[Federal Register Volume 60, Number 50 (Wednesday, March 15, 1995)]
[Notices]
[Pages 14015-14040]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-60315]



-----------------------------------------------------------------------



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 February 16, 1995, through March 3, 1995. 
The last biweekly notice was published on March 1, 1995.

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 April 14, 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. [[Page 14016]] 
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, 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.

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: January 31, 1995
    Description of amendments request: The proposed amendments would 
revise the Technical Specifications (TSs) for Calvert Cliffs, Unit Nos. 
1 and 2, to increase the amount of Trisodium Phosphate Dodecahydrate 
(TSP) located in the containment sump baskets required to be verified 
by TS surveillance. The requested change is the result of an reanalysis 
of the amount of TSP necessary to maintain the appropriate pH in the 
containment sump water subsequent to a Loss of Coolant Accident. 
Specifically, the request would change the TS value of TS 4.5.2.e.3 
from the existing amount of 100 ft3 to 289 ft3. TS 4.5.2.e.4 
would also be changed by moving the amounts of TSP and refueling water 
storage tank water to be used in the required tests to the TS Bases 
Section 3/4.5.2 and 3/4.5.3. These Bases sections would also be changed 
by modifying the test methods.
    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. Would not involve a significant increase in the probability 
orconsequences of an accident previously evaluated.
    Trisodium Phosphate Dodecahydrate (TSP) is stored in the 
containment lower level to raise the pH of the sump and spray water 
following a Loss of Coolant Accident (LOCA). As the pH of the water 
increases, more radioactive iodine is kept in solution and the 
possibility of airborne radioactivity leakage is decreased. An 
additional advantage of a higher pH is the beneficial reduction in 
chloride stress corrosion cracking of metal components in the 
containment following an accident.
    This chemical is an accident mitigator, not an accident 
initiator in that it is not used until after an accident has 
occurred. At the time it goes into solution, the accident has 
occurred, containment spray has been activated and water has 
collected in the containment sump. Therefore, increasing the 
Technical Specification minimum amount verified to be in each 
containment will not involve a significant increase in the 
probability of an accident previously evaluated.
    Updated Final Safety Analysis Report, Chapter 14.24, ``Maximum 
Hypothetical Accident'', uses an assumption of a pre-RAS minimum 
containment spray pH of 5.0 for the iodine removal calculation and a 
post-RAS sump pH of 7.0 for iodine retention. Raising the pH to 7.0 
does not increase the consequences of an accident previously 
evaluated.
    The proposed change to Technical Specification 4.5.2.e.4 would 
remove the amounts of chemical and water used in the test to the 
Bases. This relocation will not alter the test method or acceptance 
criteria, but will allow adjustments to the ratio of TSP and borated 
water under the controls of 10 CFR 50.59 to reflect changes in plant 
conditions. In the Bases, the amount of TSP used in the test is 
changed to reflect the ratio of TSP to water that would be found in 
the containment following a LOCA. The specified concentration of 
boron in the test reflects the highest concentration that could be 
found in the containment following a LOCA. The test temperature is 
changed to 120 deg.F which is well below the temperature expected to 
be found in the containment sump following a LOCA. The decanting of 
the solution does not change the intent of the test method since the 
dissolving period will still be conducted without agitation.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated. [[Page 14017]] 
    The addition of more TSP does not represent a significant change 
in the configuration or operation of the plant. Trisodium Phosphate 
Dodecahydrate is currently present in the containment lower level. 
There are no physical changes which result from the increase in 
volume. The proposed change to Technical Specification 4.5.2.e.4 
would move the amounts of chemical and water used in the test to the 
Bases. This relocation will not alter the test method or acceptance 
criteria, but will allow adjustments to the ratio of TSP and borated 
water under the controls of 10 CFR 50.59 to reflect changes in plant 
conditions. In the Bases, the amount of TSP used in the test is 
changed to reflect the ratio of TSP to water that would be found in 
the containment following a LOCA. The specified concentration of 
boron in the test reflects the highest concentration that could be 
found in the containment following a LOCA. The test temperature is 
changed to 120 deg.F which is well below the temperature expected to 
be found in the containment sump following a LOCA. The decanting of 
the solution does not change the intent of the test method since the 
dissolving period will still be conducted without agitation.
    Therefore, this change would not create the possibility of a new 
or different type of accident from any accident previously 
evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    Trisodium Phosphate Dodecahydrate is stored in the containment 
lower level to raise the pH of the sump and spray water following a 
LOCA. As the pH of the water increases, more radioactive iodine is 
kept in solution and the possibility of airborne radioactivity 
leakage is decreased. Additionally, a higher pH has a beneficial 
effect on chloride stress corrosion cracking of metal components in 
the containment.
    Technical Specification 4.5.2.e.3 requires verification that a 
minimum volume of TSP is contained in the storage baskets in each 
containment. This change proposes to increase that volume. The 
increased volume will ensure the containment sump, when filled with 
water, will have an acceptable pH following a LOCA.
    The proposed change to Technical Specification 4.5.2.e.4 would 
move the amounts of chemical and water used in the test to the 
Bases. This relocation will not alter the test method or acceptance 
criteria, but will allow adjustments to the ratio of TSP and borated 
water under the controls of 10 CFR 50.59 to reflect changes in plant 
conditions. In the Bases, the amount of TSP used in the test is 
changed to reflect the ratio of TSP to water that would be found in 
the containment following a LOCA. The specified concentration of 
boron in the test reflects the highest concentration that could be 
found in the containment following a LOCA. The test temperature is 
changed to 120 deg.F which is well below the temperature expected to 
be found in the containment sump following a LOCA. The decanting of 
the solution does not change the intent of the test method since the 
dissolving period will still be conducted without agitation.
    Therefore, this change would not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: February 9, 1995
    Description of amendment request: The proposed amendment would 
increase the Reactor High Water Level Trip Level Setting for the Group 
1 isolation. The change will allow an increase to the main steam 
isolation valve (MSIV) high water level isolation setpoint.
    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:
    In accordance with 10 CFR 50.91, Boston Edison submits the 
following analysis addressing the no significant hazards 
consideration. The proposed changes do not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Operation of the station in accordance with the proposed Trip 
Level Setting will not significantly increase the probability or 
consequences of an accident previously evaluated. The MSIV high 
water level isolation signal is provided to protect against rapid 
depressurization due to a pressure regulator malfunction during 
plant startup. The high water level isolation signal is not 
functional when the mode switch is in the RUN position. A high water 
level in the reactor vessel indicates that fuel is covered. 
Increasing the Trip Level Setting will have minimal effect on 
moisture carryover in the event of a pressure regulator failure at 
low reactor power. MSIV closure (Group 1) is initiated by low 
reactor pressure (810 psig) approximately 30 seconds into the event. 
The resulting reactor water level swell is not sufficient to reach 
the bottom elevation of the main steam lines.
    The proposed Technical Specification allowable value for the 
Reactor Low Level Trip Level Setting and the Reactor Low Low Water 
Level Trip Level setting does not involve significant increase in 
the probability or consequence of an accident.
    (2) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed change does not affect the Group 1 isolation safety 
function. The change does not involve any plant hardware changes 
that could introduce any new failure modes or effects; thus, the 
change can not create the possibility of a new or different kind of 
accident from any previously analyzed.
    (3) Involve a significant reduction in a margin of safety.
    The proposed change does not affect the Group 1 isolation safety 
function. The proposed change is consistent with the FSAR [Final 
Safety Analysis Report] and Technical Specification basis associated 
with reactor vessel inventory control and main steam line flooding.
    The proposed change to the instrument calibration range does not 
affect the margin of safety for systems or components affected by 
the change. Operating Pilgrim in accordance with the proposed Trip 
Level Setting does not involve a significant reduction in the margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: February 6, 1995
