawdown. The revision would reduce the SR acceptance criteria to 
greater than or equal to 0.20 inch of vacuum from greater than or equal 
to 0.25 inch of vacuum. Also, the licensee proposed to change the Bases 
to reflect the proposed TS revision.
    The licensee stated that the secondary containment performs no 
active function in response to either loss-of-coolant accident or fuel 
handling accident. However, its leak tightness is required to ensure 
that the release of radioactive materials from the primary containment 
is restricted to those leakage paths and associated leakage rates 
assumed in the accident analysis and that fission products entrapped 
within the secondary containment structure will be treated by the Unit 
1 and Unit 2 standby gas treatment systems prior to discharge to the 
environment. This change will continue to provide adequate margin for 
the secondary containment to be sufficiently leak tight such that the 
conclusions of the accident analysis remain valid.
    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 change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. The 
secondary containment serves a mitigation function and therefore 
this change does not increase the probability of an accident 
previously evaluated. The consequences of the previously evaluated 
accidents are not affected because at the wind conditions assumed in 
the accident analysis the building will be at a negative pressure 
and no exfiltration is postulated. Furthermore, the estimated wind 
speed at which exfiltration might take place (31 mph) is not a 
frequent occurrence (wind speeds of greater than 24 mph occur [less 
than] <0.5% of the time based on Plant Hatch specific meteorological 
data).
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
Revising the surveillance [[Page 32365]] requirement acceptance 
criteria does not physically modify the plant nor does it modify the 
operation of any existing equipment.
    3. The proposed change does not involve a significant reduction 
in the margin of safety. The change in vacuum acceptance criteria 
results in a slightly lower wind speed that may result in 
exfiltration from the building. However, this wind speed (31 mph) is 
in the realm of wind speeds which are infrequent at Plant Hatch. 
Furthermore, there are numerous conservatisms in the existing dose 
calculations including: neutral to stable meteorological conditions, 
ground level release until establishment of the required vacuum, 
accident source terms at event initiation, and no credit for 
plateout. The secondary containment would be maintained at a slight 
negative pressure shortly after the Standby Gas Treatment fans are 
running and the releases would be from the main stack (well before 
the accident source term would be present in the secondary 
containment). Some plateout would also occur and this is 
conservatively ignored. Therefore the margin of safety is not 
significantly reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 17, 1995
    Description of amendment request: The proposed license amendment 
would revise Section 3.2 of the Technical Specifications (TSs) for 
Three Mile Island Nuclear Station, Unit 1 (TMI-1) to relocate the 
requirements for volume and boron concentration of the chemical 
addition system boric acid mix tank and the reclaimed boric acid 
storage tank from the TMI-1 TSs to the TMI-1 Core Operating Limits 
Report. The licensee, in its request, stated that the proposed changes 
are consistent with the intent of NRC Generic Letter 88-16.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The proposed amendment relocates chemical addition tank 
volume and boron concentration parameters from Technical 
Specifications to the TMI-1 Core Operating Limits Report. The 
proposed amendment provides continued control of the values of these 
parameters and assures these values are developed using NRC-approved 
reload methodologies consistent with all applicable limits of the 
safety analyses addressed in the TMI-1 [Final Safety Analysis 
Report] FSAR. The Technical Specifications retain the requirement to 
maintain the plant within the appropriate bounds of these limits. 
Therefore, the proposed amendment has no effect on 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 accident previously evaluated. The 
proposed amendment relocates chemical addition tank volume and boron 
concentration parameters to the TMI-1 Core Operating Limits Report. 
The Technical Specifications retain the requirement to maintain the 
boric acid mix tank and reclaimed boric acid storage tank volume and 
boron concentration parameters within the appropriate limits. 
Therefore, the proposed amendment has no effect on the possibility 
of creating a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment provides continued control of the 
boric acid mix tank and reclaimed boric acid storage tank volume and 
boron concentration parameters and assures these values remain 
consistent with all applicable limits of the safety analyses 
addressed in the TMI-1 FSAR. Therefore, it is concluded that 
operation of the facility in accordance with the proposed amendment 
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: 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: Phillip F. McKee

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

    Date of amendment request: May 24, 1995
    Description of amendment request: The proposed license amendment 
would revise Table 4.1-1 of the Technical Specifications (TSs) for 
Three Mile Island Nuclear Station, Unit 1 (TMI-1) to revise the test 
frequency requirement for the source range nuclear instrumentation from 
7 days before reactor startup to 6 months before startup.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the facility in accordance with the proposed 
TSCR would not involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated.
    The proposed revision to the Technical Specifications does not 
involve any physical changes to the plant, and it does not impact 
the safety analysis with respect to design basis events and 
assumptions. The only change proposed is in the ``Test'' frequency 
for source-range Nuclear Instrumentation by revision of the 
appropriate Tech. Spec. tables. The revised testing requirement has 
no impact upon the probability of occurrence or the consequences of 
any accident previously evaluated, because no credit is taken in the 
accident analyses for the source range monitors nor are there any 
inputs to the Reactor Protection System. Tech. Spec. 3.1.9.2 
requires that the control rod withdraw inhibit be operable at all 
times; however, it is not affected by this change request. 
Additionally, no nuclear safety equipment or systems interface with 
source-range nuclear instrumentation, and operator ability to 
monitor and trend post-accident neutron level is not affected by the 
proposed change. Therefore, this change request will not increase 
the probability of occurrence or the consequences of any previously 
analyzed accidents as described in the Updated [Final Safety 
Analysis Report] FSAR (UFSAR).
    2. Operation of the facility in accordance with the proposed 
TSCR would not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed revision to the TMI-1 Technical Specifications does 
not involve any physical changes to the plant, and does not impact 
on the safety analysis with respect to design basis events and 
assumptions. The only change proposed is in the ``Test'' frequency 
for Nuclear Instrumentation by revision of the appropriate Tech. 
Spec. tables. No nuclear safety equipment or [[Page 32366]] systems 
interface with the source-range nuclear instrumentation, and 
operator ability to monitor and trend post-accident neutron levels 
is not adversely affected by the proposed change. In addition, the 
source-range nuclear instrument channels provide indication to the 
control room, plant computer and one of two channels provides input 
to Remote Shutdown Panel B.
    The 0.5% instrument drift over a six (6) month period will not 
affect the ability to operate other safety equipment; nor, will it 
increase the probability of failure of the rod withdrawal inhibit. 
The inhibit function is triggered by a startup rate, and a 0.5% 
drift over six (6) months will not affect the instrument's ability 
to perform the inhibit function. Therefore, this change has no 
impact upon the possibility of creating a new or different kind of 
accident from any previously evaluated in the UFSAR.
    3. Operation of the facility in accordance with the proposed 
TSCR would not involve a significant reduction in a margin of 
safety.
    The proposed revision to the TMI-1 Technical Specifications does 
not involve any physical changes to the plant, and does not impact 
on the safety analysis with respect to design basis events and 
assumptions. The only change proposed is in the surveillance 
frequency for Nuclear Instrumentation by revision of the appropriate 
Tech. Spec. tables. Startup rate instrumentation is not included in 
Technical Specifications 2.0, ``Safety Limits''; and, hence, all 
system Limiting Conditions for Operation(s) remain unchanged. 
Testing of the source-range nuclear instrument channels within six 
(6) months prior to a reactor startup will not decrease the margin 
of safety. Hence, the margin of safety for the plant is not 
diminished by this change request.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Phillip F. McKee

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

    Date of amendment request: June 1, 1995
    Description of amendment request: The proposed license amendment 
would revise Section 5.3.1.1 of the Technical Specifications (TSs) for 
Three Mile Island Nuclear Station, Unit 1 (TMI-1) to allow use of an 
alternate zirconium-based cladding material manufactured by Babcock & 
Wilcox Fuel Company to test the properties of the fuel in an operating 
core. Present TSs require fuel clad material to be either ``zircaloy'' 
or ``ZIRLO.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or the consequences of an accident 
previously evaluated. The test assemblies with the zirconium-based 
claddings are mechanically and thermal-hydraulically similar to the 
remainder of the reload batch and the rest of the core, so no 
failure probability is increased, nor is any operational practice 
changed which could introduce a new initiator of an accident. The 
only credible event which could occur as a result of this 
demonstration is clad failure of the test fuel rods. The number of 
fuel rods involved is such a small percentage of the core inventory 
that even a postulated failure of all the demonstration fuel rods 
from a cause related to the demonstration would not result in dose 
consequences greater than existing limits. A failure of the fuel 
rods from a cause not related to the demonstration would not result 
in consequences greater than those which would have occurred had the 
assemblies not been demonstrated assemblies. Therefore, this change 
does not increase the probability of occurrence or the 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 accident previously evaluated. The 
mechanical and thermal-hydraulic similarity of the test assemblies 
to the remainder of assemblies in the core precludes the credible 
possibility of creating any new failure mode or accident sequence. 
The use of the demonstration assemblies does not involve any 
alterations to plant equipment or procedures which would introduce 
any new or unique operational modes or accident precursors.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The demonstration assemblies meet the same design as the 
remainder of assemblies in the core. Existing reload design and 
safety analysis limits are maintained, and the FSAR analyses are 
bounding. No special setpoints or other safety settings are required 
as a result of the use of these two (2) test assemblies. The 
assemblies will be placed in locations which will not experience 
limiting peak power conditions. Therefore, it is concluded that 
operation of the facility in accordance with the proposed amendment 
does not involve a reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: 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: Phillip F. McKee