    Description of amendment request: The change proposes to relocate 
the cycle specific core operating limits of Figure 3.1-1, Shutdown 
Margin Versus Boron Concentration, from Technical Specification (TS) 
3.1.1.2, Shutdown Margins - Modes 3, 4, and 5, to the Core Operating 
Limits Report (COLR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed change of relocating TS Figure 3.1-1, Shutdown 
Margin Versus Boron Concentration to the COLR has no influence 
[[Page 14018]] or impact to the probability or consequences of an 
accident. The revised TS will continue to implement the shutdown 
margin limits through reference to the Shutdown Margin Curve in the 
COLR. In addition, the COLR is subject to the existing controls of 
TS 6.9.1.6. Given that this change is an administrative relocation 
of the Shutdown Margin Curve to another TS controlled document, 
there would be no increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    No safety-related equipment, safety function, or plant operation 
will be altered as a result of this proposed change. The TS will 
continue to require operation within the required core operating 
limits. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in the margin of safety.
    Relocation of the Shutdown Margin Curve to the TS controlled 
COLR has no effect on the core operating limits currently in force 
in TS 3.1.1.2. Future revisions to the Shutdown Margin Curve are 
governed by TS 6.9.1.6 which stipulates the specific TS that 
reference the COLR limits and the methodologies utilized in 
developing those limits. Given that the change is an administrative 
relocation of the Shutdown Margin Curve to another TS controlled 
document, 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: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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: January 12, 1995
    Description of amendment request: The proposed amendments would 
revise and clarify portions of Technical Specification (TS) Section 
6.0, ``Administrative Controls,'' for the McGuire, Catawba, and Oconee 
nuclear stations. The licensee submitted a combined amendment request 
covering the three Duke Power nuclear stations. The proposed changes 
are described below.
    1. Remove the specific assignment of responsibilities for the 
review, distribution, and approval activities contained in the 
Technical Review and Control Section of each station's TS. The proposed 
specifications state that these activities will be performed by a 
knowledgeable individual/organization. Approval of the affected 
documents is to be at the appropriate manager/superintendent level as 
specified in Duke administrative controls.
    2. Move the requirement for the review of proposed changes in the 
stations' TS and Operating Licenses by the Duke Nuclear Safety Review 
Board (NSRB) to Duke administrative procedures (Selected Licensee 
Commitments documents) and change the wording of the requirements 
covering NSRB meeting frequency. The Oconee TS covering the NSRB are 
being rewritten to be consistent with McGuire and Catawba.
    3. Add Technical Review and Control Program implementation and 
Plant Operations Review Committee (PORC) implementation to the list of 
required procedures and programs for each nuclear station.
    4. Change or clarify certain TS administrative requirements 
covering technical review and control activities or records retention 
requirements.
    5. For Oconee only, under ``Station Operating Procedures,'' revise 
the TS requirements covering the review and approval of station 
procedures and temporary procedure changes such that these are now 
consistent with the corresponding requirements for McGuire and Catawba.
    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:
    (It should be noted that the licensee submitted a combined 
analysis that covers McGuire, Catawba, and Oconee nuclear stations.)
    Standard 1. The proposed amendments will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The provisions of these proposed amendments concern 
administrative changes in the stations' Technical Specifications 
involving the Technical Review and Control, Procedures and Programs/
Station Operating Procedures, and Records Retention/Station 
Operating Records portions of the Administrative Controls Section. 
The requested changes primarily affect review and control 
activities, but also include other administrative changes affecting 
the approval of station procedures (Oconee only), records retention, 
and definition of the term ODCM [offsite dose calculation manual] 
(McGuire and [Catawba]). The provisions of the proposed amendment 
primarily involve the relocation of existing Technical 
Specifications review, distribution, or approval requirements to 
internal Duke administrative controls. However, implementation of 
the proposed amendment does involve changes to several review/
distribution activities. Theses review/distribution activities are 
primarily for: 1) Proposed changes to the stations' Technical 
Specifications, 2) Proposed tests and experiments which affect 
nuclear safety and are not addressed in the stations' FSAR [Final 
Safety Analysis Report] or Technical Specifications, 3) 
Environmental radiological procedures, 4) Reportable events 
documentation and reports of violations of Technical Specifications, 
5) Reports of special reviews and investigations, and 6) Reports of 
unplanned onsite releases of radiological material to the environs. 
Planned implementation of the proposed Technical Specifications 
amendments utilizing Selected Licensee Commitments will result in 
the above items being reviewed/received by a different 
organizational unit in the future. The organizational unit is to be 
either the recently initiated Plant Operations Review Committee 
(PORC) or the General Manager, Environmental Services. Personnel 
serving on the PORC, and the General Manager, Environmental Services 
will be qualified based upon education and experience to review the 
operational and technical considerations involved with the 
applicable items listed above. No required reviews are being 
eliminated by the requested amendments, only the organizational 
units responsible for performing the reviews will be changed. Future 
reviews of theses items under the auspices of the PORC or the 
General Manager, Environmental Services will maintain a quality 
level equivalent to that being currently achieved by Duke's 
Qualified Reviewer Program, the Station Managers, or the
    Duke Nuclear Safety Review Board as applicable. Consequently, 
merely changing the organizational units performing future reviews, 
or making the additional administrative changes described above, 
results in no increase in the probability or consequences of an 
accident previously evaluated because the review function will 
continue to be conducted in an equivalent manner.
    The implementing SLC will also permit proposed amendments to the 
stations' Technical Specifications and Operating Licenses to be 
approved for the Station Manager by a designee. However, this 
individual will occupy a position equivalent to, or higher, in the 
Duke organization as the Station manager.
    Additionally, the proposed changes do not directly impact the 
design or operation of any plant systems or components any more so 
than the review and approval processes currently being conducted in 
accordance [[Page 14019]] with existing approved Technical 
Specifications.
    Standard 2. The proposed amendments will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed changes are administrative in nature and primarily 
cover the review, distribution, and/or approval function performed 
for items identified in existing Technical Specifications. The 
quality level of the future reviews will not decrease and the 
ability of Duke to identify the possibility for the concurrence of 
new or different kinds of accidents prior to implementation will be 
maintained. Of specific interest in the consideration of Standard 
2 is the review of proposed tests and experiments which 
affect station nuclear safety and are not addressed in the FSAR or 
Technical Specifications. The Technical Specifications required 
reviews of these tests and experiments are not being proposed for 
removal by these requested amendments. Only the organizational unit 
conducting the review of proposed tests and experiments is being 
changed by the requested amendments. The PORC, instead of the 
Station Manager, is being assigned the responsibility for conducting 
the reviews of proposed tests and experiments in the future. It is 
believed that the combined expertise of the PORC membership will 
enhance Duke's ability to identify potential situations which could 
possibly involve a new, or different, kind of accident.
    Standard 3. The proposed amendments will not involve a 
significant reduction in any margin of safety.
    The changes contained in the requested amendments are 
administrative in nature and do not impact the design capabilities 
or operation of any plant structures, systems, or components. There 
will be no reduction in margin of safety as a result of implementing 
these requested amendments. Impact upon margin of safety is a 
consideration primarily included in the 10 CFR 50.59 evaluation 
process conducted for station procedures, procedure changes, and 
nuclear station modifications. The 10 CFR 50.59 evaluation process 
in conducted under the auspices of the Duke Qualified Reviewer 
Program and is not affected by these requested amendments. The 
impact on margin of safety for future Technical Specifications and 
Operating License changes will be reviewed by the PORC, but these 
reviews will be equivalent in quality to the reviews presently 
conducted by the Qualified Reviewers.
    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

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: January 13, 1995
    Description of amendment request: The proposed amendments would 
increase the surveillance test intervals and allowed outage times for 
Reactor Trip System (RTS) and Engineered Safety Features Actuation 
System (ESFAS) equipment based upon analyses by Westinghouse for the 
Westinghouse Owners Group and approved by the NRC. The proposed changes 
to the RTS and ESFAS instrumentation are based upon WCAP-10271, its 
supplements, and the NRC's safety evaluation reports.
    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 - Operation of McGuire in accordance with the 
proposed license amendment[s] [do] not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The determination that the results of the proposed changes are 
within all acceptable criteria was established in the SERs prepared 
for WCAP-10271, WCAP-10271 Supplement 1, WCAP-10271 Supplement 2, 
and WCAP-10271 Supplement 2, Revision 1 issued by letters dated 
February 21, 1985, February 22, 1989, and April 30, 1990. 
Implementation of the proposed changes is expected to result in an 
acceptable increase in total RTS yearly unavailability. This 
increase, which is primarily due to less frequent surveillance, 
results in an increase of similar magnitude in the probability of an 
Anticipated Transient Without Scram (ATWS) and in the probability of 
core melt resulting from an ATWS and also results in a small 
increase in core damage frequency (CDF) due to ESFAS unavailability.
    Implementation of the proposed changes is expected to result in 
a significant reduction in the probability of core melt from 
inadvertent reactor trips. This is a result of a reduction in the 
number of inadvertent reactor trips (0.5 fewer inadvertent reactor 
trips per unit per year) occurring during testing of RTS 
instrumentation. This reduction is primarily attributable to testing 
in bypass and less frequent surveillance.
    The reduction in core melt frequency from inadvertent reactor 
trips is sufficiently large to counter the increase in ATWS core 
melt probability resulting in an overall reduction in total core 
melt probability.
    The values determined by the WOG and presented in the WCAP for 
the increase in CDF were verified by Brookhaven National Laboratory 
(BNL) as part of an audit and sensitivity analysis for the NRC 
staff. Based on the small value of the increase compared to the 
range of uncertainty in the CDF, the increase is considered 
acceptable.
    Changes to surveillance test frequencies for the RTS [reactor 
trip system] interlocks do not represent a significant reduction in 
testing. The currently specified test interval for interlock 
channels allows the surveillance requirement to be satisfied by 
verifying that the permissive logic is in its required state using 
the permissive annunciator window. The surveillance as currently 
required only verifies the status of the permissive logic and does 
not address verification of channel setpoint or operability. The 
setpoint verification and channel operability are verified after a 
refueling shutdown. The definition of the channel check includes 
comparison of the channel status with other channels for the same 
parameter. The requirement to routinely verify permissive status is 
a different consideration than the availability of trip or actuation 
channels which are required to change state on the occurrence of an 
event and for which the function availability is more dependent on 
the surveillance interval. The change in surveillance requirement to 
at least once every refueling does not therefore represent a 
significant change in channel surveillance and does not involve a 
significant increase in unavailability of the RTS.The proposed 
changes do not result in an increase in the severity or consequences 
of an accident previously evaluated. Implementation of the proposed 
changes affects the probability of failure of the RTS but does not 
alter the manner in which protection is afforded nor the manner in 
which limiting criteria are established.
    Criterion 2 - The proposed license amendment[s] [do] not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The proposed changes do not result in a change in the manner in 
which the RTS provides plant protection. No change is being made 
which alters the functioning of the RTS (other than in a test mode). 
Rather, the likelihood or probability of the RTS functioning 
properly is affected as described above. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident.
    The proposed changes do not involve hardware changes except 
those necessary to implement testing in bypass. Some existing 
instrumentation is designed to be tested in bypass and current 
Technical Specifications allow testing in bypass. Testing in bypass 
is also recognized by IEEE standards. Therefore, testing in bypass 
has been previously approved and implementation of the proposed 
changes for testing in bypass does not create the possibility of a 
new or different kind of accident from any previously evaluated. 
Furthermore, since the other proposed changes do not alter the 
functioning of the RTS, the possibility of a new or different kind 
of accident from any previously evaluated has not been created.
    Criterion 3 - The proposed license amendment[s] [do] not involve 
a significant reduction in a margin of safety.
    The proposed changes do not alter the manner in which safety 
limits, limiting safety [[Page 14020]] system setpoints, or limiting 
conditions for operation are determined. The impact of reduced 
testing other than as addressed above is to allow a longer time 
interval over which instrument uncertainties (e.g., drift) may act. 
Experience has shown that the initial uncertainty assumptions are 
valid for reduced testing.
    Implementation of the proposed changes is expected to result in 
an overall improvement in safety by:
    1) Less frequent testing will result in fewer inadvertent 
reactor trips and actuation of Engineered Safety Features Actuation 
System components.
    2) Higher quality repairs leading to improved equipment 
reliability due to longer allowable repair times.
    3) Improvements in the effectiveness of the operating staff in 
monitoring and controlling plant operation. This is due to less 
frequent distraction of the operator and shift supervisor to attend 
to instrumentation testing.
    The foregoing analysis demonstrates that the proposed 
amendment[s] to McGuire's Technical Specifications [do] not involve 
a significant increase in the probability or consequences of a 
previously evaluated accident, [do] not create the possibility of a 
new or different kind of accident, and [do] not involve a 
significant reduction in a margin of safety.
    Based upon the preceding analysis, Duke Power Company concludes 
that the proposed amendment[s] [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