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

    Date of amendment request: April 27, 1995, as supplemented by 
letters dated May 4, and May 25, 1995.
    Description of amendment request: The proposed amendment would 
change the tables associated with Technical Specification (TS) 3/
4.3.3.5, Remote Shutdown System, to eliminate the core exit 
thermocouples (CETs). The proposed amendment would also change the 
tables associated with TS 3/4.3.3.6, Accident Monitoring 
Instrumentation, to require two operable channels of CETs, where each 
channel would be required to have at least two operable CETs per core 
quadrant. Each channel would also be required to have at least four 
operable CETs in at least one quadrant to support the operability of 
the subcooling margin monitors. In addition, the actions related to TS 
3/4.3.3.6 would be changed to require that a report be submitted if one 
CET channel in a quadrant is inoperable for more than 30 days, and 
require a plant shutdown if both CET channels in a quadrant are 
inoperable for more than 7 days.
    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 consequence of an accident previously evaluated?
    Change to Technical Specification 3.3.3.5: [[Page 32367]] 
    Deleting the reference to the core exit thermocouples from the 
Remote Shutdown Technical Specification will not involve a 
significant increase in the probability of an accident previously 
evaluated because the core exit thermocouples are not potential 
accident initiators. The consequences of an accident previously 
evaluated will not be increased because the core exit thermocouples 
availability is not reduced, since adequate assurance of their 
operability is provided in Technical Specification 3.3.3.6, and by 
the surveillance of other indications that require the availability 
of the displays that also provide the core exit temperatures at the 
Auxiliary Shutdown Panel.
    Change to Technical Specification 3.3.3.6:
    The proposed change reduces the number of core exit 
thermocouples required per quadrant per channel from at least 4 to 
at least 2. Thus, the Actions when less than 4 thermocouples per 
quadrant per train are Operable but more than 6 thermocouples per 
quadrant are OPERABLE, and less than 6 thermocouples per quadrant 
are OPERABLE but at least 4 thermocouples per quadrant are OPERABLE 
and with the number of OPERABLE channels less than 4 thermocouples 
per quadrant are being deleted. This change does not affect the 
probability of an accident. The Accident Monitoring Instruments are 
not initiators of any analyzed events. The consequence of an 
accident is not affected by this change. The requirement to have two 
core exit thermocouples OPERABLE per quadrant per channel is 
adequate because one OPERABLE core exit thermocouple must be located 
near the center of the core and the other OPERABLE core exit 
thermocouple must be located near the core perimeter, such that the 
pair of core exit thermocouples indicate the radial temperature 
gradient across their core quadrant. The change will not alter 
assumptions relative to the mitigation of an accident or transient 
event. Functions supported by the thermocouples will still be 
adequately supported by the system. The revised specification 
provides for at least one quadrant per channel to have at least four 
operable thermocouples to protect the subcooling margin monitor in 
the event of a single failure. The other indications used to assess 
core cooling, as described in Chapter 7B of the South Texas Project 
Updated Final Safety Analysis Report remain unaffected by the 
proposed change. Therefore, this change will not involve a 
significant increase in the probability or consequence of an 
accident previously evaluated.
    The proposed change also affects the allowed outage times for 
the thermocouples. The existing specification allows for 31 days in 
the case where there are less than four thermocouples per quadrant 
per train operable, 7 days where there are less than 6 thermocouples 
per quadrant, and 48 hours where there are less than 4 thermocouples 
per quadrant. The required action for each of these cases is a plant 
shutdown. The proposed specification will require a report to the 
Commission after 30 days in the case where one channel of core exit 
thermocouples is inoperable, and it will require the plant to go to 
HOT SHUTDOWN if two channels are inoperable for more than 7 days. A 
plant shutdown with only one channel inoperable is not warranted 
based on the fact that the redundant channel remains available to 
provide the necessary indication and the passive nature of the 
instrumentation (i.e., no critical automatic action).
    As noted above, the core exit thermocouples are not accident 
initiators; consequently, the change in allowed outage time does not 
affect the probability of an accident. The consequences of an 
accident are not significantly increased because the changes to the 
allowed outage times are not extended to allow operation of the 
system in such a degraded condition that it will not perform its 
function. In addition, the other indications used to assess core 
cooling, as described in Chapter 7B of the South Texas Project 
Updated Final Safety Analysis Report remain unaffected by the 
proposed change. As noted above, functionality of the core exit 
temperature indication is preserved by requiring at least two 
thermocouples to be operable in separate regions of the core 
quadrant.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Change to Technical Specification 3.3.3.5:
    Deleting the core exit thermocouples from the Remote Shutdown 
Technical Specification will not create the possibility of a new or 
different accident because there are no automatic actuations 
performed by the core exit thermocouples, nor are any different 
plant configurations or different operational procedures proposed. 
The existing safety analyses are unchanged and still applicable.
    Change to Technical Specification 3.3.3.6:
    The proposed change reduces the number of core exit 
thermocouples required per quadrant per channel from at least 4 to 
at least 2. Thus, the Actions when less than 4 thermocouples per 
quadrant per train are Operable but more than 6 thermocouples per 
quadrant are OPERABLE, and less than 6 thermocouples per quadrant 
are OPERABLE but at least 4 thermocouples per quadrant are OPERABLE 
and with the number of OPERABLE channels less than 4 thermocouples 
per quadrant are being deleted. This change will not physically 
alter the plant (no new or different type of equipment will be 
installed). The changes in methods governing normal plant operation 
are consistent with current safety analysis assumptions. Therefore, 
the change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The change in the allowed outage time does not alter the 
physical configuration of the plant or how the plant is operated; 
consequently, this change does not create the possibility of a new 
or different kind of accident.
    3. Does this change involve a significant reduction in a margin 
of safety?
    Change to Technical Specification 3.3.3.5:
    Deleting the core exit thermocouples from the Remote Shutdown 
Technical Specification does not involve a significant reduction in 
the margin of safety because the core exit thermocouples indications 
will still be available at the Auxiliary Shutdown Panel. In 
addition, adequate and appropriate assurance of the operability of 
the core exit thermocouples is provided in Technical Specification 
3.3.3.6 for Accident Monitoring Instrumentation, including the 
changes proposed in this letter.
    Change to Technical Specification 3.3.3.6:
    The proposed change reduces the number of core exit 
thermocouples required per quadrant per channel from at least 4 to 
at least 2. Thus, the Actions when less than 4 thermocouples per 
quadrant per train are Operable but more than 6 thermocouples per 
quadrant are OPERABLE, and less than 6 thermocouples per quadrant 
are OPERABLE but at least 4 thermocouples per quadrant are OPERABLE 
and with the number of OPERABLE channels less than 4 thermocouples 
per quadrant are being deleted. The margin of safety is not affected 
by this change. The Accident Monitoring Instrumentation provide no 
automatic actuation functions. Even though the number of core exit 
thermocouples per quadrant per channel is being reduced, the Bases 
requirement to have one core exit thermocouple located near the 
center of the core and one core exit thermocouple located near the 
core perimeter ensures that the pair of core exit thermocouples 
indicate the radial temperature gradient across their core quadrant 
which ensures the required level of information is available. The 
functions dependent on the core exit thermocouples are still 
adequately supported by the thermocouples. The revised specification 
provides for at least one quadrant per channel to have at least four 
operable thermocouples to protect the subcooling margin monitor in 
the event of a single failure. In addition, the other indications 
used to assess core cooling, as described in Chapter 7B of the South 
Texas Project Updated Final Safety Analysis Report remain unaffected 
by the proposed change. The safety analysis assumptions will still 
be maintained, thus, no question of safety exists. Therefore, the 
change does not involve a significant reduction in a margin of 
safety.
    The proposed changes to the allowed outage times have no 
significant impact on the margin of safety. A plant shutdown with 
only one channel inoperable is not warranted based on the fact that 
the redundant channel remains available to provide the necessary 
indication and the passive nature of the instrumentation (i.e., no 
critical automatic action). Based on the small likelihood of an 
accident occurring concurrent with the station being in an ACTION 
statement with regard to the thermocouples, and the small chance 
that the degradation of the system in such a situation would affect 
its functionality, and the diversity provided by other indications 
of core cooling, the changes in the allowed outage times are not 
considered significant.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior 
[[Page 32368]] College, J. M. Hodges, Learning Center, 911 Boling 
Highway, Wharton, Texas 77488
    Attorney for licensee: Jack R. Newman, Esq., Newman & Holtzinger, 
P.C., 1615 L Street, N.W., Washington, D.C. 20036
    NRC Project Director: William D. Beckner