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

    Date of amendment request: January 12, 1995
    Description of amendment request: The proposed amendments would 
revise and clarify portions of Technical Specification (TS) Section 
6.0, ``Administrative Controls,'' for the McGuire, Catawba, and Oconee 
nuclear stations. The licensee submitted a combined amendment request 
covering the three Duke Power nuclear stations. The proposed changes 
are described below.
    1. Remove the specific assignment of responsibilities for the 
review, distribution, and approval activities contained in the 
Technical Review and Control Section of each station's TS. The proposed 
specifications state that these activities will be performed by a 
knowledgeable individual/organization. Approval of the affected 
documents is to be at the appropriate manager/superintendent level as 
specified in Duke administrative controls.
    2. Move the requirement for the review of proposed changes in the 
stations' TS and Operating Licenses by the Duke Nuclear Safety Review 
Board (NSRB) to Duke administrative procedures (Selected Licensee 
Commitments documents) and change the wording of the requirements 
covering NSRB meeting frequency. The Oconee TS covering the NSRB are 
being rewritten to be consistent with McGuire and Catawba.
    3. Add Technical Review and Control Program implementation and 
Plant Operations Review Committee (PORC) implementation to the list of 
required procedures and programs for each nuclear station.
    4. Change or clarify certain TS administrative requirements 
covering technical review and control activities or records retention 
requirements.
    5. For Oconee only, under ``Station Operating Procedures,'' revise 
the TS requirements covering the review and approval of station 
procedures and temporary procedure changes such that these are now 
consistent with the corresponding requirements for McGuire and Catawba.
    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:
    (It should be noted that the licensee submitted a combined 
analysis that covers McGuire, Catawba, and Oconee nuclear stations.)
    Standard 1. The proposed amendments will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The provisions of these proposed amendments concern 
administrative changes in the stations' Technical Specifications 
involving the Technical Review and Control, Procedures and Programs/
Station Operating Procedures, and Records Retention/Station 
Operating Records portions of the Administrative Controls Section. 
The requested changes primarily affect review and control 
activities, but also include other administrative changes affecting 
the approval of station procedures (Oconee only), records retention, 
and definition of the term ODCM [offsite dose calculation manual] 
(McGuire and [Catawba]). The provisions of the proposed amendment 
primarily involve the relocation of existing Technical 
Specifications review, distribution, or approval requirements to 
internal Duke administrative controls. However, implementation of 
the proposed amendment does involve changes to several review/
distribution activities. These review/distribution activities are 
primarily for: 1) Proposed changes to the stations' Technical 
Specifications, 2) Proposed tests and experiments which affect 
nuclear safety and are not addressed in the stations' FSAR [Final 
Safety Analysis Report] or Technical Specifications, 3) 
Environmental radiological procedures, 4) Reportable events 
documentation and reports of violations of Technical Specifications, 
5) Reports of special reviews and investigations, and 6) Reports of 
unplanned onsite releases of radiological material to the environs. 
Planned implementation of the proposed Technical Specifications 
amendments utilizing Selected Licensee Commitments will result in 
the above items being reviewed/received by a different 
organizational unit in the future. The organizational unit is to be 
either the recently initiated Plant Operations Review Committee 
(PORC) or the General Manager, Environmental Services. Personnel 
serving on the PORC, and the General Manager, Environmental Services 
will be qualified based upon education and experience to review the 
operational and technical considerations involved with the 
applicable items listed above. No required reviews are being 
eliminated by the requested amendments, only the organizational 
units responsible for performing the reviews will be changed. Future 
reviews of these items under the auspices of the PORC or the General 
Manager, Environmental Services will maintain a quality level 
equivalent to that being currently achieved by Duke's Qualified 
Reviewer Program, the Station Managers, or the Duke Nuclear Safety 
Review Board as applicable. Consequently, merely changing the 
organizational units performing future reviews, or making the 
additional administrative changes described above, results in no 
increase in the probability or consequences of an accident 
previously evaluated because the review function will continue to be 
conducted in an equivalent manner.
    The implementing SLC will also permit proposed amendments to the 
stations' Technical Specifications and Operating Licenses to be 
approved for the Station Manager by a designee. However, this 
individual will occupy a position equivalent to, or higher, in the 
Duke organization as the Station Manager.
    Additionally, the proposed changes do not directly impact the 
design or operation of any plant systems or components any more so 
than the review and approval processes currently being conducted in 
accordance with existing approved Technical Specifications.
    Standard 2. The proposed amendments will not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The proposed changes are administrative in nature and primarily 
cover the review, [[Page 14021]] distribution, and/or approval 
function performed for items identified in existing Technical 
Specifications. The quality level of the future reviews will not 
decrease and the ability of Duke to identify the possibility for the 
occurrence of new or different kinds of accidents prior to 
implementation will be maintained. Of specific interest in the 
consideration of Standard 2 is the review of proposed tests 
and experiments which affect station nuclear safety and are not 
addressed in the FSAR or Technical Specifications. The Technical 
Specifications required reviews of these tests and experiments are 
not being proposed for removal by these requested amendments. Only 
the organizational unit conducting the review of proposed tests and 
experiments is being changed by the requested amendments. The PORC, 
instead of the Station Manager, is being assigned the responsibility 
for conducting the reviews of proposed tests and experiments in the 
future. It is believed that the combined expertise of the PORC 
membership will enhance Duke's ability to identify potential 
situations which could possibly involve a new, or different, kind of 
accident.
    Standard 3. The proposed amendments will not involve a 
significant reduction in any margin of safety.
    The changes contained in the requested amendments are 
administrative in nature and do not impact the design capabilities 
or operation of any plant structures, systems, or components. There 
will be no reduction in margin of safety as a result of implementing 
these requested amendments. Impact upon margin of safety is a 
consideration primarily included in the 10 CFR 50.59 evaluation 
process conducted for station procedures, procedure changes, and 
nuclear station modifications. The 10 CFR 50.59 evaluation process 
is conducted under the auspices of the Duke Qualified Reviewer 
Program and is not affected by these requested amendments. The 
impact on margin of safety for future Technical Specifications and 
Operating License changes will be reviewed by the PORC, but these 
reviews will be equivalent in quality to the reviews presently 
conducted by the Qualified Reviewers.
    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: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street, NW., Washington, DC 20036
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: January 27, 1995
    Description of amendment request: The requested change would modify 
Section 5.3.1, Fuel Assemblies, of the Waterford 3 technical 
specifications. The requested change increases the maximum enrichment 
for the spent fuel pool and containment temporary storage rack from 4.1 
to 4.9 weight percent U-235 when fuel assemblies contain fixed poisons. 
Waterford 3 plans to use higher enriched fuel in the next fuel cycle 
(Cycle 8) to meet the energy plans and maintain a reload batch size 
similar to that used in Cycles 6 and 7.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change will increase the fuel enrichment limit in 
order to meetthe cycle energy requirements while maintaining fuel 
batch sizes consistent with previous cycle designs. The calculated 
k-effective, including uncertainties, demonstrate substantial margin 
to criticality in the storage racks for both normal and accident 
conditions. No changes to the facility are required. No new modes of 
operating the fuel storage or transfer systems are required, except 
a restriction to limit the use of the new fuel vault to fuel with a 
maximum enrichment of 4.1 weight percent U-235. This restriction 
will be implemented by administrative controls. Since the plant 
equipment and operation are essentially the same, there is no 
significant increase in the probability of a criticality accident. 
Since a criticality event is demonstrated to be unfeasible, there 
are no increased adverse consequences for such a postulated event.
    As previously discussed, the proposed change will not result in 
a physical change to the facility nor will it result in a 
significant change to the operation of the facility; therefore, it 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed change has been analyzed to establish a k-
effective, including uncertainties, at or below the NRC criticality 
acceptance criteria of k-effective below 0.95 including 
uncertainties at the 95/95 probability/confidence level; therefore, 
there is no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: January 16, 1995
    Description of amendment request: The proposed amendment would 
revise the TMI-1 Technical Specifications (TS) to incorporate certain 
improvements from the Revised Standard Technical Specifications (TS) 
for Babcock & Wilcox nuclear power plants (NUREG-1430). The amendment 
would also change the bases incorporating the results of analyses to 
support allowance for drift of the pressurizer code safety valve 
setpoint. One of the proposed STS improvements involves a change to 
Chapter 6, Administrative Controls, affecting both TMI-1 and TMI-2 TSs. 
A separate notice of consideration of issuance of amendment to facility 
operating license is being issued for the proposed TMI-2 TSs Change.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. 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 proposed amendments involve a) an administrative change to 
both the TMI-1 and TMI-2 Technical Specifications which is 
consistent with the B&W Standard Technical Specifications (STS), 
NUREG-1430, and b) changes to the TMI-1 Technical Specifications 
which are consistent with the STS. This change does not involve any 
change to system or equipment configuration. The proposed amendment 
revises certain surveillance requirements, extends certain 
surveillance intervals as evaluated above, or involves changes that 
are purely
    administrative. The reliability of systems and components relied 
upon to prevent or mitigate the consequences of accidents previously 
evaluated is not degraded by the proposed changes. Assurance of 
system and equipment availability is maintained. 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 changes 
only involve changes to surveillance requirements that are 
consistent with STS and with the ASME Code. No new failure modes are 
created and thus the changes are bounded by accidents previously 
evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not [[Page 14022]] involve a significant reduction 
in a margin of safety. Each of these changes is compatible with the 
STS and has been evaluated to preserve the level of safety assured 
by the current TS.
    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: Government Publications 
Section, State Library of Pennsylvania, 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

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 20, 1995
    Description of amendment request: The proposed amendment would 
revise the fire hazards analysis for the River Bend Station (RBS) by 
allowing a deviation from 10 CFR 50, Appendix R, Section III.G.3 with 
respect to the requirement for a fixed fire suppression system in fire 
area C-17. This area houses the control building heating, ventilation 
and air conditioning (HVAC) systems and the loss due to a fire could 
cause the loss of main control room habitability. C-17 does not have a 
fixed fire suppression system but depends upon the use of the existing 
remote shutdown system as described in the updated safety analysis 
report (USAR).
    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 request does not involve a significant increase in the 
probability or consequences of accident previously evaluated.
    The event of concern is a fire in fire area C-17. The low fire 
loading and sparse concentration of exposed combustible material in 
fire area C-17 would limit fire spread. However, for this scenario 
all equipment in fire area C-17 will be assumed lost. Fire area C-17 
contains the air handling units for the main control room envelope. 
The loss of both air handling units would cause the control building 
chillers to stop running due to a logic tie requiring air flow 
through the air handling equipment for the chilled water system to 
operate during normal operation. The loss of the HVAC system in the 
control building would cause the main control room and the equipment 
rooms to begin heating up if exposed to design summer conditions. 
Operator actions can be accomplished to minimize the heat up rates 
for the rooms prior to the areas reaching equipment temperature 
limits. This would allow the operators to begin the shutdown process 
from the main control room. If the main control room continued to 
heat up, the operators could accomplish the shutdown using the 
remote shutdown system. HVAC for the remote shutdown panel is 
located in fire area C-4 and would not be damaged by a fire in fire 
area C-17. Operation of the control building HVAC system from the 
remote shutdown panel bypasses the logic between the chilled water 
system and the air handling system. This would allow restart of the 
HVAC system for all areas except the main control room. The scenario 
would conclude in a manner similar to that described in RBS USAR 
Appendix 15A, Event 52, ``Reactor Shutdown From Outside Main Control 
Room.''
    In summary, the probability of a fire occurring in fire area C-
17 is not increased. However, if a fire were to occur in fire area 
C-17 which caused the loss of main control room HVAC, the remote 
shutdown system would provide an acceptable method of shutdown. The 
low fire loading and sparse concentration of exposed combustible 
material in fire area C-17 would limit fire spread. Therefore, this 
request does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    2) The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    The event of concern is a fire in fire area C-17. Fire area C-17 
does not have a fixed suppression system as required by 10 CFR 50, 
Appendix R, Section III.G.3. Fire suppression systems are generally 
used to limit fire spread, once the heat of the fire opens thermally 
sensitive sprinklers. The low fire loading and sparse concentration 
of exposed combustible material in fire area C-17 would limit fire 
spread. However, for the purpose of event analysis, all equipment in 
fire area C-17 is assumed lost. Thus a fire in fire area C-17 is 
bounded by the same analysis with or
    without a fixed suppression system in terms of equipment 
availability.
    The proposed method of shutdown for a fire in fire area C-17 
will be changed in that the remote shutdown system will be credited. 
Use of the remote shutdown system is bounded by RBS USAR Appendix 
15A, Event 52, ``Reactor Shutdown From Outside Main Control Room.'' 
The HVAC for the remote shutdown panel is located in fire area C-4 
and would be undamaged by a fire in fire area C-17. Operation of the 
control building HVAC system from the remote shutdown panel bypasses 
the logic between the chilled water system and the air handling 
system. This would allow restart of the HVAC system for all areas 
except the main control room.
    In summary, if a fire were to occur in fire area C-17 which 
caused the loss of main control room HVAC, the remote shutdown 
system would provide an acceptable method of shutdown. Since, for 
the purpose of event analysis, all equipment in fire area C-17 is 
assumed lost, a fire in fire area C-17 is bounded by the same 
analysis with or without a fixed suppression system in terms of 
equipment availability. Therefore, this request does not create the 
possibility of occurrence of a new or different kind of accident 
from any accident previously evaluated.
    3) The request does not involve a significant reduction in a 
margin of safety.
    In this case, the margin of safety is implicit rather than being 
explicitly expressed as a numerical value. An implicit margin of 
safety involves conditions for NRC acceptance. Since the RBS 
Technical Specification Bases do not specifically address a margin 
of safety for fire protection, the SAR, the NRC's Safety Evaluation 
Report (SER), and appropriate other licensing basis documents were 
reviewed to determine if the proposed change would result in a 
reduction in a margin of safety. As stated, in part, in Attachment 4 
to NPF-47:
    EOI shall implement and maintain in effect all provisions of the 
approved fire protection program as described in the Final Safety 
Analysis Report for the facility through Amendment 22 and as 
approved in the SER dated May 1984 and Supplement 3 dated August 
1985 subject to provisions 2 and 3....
    As discussed in the Reason for Request, SSER 3 dated August 1985 
states, in part:
    On the basis of its evaluation the staff finds that the 
applicant's fire protection program with approved deviations is in 
conformance with the guidelines of BTP CMEB 9.5-1, sections III.G, 
III.J, and III.O of Appendix R to 10CFR50, and GDC 3, and is, 
therefore, acceptable.
    Thus, the margin of safety in this case can be defined as 
conformance with the specified fire protection guidelines. 10 CFR 
50, Appendix R, Section III.G.3, requires, in part, that alternative 
shutdown capability be provided for areas where adequate separation 
of redundant safe shutdown components cannot be provided. In 
addition, fire detection and a fixed fire suppression system must be 
installed in the area, room, or zone under consideration. Since fire 
area C-17 does not have a fixed suppression system, use of the 
remote shutdown system for a fire in this fire area would deviate 
from the requirements of 10 CFR 50, Appendix R, Section III.G.3. 
However, as discussed previously, the low fire loading and sparse 
amount of exposed combustibles compensate for the lack of a fixed 
fire suppression system. There is no adverse impact on the ability 
to achieve and maintain safe shutdown. Therefore, this request does 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
[[Page 14023]] amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: July 28, 1994
    Description of amendment request: The proposed amendment would add 
a footnote to Technical Specifcaiton 3.5.C. The footnote would state 
that the operability of the feedwater coolant injection (FWCI) system 
be independent of its seismic capability.
    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 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    Any postulated failure in the non-seismic portion of the FWCI 
subsystem may result in a loss of feedwater flow transient. However 
comparing the probability of occurrence of a seismic event, any 
increase in the probability of occurrence of a loss of feedwater 
event would be small. The proposed change would have no impact on 
the probability of occurrence of any other accident, including LOCAs 
[loss of coolant accidents].
    The FWCI subsystem will continue to be maintained as QA Category 
1 (except for the seismic attribute). Therefore, it will remain 
available for accident mitigation for most scenarios. Nevertheless, 
LOCA analyses have been reevaluated to demonstrate that FWCI is not 
necessary to show compliance with 10CFR50.46. Potentially limiting 
LOCA scenarios have been analyzed without the FWCI subsystem using 
an approved LOCA methodology. An active single failure was 
postulated in addition to not taking credit for the FWCI subsystem. 
Based on the results of these analyses, the current design basis 
large and small break LOCAs remain bounding. Moreover, FWCI is not 
credited in mitigating any of the non-LOCA transients/accidents.
    Safe shutdown following a seismic event can be achieved using 
the LPCI [low pressure coolant injection] and ESW [emergency service 
water] systems, and the SRVs [safety relief valves], which are all 
seismically qualified. Therefore, the FWCI system is not required to 
mitigate a seismic event.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Seismic reclassification of portions of FWCI does not create the 
possibility of a new kind of an accident. The portion of the piping 
up to the second isolation valve (from the RPV [reactor pressure 
vessel]), is seismically qualified and will remain classified as 
seismic. This ensures that a postulated failure in the non-seismic 
portion of piping or components does not degrade containment 
integrity or result in a blowdown of the RPV. Consequential and 
environmental effects of a FW [feedwater] piping failure have been 
analyzed in the HELB [high energy line break] program and have been 
found to be acceptable.
    3. Involve a significant reduction in the margin of safety.
    All accidents, including LOCAs, can be mitigated without using 
FWCI. FWCI is also not necessary for safe shutdown following a 
seismic event. The intended function of the FWCI subsystem is to 
reduce the likelihood of core uncovery during the lifetime of the 
plant. The CS [core spray] and LPCI subsystems provide redundant and 
diverse means of injecting water to the RPV. The FWCI subsystem 
provides an additional diverse means to inject water. Since FWCI 
will be maintained QA Category 1 (except for the seismic attribute), 
it will continue to provide the additional diversity to the 
injection systems. Considering the intended function of the 
subsystem and the credit taken in the accident analysis, 
reclassifying FWCI to be non-seismic does not significantly reduce 
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: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 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