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: March 31, 1995
    Description of amendment requests: The proposed amendments would 
modify the technical specifications to eliminate the requirement to 
test certain safeguards pumps via their recirculation flowpath. The 
affected pumps are the centrifugal charging pumps, residual heat 
removal pumps, motor driven auxiliary feedwater pumps, and the turbine 
driven auxiliary feedwater pumps. The proposed amendments would also 
eliminate references to specific discharge pressures and flows 
associated with these pumps and remove footnotes associated with the 
Unit 2 cycle 9-10 refueling outage which are no longer applicable.
    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:
    Per 10 CFR 50.92, a proposed change does not involve a 
significant hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.
    Criterion 1
    The purpose for conducting periodic testing of the pumps 
identified in this proposed amendment is to detect gross degradation 
as required by Section XI of the ASME [American Society of 
Mechanical Engineers] Code. The Cook Nuclear Plant IST [Inservice 
Testing] program, which encompasses Section XI of the ASME Code, is 
the basis for the existing as well as the proposed T/Ss. Testing the 
pumps utilizing a high capacity flowpath instead of a recirculation 
flow path (where applicable) will have no impact on the ability of 
the pump to perform its intended function. In fact, it is expected 
that the high capacity flowpath will provide a more accurate 
assessment of the pump/systems' conditions and ability to meet their 
safety function.
    The removal of specific test parameters, in favor of referencing 
the Cook Nuclear Plant IST Program, will not impact the ability of 
the pumps to perform their safety related function. IST Program 
parameters ensure that the pumps under test provide the support 
assumed in the plant's safety analyses.
    Therefore, based on these considerations, it is concluded that 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2
    The proposed change will preclude the need to realign selected 
pumps to their recirculation flowpaths for testing purposes (where 
applicable). Eliminating the need for alignment to the recirculation 
flowpath aids in maximizing the pump's availability to perform its 
safety function.
    As stated previously the removal of the specific test 
parameters, in favor of referencing the Cook Nuclear Plant IST 
Program will not impact the ability of the pumps to perform their 
intended safety function.
    Thus, it is concluded that the proposed changes do not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    Criterion 3
    As stated previously, testing of the selected pumps utilizing a 
high capacity flowpath will provide greater assurance of pump 
capability and maximize pump availability. Additionally, removing 
specific test parameters in favor of referencing the Cook Nuclear 
Plant IST Program will have no impact on the ability of the pumps to 
perform their intended safety function. Therefore, we believe that 
the margin for safety as defined int 10 CFR [Part] 100 has not been 
reduced. Based on these considerations, it is concluded that the 
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. Although not specifically addressed in the licensee's 
analysis, the elimination of specific discharge pressures and flows is 
encompassed in the elimination of the recirculation testing requirement 
and presents no additional significant hazards consideration. 
Therefore, the NRC staff proposes to determine that the amendment 
requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Cynthia A. Carpenter, Acting

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, 
Donald C. Cook Nuclear Plant, Unit Nos. 1 and 2, Berrien County, 
Michigan

    Date of amendment requests: May 19, 1995
    Description of amendment requests: The proposed amendments would 
modify the Technical Specification action statement associated with the 
Main Steam Safety Valves (MSSVs). The action statement would reflect 
different requirements based on operating Mode and the power range 
neutron flux high setpoint with inoperable MSSVs would be revised in 
response to an issue raised in Westinghouse Nuclear Safety Advisory 
Letter 94-001.
    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:
    Per 10 CFR 50.92, a proposed change does not involve a 
significant hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.
    Criterion 1
    Correction of the setpoint methodology does not represent a 
credible accident initiator. The new methodology reduces the 
allowable power level setpoints and is conservative compared to the 
presently evaluated setpoints. The consequences of any previously 
evaluated accident are not adversely affected by this action because 
the decrease in the setpoints resulting from the new calculational 
methodology will ensure that the MSSVs are capable of relieving the 
pressure at the allowable power levels. Based on these 
considerations, it is concluded that the changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Correcting the overly restrictive action statements of T/S 3.7.1 
does not involve a significant increase in the probability of an 
accident. The proposed changes modify existing text to more 
accurately reflect the intention of the restrictions imposed by the 
action statements. The changes do not create any situation that 
would initiate a credible accident sequence.
    Criterion 2
    The change in Table 3.7-1 reduces the allowable power levels 
that can be achieved in the event that one or more main steam safety 
valve(s) is inoperable. This change is a result of vendor guidance 
to correct an error in the existing methodology used to determine 
the setpoints for the power level. Changing the methodology used to 
determine the setpoints, and lowering the setpoints themselves, do 
no create a new condition [[Page 32369]] that could lead to a 
credible accident. Therefore, it is concluded that the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The action statements remain in effect to perform the intended 
function of protecting the plant's secondary side when the main 
steam safety valves are inoperable. They have only been modified to 
correct the overly restrictive language that specifies when, in each 
MODE, specific actions must be taken. Therefore, the proposed change 
does not create a new or different type of accident.
    Criterion 3
    The margin of safety presently provided is not reduced by the 
proposed change in the setpoints. The change will correct the 
limiting power levels that are to be implemented when MSSVs are 
inoperable. This action does not adversely affect the margin that 
was previously allocated for the ability of the MSSVs to relieve 
secondary side pressure. Based on these considerations, it is 
concluded that the changes do not involve a significant reduction in 
a margin of safety.
    The margin of safety is also not significantly reduced by the 
proposed change to the action statements of the T/S. The proposed 
revision clarifies when specific actions are to be taken in response 
to inoperable main steam safety valves. The changes do not decrease 
the effectiveness of the actions to be taken; therefore, they do not 
significantly reduce 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 requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Cynthia A. Carpenter, Acting