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

    Date of amendment requests: January 9, 1995, as supplemented 
February 7, 1995
    Description of amendment requests: The proposed amendments would 
revise Prairie Island Nuclear Generating Plant Technical Specification 
(TS) 4.12, ``Steam Generator Tube Surveillance,'' to incorporate 
revised acceptance criteria for steam generator tubes with degradation 
in the tubesheet roll expansion region. These criteria for steam 
generator tube acceptance were developed by Westinghouse Electric 
Corporation and are known as F* (F-Star'') and L* 
(L-Star''). These criteria would be utilized to avoid 
unnecessary plugging and sleeving of steam generator tubes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    F* Steam Generator Tube Repair Criteria
    The supporting technical and safety evaluations of the subject 
criterion demonstrate that the presence of the tubesheet will 
enhance the tube integrity in the region of the hardroll by 
precluding tube deformation beyond its initial expanded outside 
diameter. The resistance to both tube rupture and tube collapse is 
strengthened by the presence of the tubesheet in that region. The 
results of hardrolling of the tube into the tubesheet is an 
interference fit between the tube and the tubesheet. Tube rupture 
cannot occur because the contact between the tube and tubesheet does 
not permit sufficient movement of tube material. The radial preload 
developed by the rolling process will secure a postulated separated 
tube end within the tubesheet during all plant conditions. In a 
similar manner, the tubesheet does not permit sufficient movement of 
tube material to permit buckling collapse of the tube during 
postulated LOCA [loss-of-coolant accident] loadings.
    The F* length of roll expansion is sufficient to preclude tube 
pullout from tube degradation located below the F* distance, 
regardless of the extent of the tube degradation. The existing 
Technical Specification leakage rate requirements and accident 
analysis assumptions remain unchanged in the unlikely event that 
significant leakage from this region does occur. As noted above, 
tube rupture and pullout is not expected for tubes using the F* 
criterion. Any leakage out of the tube from [[Page 14024]] within 
the tubesheet at any elevation in the tubesheet is fully bounded by 
the existing steam generator tube rupture analysis included in the 
Prairie Island Plant USAR [Updated Safety Analysis Report]. For 
plants with partial depth roll expansion like Prairie Island, a 
postulated tube separation within the tube near the top of the roll 
expansion (with subsequent limited tube axial displacement) would 
not be expected to result in coolant release rates equal to those 
assumed in the USAR for a steam generator tube rupture event due to 
the limited gap between the tube and tubesheet. The proposed 
plugging criterion does not adversely impact any other previously 
evaluated design basis accident.
    Leakage testing of roll expanded tubes indicates that for roll 
lengths approximately equal to the F* distance, any postulated 
faulted condition primary to secondary leakage from F* tubes would 
be insignificant.
    L* Steam Generator Tube Repair Criteria
    The presence of the tubesheet enhances steam generator tube 
integrity in the region of the hardroll by precluding tube 
deformation beyond its initial expanded outside diameter. The 
resistance to both tube rupture and tube collapse is strengthened by 
the presence of the tubesheet in that region. The result of the 
hardroll of the tube into the tubesheet is an interference fit 
between the tube and the tubesheet. Tube rupture cannot occur 
because the contact between the tube and tubesheet does not permit 
sufficient movement of tube materials. In a similar manner, the 
tubesheet does not permit sufficient movement of tube material to 
permit buckling collapse of the tube during postulated LOCA 
loadings.
    The type of degradation for which the L* criteria has been 
developed (cracking with an axial or near axial orientation) has 
been found not to significantly reduce the axial strength of a tube. 
An evaluation including analysis and testing has been done to 
determine the strength reduction for the axial loads with simulated 
axial and near axial cracks. This evaluation provided the basis for 
the acceptance criteria for tube degradation subject to the L* 
criteria.
    The length of roll expansion above L* is sufficient to preclude 
significant leakage from tube degradation located below the L* 
distance. The existing Technical Specification leakage rate 
requirements and accident analysis assumptions remain unchanged in 
the unlikely event that significant leakage from this region does 
occur. As noted above, tube rupture and pullout is not expected for 
tubes using the alternate plugging criteria.
    Any leakage out of the tube from within the tubesheet at any 
elevation in the tubesheet is fully bounded by the existing steam 
generator tube rupture analysis included in the Prairie Island 
Updated Safety Analysis Report. The proposed alternate plugging 
criteria do not adversely impact any other previously evaluated 
design basis accident.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    F*
    Implementation of the proposed F* criterion does not introduce 
any significant changes to the plant design basis. Use of the 
criterion does not provide a mechanism to initiate an accident 
outside of the region of the expanded portion of the tube. Any 
hypothetical accident as a result of any tube degradation in the 
expanded portion of the tube would be bounded by the existing tube 
rupture accident analysis. Tube bundle structural integrity will be 
maintained. Tube bundle leaktightness will be maintained such that 
any postulated accident leakage from F* tubes will be negligible 
with regards to offsite doses.
    L*
    Implementation of the proposed alternate tubesheet tube plugging 
criteria does not introduce changes to the plant design basis. Use 
of the criteria does not provide a mechanism to result in an 
accident outside of the region of the tubesheet expansion. Any 
hypothetical accident as a result of any tube degradation in the 
expanded portion of the tube would be bounded by the existing tube 
rupture accident analysis.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    F*
    The use of the F* criterion has been demonstrated to maintain 
the integrity of the tube bundle commensurate with the requirements 
of Reg Guide 1.121 [Bases for Plugging Degraded PWR Steam 
Generator Tubes] (intended for indications in the free 
span of tubes) and the primary to secondary pressure boundary under 
normal and postulated accident conditions. Acceptable tube 
degradation for the F* criterion is any degradation indication in 
the tubesheet region, more than the F* distance below the bottom of 
the transition between the roll expansion and the unexpanded tube. 
The safety factors used in the verification of the strength of the 
degraded tube are consistent with the safety factors in the ASME 
Boiler and Pressure Vessel Code used in steam generator design. The 
F* distance has been verified by testing to be greater than the 
length of roll expansion required to preclude both tube pullout and 
significant leakage during normal and postulated accident 
conditions. Resistance to tube pullout is based upon the primary to 
secondary pressure differential as it acts on the surface area of 
the tube, which includes the tube wall cross-section, in addition to 
the inner diameter based area of the tube. The leak testing 
acceptance criteria are based on the primary to secondary leakage 
limit in the Technical Specifications and the leakage assumptions 
used in the USAR accident analysis.
    Implementation of the tubesheet plugging criterion will decrease 
the number of tubes which must be taken out of service with tube 
plugs or repaired with sleeves. Both plugs and sleeves reduce the 
RCS (reactor coolant system) flow margin; thus, implementation of 
the F* criterion will maintain the margin of flow that would 
otherwise be reduced in the event of increased plugging or sleeving.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the USAR or the Technical Specification 
Bases.
    L*
    The use of the alternate tubesheet plugging criteria has been 
demonstrated to maintain the integrity of the tube bundle 
commensurate with the requirements of Reg. Guide 1.121 for 
indications in the free span of tubes and the primary to secondary 
pressure boundary under normal and postulated accident conditions. 
Acceptable tube degradation for the L* criteria is any degradation 
indication with axial or nearly axial cracking in the tubesheet 
region, more than the L* distance below the bottom of the transition 
between the roll expansion and the unexpended tube. For tubes with 
axial or nearly axial cracks the strength of the tube relative to an 
axial load would not be reduced below the strength required to 
resist potential axial loads. The safety factors used in the 
verification of the strength of the degraded tube are consistent 
with the safety factors in the ASME Boiler and Pressure Vessel Code 
used in steam generator design. The L* distance has been verified by 
testing to be greater than the length of roll expansion required to 
preclude significant leakage during normal and postulated accident 
conditions. The leak testing acceptance criteria are based on the 
primary to secondary leakage limit in the Technical Specifications 
and the leakage assumptions used in the USAR accident analyses.
    Implementation of the proposed tubesheet plugging criteria will 
decrease the number of tubes which must be taken out of service with 
tube plugs or repaired with sleeves. Both plugs and sleeves reduce 
the RCS flow margin, thus implementation of the alternate plugging 
criteria will maintain the margin of flow that would otherwise be 
reduced in the event of increased plugging or sleeving.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the Updated Safety Analysis Report or the 
bases of the Technical Specifications.
    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. This 
notice supersedes the staff's previous notice which was published in 
the Federal Register February 1, 1995 (60 FR 6307).
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Cynthia Carpenter, Acting [[Page 14025]] 

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

    Date of amendment requests: February 23, 1995
    Description of amendment requests: The proposed amendments would 
revise the wording in the Prairie Island technical specifications to 
allow implementation of exemptions to the schedule requirements of 10 
CFR Part 50, Appendix J. A related exemption request would grant 
temporary relief from the requirements of 10 CFR Part 50, Appendix J, 
Section III.D.1.(a) which requires Prairie Island Unit 2 to perform a 
Type A test in the May 1995 refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed amendment is an administrative change which allows 
implementation of approved exemptions to the regulations and by 
itself does not change any retest schedules.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by the proposed amendment.
    2. The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment is an administrative change which allows 
implementation of approved exemptions to the regulations and by 
itself does not change any retest schedules.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be created 
by the proposed amendment.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety
    The proposed amendment is an administrative change which allows 
implementation of approved exemptions to the regulations and by 
itself does not change any retest schedules.
    Therefore, a significant reduction in the margin of safety would 
not be involved with the proposed amendment.
    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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Cynthia Carpenter, Acting

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

    Date of amendment request: February 10, 1995
    Description of amendment request: The proposed amendment to the 
technical specifications (TSs) would relocate the requirements for the 
incore instrumentation (ICI) system from the TS to the Updated Safety 
Analysis Report (USAR).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed changes do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The Incore Instrumentation (ICI) System is used to measure core 
power distribution for the purpose of Limiting Conditions for 
Operation (LCO) monitoring of Technical Specification (TS) limits on 
linear heat rate, unrodded planer radial peaking factor, unrodded 
integrated radial peaking factor, and azimuthal power tilt. The ICI 
System has no safety purpose itself; it measures parameters which 
have safety significance. No change to the monitored parameters is 
proposed. The proposed changes will relocate requirements on the 
number and distribution of incore detectors used by the ICI System 
when measuring these parameters from the TS to the Updated Safety 
Analysis Report (USAR). Changes to the requirements can be made 
without NRC approval when the changes meet the criteria of 10 CFR 
50.59. Changes to the ICI System requirements that do not meet the 
criteria of 10 CFR 50.59 must be approved by the NRC by license 
amendment.
    Relocation of the requirements on the ICI System from the TS to 
the USAR does not increase the probability or consequences of any 
accident previously analyzed because the ICI System is neither a 
precursor nor a mitigator for any analyzed accident. The ICI System 
is used to ensure that operation within the LCOs for linear heat 
rate, unrodded planer radial peaking factor, unrodded integrated 
radial peaking factor, and azimuthal power tilt is maintained. 
However, its operation serves no mitigation function associated with 
any USAR Section 14 accident analysis. The parameters measured by 
the ICI System are important parameters in many accident analyses; 
however, this proposed change does not remove or revise the limits 
on these parameters.
    Additionally, it is proposed to revise TS 2.10.4(1)(b) to 
clarify its requirements. Currently TS 2.10.4(1) part (b) applies 
while operating under the provisions of part (a) if the plant 
computer incore detector alarms become inoperable. This is incorrect 
in that part (a) applies when the linear heat rate is being 
monitored by the ICI System and the linear heat rate is exceeding 
its limits as indicated by valid detector alarms. Part (b) of this 
specification should apply only if the linear heat rate is being 
monitored by the ICI System, is within its limits, and the plant 
computer incore detector alarms are inoperable.
    Administrative changes are also proposed which correct grammar 
and renumber/relocate portions of the TS and bases to other TS, to 
correspond to the proposed change to relocate ICI System 
requirements.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (2) The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The ICI System will continue to be used to monitor TS limits on 
core power distribution. There will be no physical alterations to 
the plant configuration, changes to setpoint values, or changes to 
the implementation of setpoints or limits as a result of this 
proposed change.
    The proposed change to TS 2.10.4(1)(b) only clarifies its 
requirements. The proposed change is more restrictive in that TS 
2.10.4(1)(b), as currently written, could be interpreted to allow 
continued operation for up to seven days with the linear heat rate 
exceeding its limits. The proposed change clarifies this 
specification to ensure that TS 2.10.4(1)(a) is applied if the 
linear heat rate is exceeded while being monitored by the ICI 
System. TS 2.10.4(1)(a) requires that the linear heat rate be 
restored within one hour or a plant shutdown initiated.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    (3) The proposed changes do not involve a significant reduction 
in a
    margin of safety.
    The ICI System is used to measure core power distribution 
parameters which are a direct measure of the margin of safety. The 
limits on these parameters are not changed. Therefore, the proposed 
change (i.e., relocation of the ICI System operability requirements 
to the USAR and/or plant procedures) does not involve a significant 
reduction in a margin of safety.
    The proposed change to TS 2.10.4(1)(b) helps ensure that the 
margin of safety is maintained by clarifying when the TS is 
applicable. This clarification ensures that the more restrictive 
actions of TS 2.10.4(1)(a) are taken if the linear heat rate is 
exceeded while being monitored by the ICI System. Therefore, the 
proposed change does not involve a significant reduction in a margin 
of safety. [[Page 14026]] 
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875 
Connecticut Avenue, NW., Washington, DC 20009-5728
    NRC Project Director: Theodore R. Quay