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

    Date of amendment request: April 16, 1995
    Description of amendment request: The proposed amendment would 
modify certain requirements of the Seabrook Station Technical 
Specifications relating to containment building penetrations during 
refueling operations. One change would allow both doors of the 
containment personnel airlock (PAL) to be open during core alterations 
or movement of irradiated fuel within containment provided at least one 
PAL door is capable of being closed and a designated individual is 
available outside the PAL to close the door. Another change would allow 
the use of alternate containment building penetration closure 
methodologies during refueling operations and provide for the manual 
closure of a penetration provided a designated individual is available 
at the penetration.
    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)). The changes do not affect the events or conditions 
which could result in a fuel handling accident and do not affect any 
equipment or procedures used for fuel handling. The changes would 
continue to ensure that penetrations which provide direct access of 
the containment atmosphere to outside containment are capable of 
restricting a release of radioactive material to the environment. 
Therefore, the changes do not involve a significant increase in the 
probability of an accident previously evaluated.
    The changes do have the potential for increased dose at the site 
boundary due to a postulated fuel handling accident. However, the 
licensee's radiological evaluations show that the resulting offsite 
and control room doses would be well within the acceptance limits of 
10 CFR Part 100 and within the acceptance limits of GDC 19.
    The Commission has provided guidance concerning the application 
of standards in 10 CFR 50.92 by providing certain examples (cf. 
FEDERAL REGISTER, March 6, 1986 51 FR 7751) of amendments that are 
considered not likely to involve a significant hazards 
consideration. These changes are similar to example (vi) in the 
Federal Register notice, in that they result in an increase in the 
consequences of a previously analyzed accident, but the results of 
the change are clearly within all acceptance criteria.
    B. The changes do not create the possibility of a new or 
different kindof accident from any accident previously evaluated (10 
CFR 50.92(c)(2))because the changes do not affect the events or 
conditions which could result in a fuel handling accident and do not 
affect any equipment or procedures used for fuel handling. The 
changes do not make any modifications to existing plant structures, 
systems, or components, or otherwise affect the manner by which the 
facility is operated.
    C. The changes do not involve a significant reduction in a 
margin ofsafety (10 CFR 50.92(c)(3)) because the increase in 
calculated offsite and control room doses resulting from a 
postulated fuel handling accident are within the acceptance limits 
of 10 CFR Part 100 and within the acceptance limits of GDC 19. 
Additionally, the changes do not otherwise affect the manner by 
which the facility is operated or involve modifications to equipment 
or features which affect the operational characteristics of the 
facility.
    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: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston MA 02110-2624.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
Millstone Nuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of amendment request: May 18, 1995
    Description of amendment request: The proposed amendment revises 
the minimum temperature at which the reactor vessel head bolting studs 
are allowed to be placed under tension. In addition, the proposed 
amendment revises the minimum reactor vessel metal temperature during 
core critical operation, revises the minimum reactor vessel metal 
temperature for pressure tests, makes editorial changes, and revises 
the bases for the applicable section.
    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 changes against the criteria set 
forth in 10CFR50.92 and has concluded that the changes do not 
involve a significant hazards consideration (SHC). The bases for 
this conclusion are that the three criteria of 10CFR50.92(c), 
discussed separately below, are not compromised. The proposed 
changes do not involve a SHC because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    Revising the boltup temperature of the reactor vessel head, from 
86F to 70F, does not decrease the margins of safety, as required 
by 10CFR 50 Appendix G, against non-ductile failure of the reactor 
vessel. Therefore, the probability of occurrence of an accident 
previously evaluated in the safety analysis report (i.e., a 
LOCA)[loss of coolant accident] is not increased since the revised 
boltup temperature does not increase the probability of failure of 
the vessel head flange region. The reactor vessel is a passive 
[[Page 32370]] component which does not initiate or play a role in 
any previously evaluated accidents or in mitigating the consequences 
of any previously evaluated accidents. Therefore, the proposed 
changes do not involve a significant increase in the probability or 
the consequences of any accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated:
    Revising the boltup temperature of the reactor vessel head, from 
86 deg.F to 70 deg.F, does not decrease the margins of safety, as 
required by 10CFR 50 Appendix G, against non-ductile failure of the 
reactor vessel. Therefore, the possibility for a new or different 
kind of accident than previously evaluated (i.e., a LOCA through the 
vessel flange) is not created.
    3. Involve a significant reduction in a margin of safety.
    Using the proposed boltup temperature of 70 deg.F still provides 
a self-imposed ``margin'' over the most limiting vessel flange 
region RT<INF>NDT of 22 deg.F (i.e., 70 deg. - 48 deg. = 22 deg.). 
This is a ``margin'' over and above the boltup temperature required 
by Appendix G to the 1992 ASME Section XI Code, since Appendix G 
would allow a boltup temperature of 48F.
    The above proposed changes to the Limiting Condition for 
Operation for tensioning the reactor vessel head studs do not alter 
the configuration, normal operation, design bases, function, 
mission, or performance of the subject components. Therefore, the 
proposed changes do not affect the margin of safety inherent in the 
design, analysis, function, or operation of the reactor vessel head 
flange region. The proposed changes do not alter the fuel clad 
barrier, fuel integrity, reactor vessel integrity, reactor coolant 
system integrity, or the containment boundary integrity; thus the 
margin of safety related to these barriers remains unchanged.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
Millstone Nuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of amendment request: May 24, 1995
    Description of amendment request: The proposed amendment would 
permit an individual who does not have a current senior reactor 
operator (SRO) license to hold the Operations Manager position. The 
position will require the individual to have previously held an SRO 
license at a boiling water reactor (BWR). An individual serving in the 
capacity of the Assistant Operations Manager will hold a current SRO 
license for Millstone Unit 1, if the Operations Manager does not. In 
addition, the proposed amendment would renumber the applicable sections 
of the related 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:
    NNECO has reviewed the proposed change in accordance with 
10CFR50.92 and concluded that the change does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed change does not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed change affects an administrative control, which was 
based on the guidance of ANSI N18.1-1971. ANSI N18.1-1971 
recommended that the Operations Manager hold an SRO license. The 
current guidance in Section 4.2.2 of ANSI/ANS 3.1-1987 recommends, 
as one option, that the Operations Manager have held a license for a 
similar unit and the Operations Middle Manager hold an SRO license. 
While the Operations Middle Manager position does not exist at 
Millstone Unit No. 1, NNECO has created the position of Assistant 
Operations Manager. The individual in this position would meet the 
requirements for, and would have responsibilities as recommended in, 
ANSI/ANS 3.1-1987 for the Operations Middle Manager position.
    Therefore, the proposed change requests an exception to ANSI 
N18.1-1971 to allow use of ANSI/ANS 3.1-1987 in a limited 
circumstance. Specifically, the proposed revision to Technical 
Specification 6.3.1 would require the Operations Manager to either 
hold an SRO license at Millstone Unit No. 1 or have held an SRO at a 
BWR.
    If the Operations Manager does not hold an SRO license at 
Millstone Unit No. 1, the specification will require the Assistant 
Operations Manager to hold, and continue to hold, an SRO license. 
The proposed change includes the requirement to have held a license 
for a similar unit (a BWR) in accordance with Section 4.2.2 of ANSI/
ANS 3.1-1987, if the Operations Manager does not hold an SRO license 
at Millstone Unit No. 1. For those areas of knowledge that require 
an SRO license, the Assistant Operations Manager will provide the 
technical guidance typically provided by the Operations Manager.
    The proposed change does not alter the design of any system, 
structure, or component, nor does it change the way plant systems 
are operated. It does not reduce the knowledge, qualifications, or 
skills of licensed operators, and does not affect the way the 
Operations Department is managed by the Operations Manager. The 
Operations Manager will continue to maintain the effective 
performance of his personnel and ensure the plant is operated safely 
and in accordance with the requirements of the operating license. 
Additionally, the control room operators will continue to be 
supervised by the licensed Shift Supervisor.
    The proposed change does not detract from the Operations 
Manager's ability to perform his primary responsibilities. In this 
case, by having previously held an SRO license, the Operations 
Manager has achieved the necessary training, skills, and experience 
to fully understand the operation of plant equipment and the watch 
requirements for operators. In summary, the proposed change does not 
affect the ability of the Operations Manager to provide the plant 
oversight required of that position. Thus, it does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed change to Technical Specification 6.3.1 does not 
affect the design or function of any plant system, structure, or 
component, nor does it change the way plant systems are operated. It 
does not affect the performance of licensed operators. Operation of 
the plant in conformance with technical specifications and other 
license requirements will continue to be supervised by personnel who 
hold an SRO license. The proposed change to Technical Specification 
6.3.1 ensures that the Operation Manager will be a knowledgeable and 
qualified individual by requiring the individual to have held an SRO 
license at a BWR. Based on the above, the proposed change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed change involves an administrative control that is 
not related to the margin of safety. The proposed change does not 
reduce the level of knowledge or experience required of an 
individual who fills the Operations Manager position, nor does it 
affect the conservative manner in which the plant is operated. The 
Control Room operators will continue to be supervised by personnel 
who hold an SRO license. Thus, 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 
[[Page 32371]] amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

Northeast Nuclear Energy Company (NNECO), Docket No. 50-245, 
Millstone Nuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of amendment request: May 26, 1995
    Description of amendment request: The proposed amendment will 
delete the old limiting conditions for operation (LCOs) and 
surveillance requirements and add new LCOs, surveillance requirements, 
and bases for the loss of normal power (LNP) instrumentation system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    NNECO has reviewed this proposed change in accordance with 
10CFR50.92 and concluded that this change does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed change does not involve an SHC because the 
changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The change does not increase the probability of a loss of off-
site power event or the occurrence of any accidents which assume 
loss of off-site power. This is ensured by the LNP instrumentation 
system design which uses multiple sensing relays, redundancy, and 
qualified Class 1E components, as well as conservative operability 
and surveillance requirements.
    Full LNP logic requires two sets of relays to trip in one of two 
redundant groups. One set monitors bus 14E and the other set 
monitors bus 14F. Separate sets are provided for loss of voltage and 
degraded voltage monitoring. This design minimizes the likelihood of 
an inadvertent full LNP initiation. To maintain redundancy in the 
instrumentation, two separate groups are provided, each group being 
powered from an independent DC supply. Partial LNP logic is also 
provided to detect a loss of voltage on a single emergency bus. 
Redundancy in the partial LNP logic is achieved by providing an 
independent logic for each emergency power train.
    The proposed technical specification would require that the LNP 
instrumentation be maintained operable except when the unit is in 
cold shutdown or refueling conditions. If redundancy in the ability 
to detect a loss of voltage or degraded voltage and initiate a full 
LNP is not maintained, reactor operation would be permitted for 
seven days. In this situation, both full and partial LNP (and both 
emergency power sources) remain operable. An action statement of 
seven days, which is the same as the action statement duration for 
an inoperable EDG [emergency diesel generator], is justified based 
on continued operability of the other LNP group. Additionally, it 
allows a reasonable amount of time to perform repairs.
    The time delays and voltage setpoints specified in Table 3.2.4 
ensure that the emergency power source starting and loading times 
continue to meet the current technical specification requirements. 
Also, these time delays are long enough to preclude false trips due 
to anticipated voltage transients (e.g., during motor starts). The 
relay calibration surveillance procedure will establish acceptance 
criteria for each relay to ensure that the total times specified in 
Table 3.2.4 are not exceeded. The proposed surveillance testing and 
calibration frequency of every refueling outage is consistent with 
the requirements in the current technical specification.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    There are no new failure modes associated with this change since 
the proposed requirements will ensure the LNP instrumentation system 
is available to perform its safety function. Individual voltage 
sensing relays, when removed from their cases, would provide the 
tripped contact configuration. The proposed technical specification 
would allow relays to be placed in the tripped condition as long as 
it would not inhibit the LNP function or cause an inadvertent 
initiation. Additionally, since the design function to ensure that 
adequate power is available to operate the emergency safeguards 
equipment has not changed, no new accident or accident of a 
different kind is created.
    3. Involve a significant reduction in the margin of safety.
    The protective boundaries are not affected because the 
consequences of any design basis accident are not changed. Since the 
protective boundaries are not affected, the safety limits are also 
unaffected. The proposed change maintains the basis of the technical 
specifications by ensuring that adequate electrical power is 
available to operate the emergency safeguards equipment. By 
maximizing the operability of the LNP instrumentation without 
requiring high risk testing, the proposed change will improve the 
margin of safety as related to availability of electric power to 
safety related loads.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: January 13, 1995
    Description of amendment request: The proposed amendment would 
revise the Administrative Controls Section (6.0) of the Technical 
Specifications (TS) for Hope Creek Generating Station to reflect 
organizational changes and resultant management title changes. As 
indicated on the marked-up pages in Attachment 2, PSE&G requests that: 
1) Vice President and Chief Nuclear Officer will be replaced with Chief 
Nuclear Officer and President - Nuclear Business Unit in TS 6.1.2, 
6.2.1.c, 6.5.2.4.3.g, 6.5.2.4.4.a, 6.5.2.4.4.b, 6.5.2.6, 6.6.1.b, 
6.7.1.a, and 6.7.1.c. 2) Vice President and Chief Nuclear Officer will 
be replaced with Vice President - Nuclear Operations in TS 6.5.1.8.b, 
and 6.5.1.9. 3) In addition, General Manager - Quality Assurance and 
Nuclear Safety will be replaced with Director - Quality Assurance and 
Nuclear Safety Review in TS 6.5.1.8.b, 6.5.1.9, 6.5.2.2, 6.5.2.4.3.g, 
6.7.1.a, 6.7.1.c.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed management title changes from Vice President and 
Chief Nuclear Officer to Chief Nuclear Officer and President - 
Nuclear Business Unit or Vice President - Nuclear Operations, and 
from General Manager - Quality Assurance and Nuclear Safety to 
Director - Quality Assurance and Nuclear Safety Review are 
administrative in nature and do not affect assumptions contained in 
the plant safety analysis, the physical design and/or operation of 
the plant, nor do they affect Technical Specifications that preserve 
safety analysis assumptions. Therefore, the proposed changes do not 
[[Page 32372]] involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The changes being proposed are purely administrative and will 
not lead to material procedure changes or to physical modifications. 
Therefore, the proposed changes do not create the possibility of a 
new or different type of accident from any accident previously 
evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The changes being proposed are administrative in nature and do 
not relate to or modify the safety margins defined in and maintained 
by the Technical Specifications. The changes discussed herein do not 
reduce the Technical Specification safety margin since all 
organizational responsibilities are being adequately implemented, 
and all personnel in place are properly qualified. 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: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070
    Attorney for licensee: M. J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: May 19, 1995 (TS 95-07)
    Description of amendment request: The proposed change would (1) 
modify Surveillance Requirement (SR) 4.1.1.3 to allow suspension of the 
end of life (EOL) moderator temperature coefficient (MTC) surveillance 
measurement provided the benchmark criteria and the Revised Prediction 
as documented in the Core Operating Limits Report (COLR) are satisfied. 
The SR would also indicate that the data required for the calculation 
of the Revised Prediction is provided in the Most Negative Temperature 
Coefficient Limit Report per Specification 6.9.1.15. In addition, a 
grammatical error affecting the Unit 1 SR would be corrected; (2) 
modify Technical Specifications (TS) 6.9.1.14, COLR, by adding to the 
list of references: WCAP-13749-P-[A], ``Safety Evaluation Supporting 
the Conditional Exemption of the Most Negative EOL Moderator 
Temperature Coefficient Measurement,'' May 1993 (Proprietary) 
(Methodology for Specification 3.1.1.3 - Moderator Temperature 
Coefficient); (3) add Specification 6.9.1.15, which would require that 
the Most Negative MTC Report be prepared at least 60 days prior to the 
date the limit would become effective and be maintained on file. Also, 
the TS would require that the data required for the determination of 
the Revised Prediction of the 300 ppm/RTP MTC per WCAP-13749-P-[A] be 
included in the report.
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The conditional exemption of the most negative moderator 
temperature coefficient (MTC) measurement does not change the most 
negative MTC surveillance requirement (SR) and limiting condition of 
operation (LCO) limits in the TSs. Since these MTC values are 
unchanged, and since the basis for the derivation of these values 
from the safety analysis moderator density coefficient (MDC) is 
unchanged, the constant MDC assumed for the Updated Final Safety 
Analysis Report (UFSAR) safety analyses will also remain unchanged. 
Therefore, no change in the modeling (i.e., probabilities) of the 
accident analysis conditions or response is necessary in order to 
implement the change to the conditional exemption methodology. In 
addition, since the constant MDC assumed in the safety analyses is 
not changed by the conditional exemption of the most negative MTC SR 
measurement, the consequences of an accident previously evaluated in 
the UFSAR are not increased. The dose predictions presented in the 
UFSAR for a steam generator tube rupture remain valid such that more 
severe consequences will not occur. Additionally, since mass and 
energy releases for a loss-of-coolant accident and a steamline break 
are not increased as a result of the unchanged MDC, the dose 
predictions for these events presented in the UFSAR also remain 
bounding.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Since the end-of-life MTC is not changed by the conditional 
exemption methodology of WCAP-13749-P, the possibility of an 
accident, which is different than any already evaluated in the 
UFSAR, has not been created. No new or different failure modes have 
been defined for any system or component nor has any new limiting 
single failure been identified. Conservative assumptions for the MDC 
have already been modeled in the UFSAR analyses. These assumptions 
will remain valid since the conditional exemption methodology 
documented in WCAP-13749-P does not change the safety analysis MDC 
nor the TS values of the MTC.
    3. Involve a significant reduction in a margin of safety.
    The conditional exemption methodology is documented in WCAP-
13749-P. This WCAP has been evaluated (Reference: SECL 93-117,R1) 
relative to the design basis, including the TSs, and has been 
determined to bound the conditions under which the specifications 
permit operation. The results as presented in the UFSAR remain 
bounding since the MDC assumed in the safety analyses and the 
limiting conditions for operation and SR MTCs in the TSs remain 
unchanged. Therefore, the margin of safety, as defined in the bases 
to these TSs, is not reduced.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library,1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

Tennessee Valley Authority, Docket No. 50-328, Sequoyah Nuclear 
Plant, Unit 2, Hamilton County, Tennessee

    Date of amendment request: May 19, 1995 (TS 95-13)
    Description of amendment request: The proposed change would revise 
License Condition 2.C.(17) to extend the required surveillance interval 
to May 4, 1996, for Surveillance Requirement 4.3.2.1.3. The proposed 
change would extend the Engineered Safety Features Response Time 
instrument tests required at 36-month intervals shown in Table 3.3-3 
associated with safety injection, feedwater isolation, containment 
isolation Phase A, auxiliary feedwater pump, essential raw cooling 
water system, emergency gas treatment system, containment spray, 
containment isolation Phase B, turbine trip, 6.9-kilovolt shutdown 
board-degraded voltage or loss of voltage, and automatic switchover to 
containment sump actuations. The proposed extension will limit the 
interval past the allowable extension provided by TS 4.0.2 to 4.5 
months. [[Page 32373]] 
    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:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c).
    Operation of Sequoyah Nuclear Plant (SQN) in accordance with the 
proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is temporary and allows a one-time extension 
of Surveillance Requirement 4.3.2.1.3 for Cycle 7 to allow 
surveillance testing to coincide with the seventh refueling outage. 
The proposed surveillance interval extension will not cause a 
significant reduction in system reliability nor affect the ability 
of the systems to perform their design function. Current monitoring 
of plant conditions and continuation of the surveillance testing 
required during normal plant operation will continue to be performed 
to ensure conformance with TS operability requirements. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Extending the surveillance interval for the performance of 
specific testing will not create the possibility of a new or 
different kind of accidents. No changes are required to any system 
configurations, plant equipment, or analyses. Therefore, this change 
will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    Surveillance interval extensions will not impact any plant 
safety analyses since the assumptions used will remain unchanged. 
The safety limits assumed in the accident analyses and the design 
function of the equipment required to mitigate the consequences of 
any postulated accidents will not be changed since only the 
surveillance test interval is being extended. Historical performance 
generally indicates a high degree of reliability, and surveillance 
testing performed during normal plant operation will continue to be 
performed to verify proper performance. Therefore, the plant will be 
maintained within the analyzed limits, and the proposed extension 
will not significantly reduce the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: May 1, 1995
    Brief description of amendments: The proposed amendment would: (1) 
reduce the minimum fuel oil volume requirement during MODES 5 and 6, 
for OPERABLE emergency diesel generators (EDG), and (2) allow continued 
OPERABLE status of diesel generators during all MODES, for 48 hours 
with greater than 6-day supply of diesel fuel for the associated diesel 
generator.
    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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    REDUCTION IN MINIMUM DIESEL FUEL STORED VOLUME WHILE SHUTDOWN
    The first proposed change reduces the diesel fuel oil inventory 
required during plant shutdown conditions (MODES 5 and 6). The 
current fuel oil inventory requirement is the same for plant 
operation (MODES 1, 2, 3 and 4) and for plant shutdown. This current 
inventory requirement is based upon the seven days continuous 
operation of a diesel generator at its rated capacity which 
encompasses all load demands for the Loss of Coolant Accident 
concurrent with a Loss of Offsite Power (LOCA/LOOP) scenario. 
Because of reduced temperature and pressure, LOCA/LOOP is a less 
significant and probable event in MODES 5 and 6. The bounding 
scenario is considered to be a Loss of Offsite Power (LOOP) while 
the plant is shutdown (in MODES 5 and 6). The new diesel fuel oil 
inventory required during plant shutdown conditions is based on 
LOOP. Because this change only affects diesel fuel inventory, there 
is no impact on the probability of an accident. The consequences of 
LOOP event are unchanged since sufficient fuel remains available to 
allow the diesel generators to support mitigation of the event. 
Because seven days of fuel are required, there is no change in the 
consequences of any event which requires the diesel generators. 
Therefore, there is no significant increase in the probability or 
consequences of an accident previously evaluated as a result of this 
proposed change.
    ADDITION OF REMEDIAL ACTION TO RESTORE THE STORED VOLUME OF 
DIESEL FUEL
    The second proposed change applies to all MODES of operation. 
This change allows the diesel generator to remain OPERABLE if the 
fuel oil inventory falls below the minimum required in the storage 
system (i.e., fuel volume for 7-day operation of the diesel 
generator) but remains above a fuel volume for 6 days operation of 
the diesel generator. The minimum required fuel oil volume must be 
restored within 48 hours of falling below the limit. This relaxation 
by 48 hours allows sufficient time to replenish the required fuel 
oil volume and complete any required analysis prior to fuel oil 
addition to the storage tank. Because this change only affects 
diesel generator fuel inventory, there is no impact on the 
probability of an accident. Since the fuel oil replenishment can be 
obtained in less than six days after an event, there is no 
significant increase in the probability of a loss of all AC power 
(i.e., Station Blackout). Because the remaining fuel oil volume is 
larger than 6-day fuel supply and actions are initiated to obtain 
replenishment within this brief period, the proposed change does not 
involve a significant increase in the consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    REDUCTION IN MINIMUM DIESEL FUEL STORED VOLUME WHILE SHUTDOWN
    The first proposed change reduces the diesel fuel oil inventory 
required for plant shutdown conditions. As described above, LOOP is 
the limiting condition for diesel fuel oil inventory requirements 
for a plant in the shutdown condition. As the proposed fuel 
inventory is adequate for a shutdown LOOP and no hardware changes or 
system operation changes are involved, no new failure modes are 
introduced and hence, no new or different accidents from any 
previously evaluated are created.
    ADDITION OF REMEDIAL ACTION TO RESTORE THE STORED VOLUME OF 
DIESEL FUEL
    The second proposed change only affects diesel generator fuel 
inventory as well. There are no hardware changes and no changes in 
system operations involved; therefore, no new or different accidents 
from any accident previously evaluated are created.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The intent of the Technical Specification is to conservatively 
assure sufficient fuel to assure diesel generator operation to 
support mitigation of postulated events. This intent is accomplished 
by conservatively assuring a seven day supply of fuel. Seven days 
fuel supply is considered sufficient to support the initial 
mitigation activities, identify the need for additional fuel, 
arrange for delivery, test and then add fuel to the storage tanks, 
if needed. The current diesel fuel oil inventory for operating 
conditions (MODES 1, 2, 3 and 4), is sufficient to conservatively 
support seven days of diesel generator operation for a LOCA with 
LOOP condition.
    REDUCTION IN MINIMUM DIESEL FUEL STORED VOLUME WHILE SHUTDOWN 
[[Page 32374]] 
    The proposed diesel fuel oil inventory for shutdown conditions 
(MODES 5 and 6), is adequate to conservatively support seven days of 
diesel generator operation for LOOP conditions. The proposed 
reduction in inventory between operating and shutdown conditions 
continues to support the different transient conditions which are 
applicable to the different modes of operation. Even though the 
minimum storage requirement during shutdown is being reduced, the 
basis of this specification continues to be conservatively satisfied 
and therefore this license amendment request does not involve a 
significant reduction in a margin of safety.
    ADDITION OF REMEDIAL ACTION TO RESTORE THE STORED VOLUME OF 
DIESEL FUEL
    The second proposed change which is applicable to all MODES of 
operation, allows 48 hours to restore diesel generator fuel oil 
inventory to the seven-day level as long as the inventory does not 
fall below the six-day level. The probability of a LOOP during this 
period is low. The 6-day fuel oil supply is calculated with adequate 
margin similar to the calculation of 7-day fuel oil inventory. In 
spite of the potential that there may be slightly less fuel 
available inlenishment within this brief period. Based on this and 
the low probability of an event during this brief period, it is 
considered that this change request does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019
    Attorney for licensee: George L. Edgar, Esq., Newman and 
Holtzinger, 1615 L Street, N.W., Suite 1000, Washington, D.C. 20036
    NRC Project Director: William D. Beckner

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: May 4, 1994
    Brief description of amendments: The amendments revise Limiting 
Condition for Operation (LCO) 3.4.8.3 and Surveillance Requirement 
4.4.8.3.1, ``Overpressure Protection Systems.'' Specifically, the LCO 
and surveillance requirements are revised to clarify that both shutdown 
cooling system (SCS) suction relief valves shall be OPERABLE and 
aligned to provide overpressure protection not only during reactor 
coolant system (RCS) cooldown and heatup evolutions, but also during 
any steady-state temperature periods in the course of RCS cooldown or 
heatup evolutions.
    Date of issuance: June 2, 1995
    Effective date: June 2, 1995
    Amendment Nos.: Unit 1 - Amendment No. 93; Unit 2 - Amendment No. 
80; Unit 3 - Amendment No. 63
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42333) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 2, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station,Plymouth County, Massachusetts

    Date of application for amendment: November 22, 1994
    Brief description of amendment: This amendment revises the 
suppression chamber water level operating range, increasing it 2 
inches, and revises the water level recorder range in response to a 
commitment from an inspection.
    Date of issuance: June 1, 1995
    Effective date: June 1, 1995
    Amendment No.: 163
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 18, 1995 (60 FR 
3672) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated June 1, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location:  Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson 
SteamElectric Plant, Unit No. 2, Darlington County, South Carolina

    Date of application for amendment: February 24, 1995
    Brief description of amendment: The proposed change would remove 
Section 4.3 from the Technical Specifications (TS) because the primary 
system testing following opening is already performed in accordance 
with the American Society of Mechanical Engineers Boiler and Pressure 
Vessel Code, as implemented in the licensee's inservice inspection 
program as required by TS 4.0.1.
    Date of issuance: May 30, 1995Effective date: May 30, 1995
    Amendment No.: 165
    Facility Operating License No. DPR-23. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16183) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 30, 1995.No significant 
hazards consideration comments received: No [[Page 32375]] 
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois; 
Docket Nos. 50-254 and 50-265, Quad Cities Nuclear Power Station, 
Units 1 and 2, Rock Island County, Illinois

    Date of application for amendments: October 15, 1992, as 
supplemented March 9, 1993.
    Brief description of amendments: The amendments would modify the 
existing Dresden and Quad Cities Technical Specifications (TS) to 
format them in the style of the Boiling Water Reactor 4 (BWR) Standard 
Technical Specifications (STS). The amendments deal specifically with 
Section 3/4.4, ``Standby Liquid Control System (SLCS).''
    Date of issuance: June 8, 1995
    Effective date: For Dresden, immediately, to be implemented no 
later than December 31, 1995; for Quad Cities, immediately, to be 
implemented no later than June 30, 1996.
    Amendment Nos.: 133, 127, 154, and 150
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29, and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36429) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 8, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

    Date of application for amendments: April 10, 1995
    Brief description of amendments: The amendments would change the 
Technical Specifications by (1) revising the low pressure value at 
which the High Pressure Coolant Injection (HPCI) and Reactor Core 
Isolation Cooling (RCIC) systems can be tested to 150 psig, and (2) 
testing these systems against a system head corresponding to reactor 
vessel pressure when steam is supplied to the turbines at 920 psig to 
1005 psig for high pressure testing and 150 psig to 325 psig for low 
pressure testing.
    Date of issuance: May 30, 1995
    Effective date: Immediately and shall be implemented within 60 
days.
    Amendment Nos.: 153 and 149
    Facility Operating License Nos. DPR-29 and DPR-30: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1995 (60 FR 
21009) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Dixon Public Library, 221 
Hennepin Avenue, Dixon, Illinois 61021.

Commonwealth Edison Company, Docket Nos. 50-295 and 50-304, Zion 
Nuclear Power Station Units 1 and 2, Lake County, Illinois

    Date of application for amendments:  November 21, 1994.
    Brief description of amendments: The amendments add footnotes in 
Limiting Condition for Operation 3.15.2.A of the Technical 
Specifications (TS) to allow a one-time extension of the allowed outage 
time (AOT) for an inoperable reserve offsite power source from 72 hours 
to 14 days. To provide additional assurance that redundant sources of 
power to the operating unit are operable during the AOT outage, the 
amendment also adds footnotes in Surveillance Requirement 4.15.2.A of 
the TS to modify the emergency diesel generator and the normal offsite 
power source testing requirements.
    Date of issuance: May 31, 1995
    Effective date: May 31, 1995
    Amendment Nos.: 163 and 151
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
500). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 31, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, MichiganDate of application for amendment: October 
20, 1992

    Brief description of amendment: This amendment revises Technical 
Specification 5.3.1a to account for changes being made to the Palisades 
Final Safety Analysis Report (FSAR) Section 4.2 following replacement 
of the steam generators.
    Date of issuance: May 22, 1995
    Effective date: May 22, 1995
    Amendment No.: 166
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18624) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 22, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendment: January 13, 1995, as 
supplemented April 12 and 27, 1995
    Brief description of amendment: This amendment revises the 
Technical Specifications to allow installed primary and secondary 
safety valve settings to be within a 3% tolerance of their nominal 
settings, but would require returning the valve settings to within 1% 
of the nominal settings if the valves are removed from the piping for 
maintenance or testing.
    Date of issuance: June 8, 1995
    Effective date: June 8, 1995
    Amendment No.: 167
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11130) The April 12 and 27, 1995, letters provided clarifying 
information in response to the staff's request for additional 
information of April 11, 1995, and a telephone request for information 
on the Palisades loss of load analysis contained in the January 13, 
1995, submittal. This information was within the scope of the original 
application and did not change the staff's initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
June 8, 1995.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423. [[Page 32376]] 

Detroit Edison Company, Docket No. 50-341, Fermi-2, Monroe County, 
Michigan Date of application for amendment: September 13, 1993

    Brief description of amendment: The amendment revises Technical 
Specification (TS) 6.5.2.8 to relocate audit frequencies from the TS to 
the Quality Assurance Program located in Chapter 17.2 of the Updated 
Final Safety Analysis Report. A related change to extend the frequency 
of the use of an independent fire contractor to every third fire 
protection audit was denied.
    Date of issuance: May 23, 1995
    Effective date: May 23, 1995, with full implementation within 45 
days.
    Amendment No.: 104
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18625) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 23, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

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

    Date of amendment request: September 16, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications by removing the incore detection system requirements. 
These requirements are to be relocated in the Updated Final Safety 
Analysis Report.
    Date of issuance: May 30, 1995
    Effective date: May 30, 1995, to be implemented within 60 days.
    Amendment No.: 107
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57851) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 30, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122.

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

    Date of application for amendments: January 20, 1995
    Brief description of amendments: These amendments will relocate the 
operability requirements for Incore Detectors in Technical 
Specification 3/4.3.3.2 to the Updated Final Safety Analysis Report, 
and revise Linear Heat Rate Surveillance 4.2.1.4, and Special Test 
Exceptions Surveillances 4.10.2.2, 4.10.4.2 (Unit 2 only), and 
4.10.5.2, accordingly.
    Date of issuance: June 6, 1995
    Effective date: June 6, 1995
    Amendment Nos.: 136 and 75
    Facility Operating License Nos. DPR-67 and NPF-16: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11132) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 6, 1995No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

Georgia Power Company, Oglethorpe Power Corporation, Municipal 
Electric Authority of Georgia, City of Dalton, Georgia, Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of application for amendments: December 29, 1994, as 
supplemented by letter dated May 2, 1995.
    Brief description of amendments: The amendments revise TS 3/4.3, 
Instrumentation and its associated Bases, and TS 3/4.8, Electrical 
Power Systems to specify the appropriate actions to take in the event 
that an automatic load sequencer must be taken out of service or 
becomes inoperable.
    Date of issuance: May 31, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 86 and 64
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6301). The May 2, 1995, letter provided minor editorial changes that 
did not change the scope of the December 29, 1994, application and 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated May 31, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Entergy Operations, Inc., Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 13, 1993, as supplemented by 
letter dated October 18, 1993
    Brief description of amendment: The amendment revised the River 
Bend Station, Unit 1 operating license to reflect a change in ownership 
of Gulf States Utilities (GSU). GSU, which ownes a 70 percent undivided 
interest in the River Bend Station, is a wholly-owned subsidiary 
company of Entergy Corporation. This amendment was originally issued on 
December 16, 1993, as License Amendment No. 69.
    Date of issuance: June 8, 1995.
    Effective date: June 8, 1995.
    Amendment No.: 78
    Facility Operating License No. NPF-47. The amendment revised the 
operating license.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36436) The October 18, 1993, supplemental letter provided 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 June 8, 1995.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803.

Gulf States Utilities Company, Cajun Electric Power Cooperative, 
and Entergy Operations, Inc., Docket No. 50-458, River Bend 
Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: January 13, 1993, as supplemented by 
letter dated June 29, 1993
    Brief description of amendment: The amendment revised the River 
Bend Station, Unit 1 operating license to include as a licensee, 
Entergy Operations, Inc. (EOI), and to authorize EOI to use and operate 
River Bend and to possess and use related licensed nuclear materials. 
This amendment was originally issued on December 16, 1993 as License 
Amendment No. 70.
    Date of issuance: June 8, 1995
    Effective date: June 8, 1995
    Amendment No.: 79 [[Page 32377]] 
    Facility Operating License No. NPF-47. The amendment revised the 
operating license.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36436) The June 29, 1993, supplemental letter provided 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 June 8, 1995.No 
significant hazards consideration comments received. Yes. Comments and 
a request for hearing were received from Cajun Electric Power 
Cooperative of Baton Rouge, Louisiana.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center, 
Linn County, Iowa

    Date of application for amendment: March 1, 1995
    Brief description of amendment: The amendment revised the 
surveillance criteria for certain pumps and valves in the Low Pressure 
Coolant Injection (LPCI) subsystem; the Core Spray subsystems; and the 
Residual Heat Removal (RHR) Service Water, High Pressure Coolant 
Injection (HPCI), Emergency Service Water (ESW), and River Water Supply 
systems. The surveillance criteria changed from every three months to 
the testing frequency specified in the Inservice Testing program.
    Date of issuance: May 18, 1995
    Effective date: May 18, 1995
    Amendment No.: 210
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18626) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 18, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

IES Utilities Inc., Docket No. 50-331, Duane Arnold Energy, Center, 
Linn County, Iowa

    Date of application for amendment: March 10, 1995
    Brief description of amendment: The amendment deletes Technical 
Specification Sections 3.7/4.7.H.3 to eliminate redundant Limiting 
Conditions of Operation and Surveillance Requirements for the 
containment hydrogen and oxygen analyzers.
    Date of issuance: May 31, 1995
    Effective date: May 31, 1995
    Amendment No.: 211
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20518) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 31, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

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

    Date of amendment request: September 28, 1990
    Brief description of amendment: The amendment revised the Technical 
Specifications to establish periodic operability testing of the reactor 
vessel overfill protection system. The changes were requested to 
satisfy a commitment in the licensee's response to Generic Letter 89-
19, ``Request for Action Related to Resolution of Unresolved Safety 
Issue (USI) A-47.''
    Date of issuance: June 8, 1995
    Effective date: June 8, 1995
    Amendment No.: 169
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 31, 1990 (55 FR 
45885) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 8, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, NE 68305.

Northeast Nuclear Energy Company, Docket No. 50-245, 
MillstoneNuclear Power Station, Unit 1, New London County, 
Connecticut

    Date of application for amendment: March 31, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to increase the as-found setpoint tolerance of the 
safety/relief valves (SRVs) from plus or minus 1% to plus or minus 3%. 
In addition, the amendment (1) allows the as-found condition of one SRV 
to be inoperable, (2) clarifies the 1325 psig safety limit wording, (3) 
increases the number of SRVs to be tested during each refueling outage, 
(4) makes editorial changes to reflect the TS changes, and (5) revises 
the bases for the applicable sections.
    Date of issuance: May 31, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 82
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20520) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 31, 1995. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendment: January 23, 1995.
    Brief description of amendment: The amendment revises the Technical 
Specifications to modify the containment spray system by replacing the 
present sodium hydroxide spray additive with the trisodium phosphate 
dodecahydrate pH control agent.
    Date of issuance: May 26, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 115
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11136). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 26, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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 application for amendments: September 20, 1994, as 
supplemented by letter dated April 14, 1995. [[Page 32378]] 
    Brief description of amendments: The proposed amendments revise 
surveillance requirements (SRs) as recommended by NRC Generic Letter 
(GL) 93-05, ``Line-Item Technical Specification Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation'' of the 
combined Technical Specifications (TS) for the Diablo Canyon Nuclear 
Power Plant Unit Nos. 1 and 2. The specific TS changes are as follows:
    (1) TS SR 4.1.3.1.2 is revised to change the frequency for testing 
the movability of the control rods from at least once per 31 days to at 
least once per 92 days.
    (2) TS 3/4.3.2, Table 4.3-2, ``Engineered Safety Features Actuation 
System Instrumentation Surveillance Requirements,'' Functional Unit 
3.c.4), and TS 3/4.3.3.1, Table 4.3-3, ``Radiation Monitoring 
Instrumentation for Plant Operations SRs,'' is revised to change the 
monthly channel functional test to quarterly.
    (3) TS 3/4.5.1 is changed as follows: (a) TS SR 4.5.1.1a.1) is 
revised to more clearly state that the accumulator water volume and 
pressure must be verified to be within their limits. (b) TS SR 
4.5.1.1b. is revised to specify that the boron concentration 
surveillance is not required to be performed if the accumulator makeup 
source was the refueling water storage tank (RWST). (c) TS SR 4.5.1.2 
is relocated to plant procedures.
    (4) TS SR 4.5.2c.2) is revised to clarify that a separate 
containment entry to verify the absence of loose debris is not required 
after each containment entry.
    (5) TS SR 4.6.2.1d. is revised to change the frequency for a 
containment spray header flow test from at least once per 5 years to at 
least once per 10 years.
    (6) TS SR 4.6.4.2a. is revised to change the verification of the 
minimum hydrogen recombiner sheath temperature from at least once per 6 
months to at least once each refueling interval.
    (7) TS SR 4.7.1.2.1 is revised to change the surveillance frequency 
for testing each auxiliary feedwater (AFW) pump from at least once per 
31 days to at least once per 92 days on a staggered test basis.
    (8) TS SR 4.10.1.2 is revised to lengthen the allowed period of 
time for a rod drop test from 24 hours to 7 days prior to reducing 
shutdown margin to less than the limits of TS 3.1.1.1.
    (9) TS SR 4.11.2.6 is revised to change the surveillance frequency 
from 24 hours to 7 days when radioactive material is being added to the 
gas decay tanks and to add a requirement to monitor radioactive 
material concentrations in the gas decay tanks at least once per 24 
hours when system degassing operations are in progress.
    Date of issuance: May 26, 1995
    Effective date: May 26, 1995, to be implemented within 60 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 102; Unit 2 Amendment No. 
101
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53843) The April 14, 1995, letter provided clarifying information and 
did not change the initial no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 26, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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 application for amendments: December 30, 1994 (LAR 94-12)
    Brief description of amendments: These amendments clarify the 
technical specifications (TS) issued in license amendments 84/83 
associated with the Eagle 21 reactor protection system modification, 
delete TS references to RM-14A and RM-14B, remove cycle-specific TS 
requirements, and incorporate editorial corrections.
    Date of issuance: June 2, 1995
    Effective date: June 2, 1995, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 103; Unit 2 - Amendment No. 
102
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14026) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 2, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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 application for amendments: February 6, 1995, as 
supplemented by letters dated March 23, 1995, and May 22, 1995.
    Brief description of amendments: The amendments would allow the 
storage of fuel with enrichments up to and including 5.0 weight percent 
U-235, would clarify that substitution of fuel rods with filler rods is 
acceptable for fuel designs that have been analyzed with applicable 
NRC-approved codes and methods, and would allow the use of ZIRLO fuel 
cladding in the future in addition to Zircaloy-4.
    Date of issuance: June 7, 1995
    Effective date: June 7, 1995, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 104; Unit 2 - Amendment No. 
103
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11138) The licensee's supplemental letters provided additional 
clarifying information. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated June 7, 1995. No 
significant hazards consideration comments received: Yes. Comments were 
submitted by Jill ZamEk on behalf of the San Luis Obispo Mothers for 
Peace by letter dated March 30, 1995.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendment: November 23, 1994, as 
supplemented April 27, 1995.
    Brief description of amendment: This amendment revised the 
Technical Specifications Section VII.C., Plant Staff, to decrease the 
minimum staff requirements for the shift operating organization from 
five to two persons.
    Date of issuance: May 31, 1995
    Effective date: This license amendment is effective as of the date 
of [[Page 32379]] its issuance and must be fully implemented no later 
than 30 days from the date of issuance.
    Amendment No.: 28Facility License No. DPR-7: The amendment revised 
the TS.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11139) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 31, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Humboldt County Library, 636 F 
Street, Eureka, California 95501.

PECO Energy Company, Public Service Electric and Gas 
CompanyDelmarva Power and Light Company, and Atlantic City Electric 
Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
Station, Unit Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: February 10, 1995
    Brief description of amendments: These amendments correct 
administrative errors in Section 4.11.A of the Technical Specifications 
(TSs). The errors were made in the TSs by Amendments 9 and 7 dated June 
25, 1975.
    Date of issuance: May 30, 1995
    Effective date: May 30, 1995
    Amendments Nos.: 202 and 205
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20521) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 30, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

PECO Energy Company, Public Service Electric and Gas 
CompanyDelmarva Power and Light Company, and Atlantic City Electric 
Company,Docket Nos. 50-277 and 50-278, Peach Bottom Atomic Power 
Station,Unit Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: September 26, 1994
    Brief description of amendments: These amendments extend the 
surveillance test intervals and allowable out-of service times for the 
testing and or repair of instrumentation that actuate the Reactor 
Protection System, Primary Containment Isolation, Core and Containment 
Cooling systems, Control Rod Blocks, Radiation Monitoring systems and 
Alternate Rod Insertion/Recirculation Pump Trip.
    Date of issuance: June 6, 1995
    Effective date: June 6, 1995
    Amendments Nos.: 203 and 206
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14027) The supplemental letters dated January 5, and March 23, 1995, 
provided clarifying information and did not change the initial proposed 
no significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated June 6, 1995.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

Pennsylvania Power and Light Company, Docket Nos. 50-387 and 50-388 
Susquehanna Steam Electric Station, Units 1 and 2, Luzerne County, 
Pennsylvania

    Date of application for amendments: April 30, 1993
    Brief description of amendments: These amendments changed the 
Technical Specifications by deleting Section 3/4.3.8 of the Turbine 
Overspeed Protection System.
    Date of issuance: June 1, 1995
    Effective date: June 1, 1995
    Amendment Nos.: 146 and 116
    Facility Operating License Nos. NPF-14 and NPF-22. These amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 9, 1993 (58 FR 
32389) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated July 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701.

Philadelphia Electric Company, Docket Nos. 50-352 and 50-353, 
Limerick Generating Station, Units 1 and 2, Montgomery County, 
Pennsylvania

    Date of application for amendments: August 25, 1993, as 
supplemented by letters dated June 27, 1994, and May 5, 1995
    Brief description of amendments: These amendments modify Technical 
Specification Surveillance Requirement 4.7.1.3 to require that all 
spray pond spray network piping above the frost line be drained at an 
ambient temperature below 40 deg.F, and within 1 hour after being used 
only when the ambient air temperature is below 40 deg.F.
    Date of issuance: June 1, 1995
    Effective date: June 1, 1995
    Amendment Nos.: 90 and 54
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50972) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 1, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: August 30, 1994
    Brief description of amendment: The changes relocate Technical 
Specification (TS) 3.3.7.9, Loose Parts Detection System (LPDS), 
Surveillance Requirement 4.3.7.9, and associated Bases from the TSs to 
the Updated Final Safety Analysis Report. The TS index is also revised 
by removing the reference to LPDS.
    Date of issuance: May 25, 1995
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 73
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16197) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 25, 1995. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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

    Date of application for amendments:  February 9, 1995
    Brief description of amendments: The amendments revise the 
Administrative [[Page 32380]] Controls section of the Technical 
Specifications to reflect organizational changes and resultant 
management title changes.
    Date of issuance: June 6, 1995
    Effective date: June 6, 1995
    Amendment Nos.: 168 and 150
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 29, 1995 (60 FR 
16200) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated June 6, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079.

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

    Date of application for amendments: March 6, 1995
    Brief description of amendments: The amendments relocate the 
seismic and meteorological monitoring instrumentation from the 
Technical Specifications to the Final Safety Analysis Report in 
accordance with the ``Final Policy Statement on Technical 
Specifications Improvements for Nuclear Power Reactors,'' dated July 
22, 1993.
    Date of issuance: May 22, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 115 and 107
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18628) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 22, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear 
Plant, Unit 1, Hamilton County, Tennessee

    Date of application for amendment: April 6, 1995 (TS 95-09)
    Brief description of amendment: The amendment modifies Operating 
License Condition 2.C.(25) to provide a limited extension of the ice 
condenser surveillance test interval on Unit 1 to coincide with the 
Cycle 7 refueling outage.
    Date of issuance: May 30, 1995
    Effective date: May 30, 1995
    Amendment No.: 200
    Facility Operating License Nos. DPR-77: Amendment revises the 
technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20526) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 30, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: April 6, 1995
    Brief description of amendments: The amendments revise the 
surveillance requirement for the power range neutron flux channel 
calibration frequency from monthly to every 31 effective full power 
days and delays first performance of the surveillance after reaching 15 
percent power for 96 hrs.
    Date of issuance: May 30, 1995
    Effective date: May 30, 1995
    Amendment Nos.: 199 and 190
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20530) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 30, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location:  Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: April 6, 1995
    Brief description of amendments: The amendments revise the 
definition of core alteration, quadrant power tilt ratio, and modifies 
the operational mode parameters table in the Unit 1 technical 
specifications.
    Date of issuance: June 1, 1990
    Effective date: June 1, 1990
    Amendment Nos.: 201 and 191
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20531) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated June 1, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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: February 14, 1994 (TXX-94046 LAR 94-006)
    Brief description of amendments: The proposed changes would revise 
the Technical Specifications (TSs) for Comanche Peak Steam Electric 
Station, Units 1 and 2 in the following three areas: 1) a change to the 
allowable value for the Unit 2 pressurizer pressure-low and Unit 2 
overtemperature N-16 (OTN-16) reactor trip setpoints; 2) an 
administrative change to delete an option which allowed continued 
operation for a period of time when a reactor trip system (RTS) or 
engineered safety features actuation system (ESFAS) instrumentation or 
interlocks trip setpoint is found less conservative than the allowable 
value; and 3) an administrative change to combine the Unit 1 and Unit 2 
line items for RTS or ESFAS trip setpoint and allowable values which 
are the same.
    Date of issuance: May 31, 1995
    Effective date: May 31, 1995, to be implemented within 30 days.
    Amendment Nos.: Unit 1 - Amendment No. 41; Unit 2 - Amendment No. 
27
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32238) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 31, 1995.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.

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

    Date of amendment request: March 24, 1995
    Brief description of amendment: The amendment relaxes the 
requirement to [[Page 32381]] sample the accumulator after refilling 
from the RWST.
    Date of issuance: May 30, 1995
    Effective date: May 30, 1995, to be implemented within 30 days of 
issuance.
    Amendment No.: Amendment No. 87
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 12, 1995 (60 FR 
18632) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 30, 1995.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.

    Dated at Rockville, Maryland, this 14th day of June, 1995.

    For the Nuclear Regulatory Commission
John N. Hannon,
Acting Deputy Director, Division of Reactor Projects - III/IV, Office 
of Nuclear Reactor Regulation
[Doc. 95-15057 Filed 6-20-95; 8:45]
BILLING CODE 7590-01-F