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

    Date of amendment requests: December 30, 1994 (Reference LAR 94-12)
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2, to revise TS 2.2, 3/4.3.1, 3/
4.3.2, 3/4.3.3, 3/4.4.4, 3/4.4.9, 3/4.5.2, 3/4.8.1, 3/4.8.2, 3/4.9.2, 
3/4.9.9, and 3/4.10.3. The specific TS changes proposed are as follows:
    (1) The TS issued in License Amendments (LAs) 84/83 would be 
changed to (a) revise the value of the overpower Delta-temperature 
(OPDT) constant K6 in TS 2.2.1, Table 2.2-1, Note 3; (b) revise the 
reactor coolant system (RCS) loop Delta-T function; and (c) make 
editorial corrections for clarification and consistency to TS 2.2.1 
(and TS 2.2.1 Bases), TS 3/4.3.1, and TS 3/4.3.2.
    In revising the RCS loop Delta-T function, the licensee would (a) 
incorporate the 0.99 multiplying factor listed in TS 2.2.1, Table 2.2-
1, Note 5, and TS 3/4.3.2, Table 3.3-4, Note 2, into constants B1 
through B4; (b) change ``Steam Generator (SG) Water Level Low-Low'' in 
TS 3/4.3.2, Table 3.3-3 and Table 4.3-2, Functional Unit 6.c, 
``Auxiliary Feedwater'' (AFW), by deleting the Mode 3 applicability of 
the RCS loop Delta-T function and by adding a footnote to the Mode 3 
applicability of the SG water level low-low function requiring that the 
trip time delay (TTD) associated with the SG water level low-low 
channel be less than or equal to 464.1 seconds; (c) change TS 3/4.3.1, 
Table 3.3-1, Action 27, and TS 3/4.3.2, Table 3.3-3, Action 29, by 
allowing up to four RCS loop Delta-T channels to be inoperable with the 
TTD threshold power level for zero seconds time adjusted to 0-percent 
rated thermal power (RTP) and by allowing the affected SG water level 
low-low channels to be placed in the tripped condition, with one 
inoperable RCS loop Delta-T channel; and (d) change the Table 3.3-1 and 
Table 3.3-3 ``Channels to Trip'' and ``Minimum Channels Operable'' 
columns to not applicable (N.A.).
    (2) The TS issued in LAs 70/69 would be changed to (a) delete 
references to the plant vent noble gas activity monitors (RM-14A and 
RM-14B) and footnote references to applicability of the containment 
ventilation exhaust radiation monitors (RM-44A and RM-44B) in TS Tables 
3.3-3, 3.3-4, 3.3-5, 3.3-6, 4.3-2, and 4.3-3 and TS 4.9.9; and (b) 
revise the ``Trip Setpoint and Allowable Values'' column in TS Table 
3.3-4, Functional Unit 3.c.4), to reference the offsite dose 
calculation procedure (ODCP).
    (3) Cycle-specific information in TS 4.3.2.1, TS 3.3.3.6, TS 
4.4.4.1, TS 4.5.2, TS 3.8.1.1, TS 3.8.2.1, and TS 3.8.2.2 that is no 
longer necessary would be deleted.
    (4) The word ``analog'' would be deleted from TS 4.4.9.3.1, TS 
4.9.2, and TS 4.10.3.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change to the OPDT constant K6 is conservative and 
will not cause any design or analysis acceptance criteria to be 
exceeded. There is no effect on the structural and functional 
integrity of any plant system. The OPDT function is part of the 
accident mitigation response and is not itself an initiator for any 
transient. This change does not affect the integrity of the fission 
product barriers for mitigation of radiological dose consequences as 
a result of an accident.
    The proposed change to incorporate the 0.99 multiplier into the 
TTD constants is an administrative change and has no effect on plant 
operation. The proposed change to delete Mode 3 applicability of the 
RCS Loop Delta-T function does not affect any design or analysis 
results. Allowing up to 4 RCS Loop Delta-T channels to be inoperable 
with the TTD threshold power level for zero seconds time delay 
adjusted to 0% RTP is conservative with respect to ESFs [engineered 
safety features] and reactor trip actuation time. Allowing the SG 
[steam generator] water level low-low channels affected by the 
inoperable RCS Loop Delta-T channels to be placed in the tripped 
condition is also conservative with respect to reactor trip and AFW 
pumps start. The change to the Channels to Trip and Minimum Channels 
Operable columns is a clarifying change to reflect the proposed 
changes to the action statements and identifies that the RCS Loop 
Delta-T does not provide a reactor trip function. Therefore, the 
proposed changes to the RCS Loop Delta-T function do not affect any 
of the accident analysis results.
    The proposed changes to revise Table 3.3-4, Functional Unit 
3.c.4), and to delete cycle-specific TS, TS references to RM-14A and 
RM-14B, and the word ``analog'' from the analog channel operation 
test are administrative and have no effect on plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change to the OPDT constant K6 does not affect the 
assumed accident initiation sequences. No new operating 
configuration is being imposed by the change to K6 that would create 
a new failure scenario. No new failure modes are being created for 
any plant equipment.
    The proposed changes to the RCS Loop Delta-T function do not 
involve any physical modification to any plant system or change the 
methodology by which any safety-related system performs its 
function.
    1The proposed changes to revise Table 3.3-4, Functional Unit 
3.c.4), and to delete cycle-specific TS, TS references to RM-14A and 
RM-14B, and the word ``analog'' from the analog channel operation 
test are administrative, would not result in any physical alteration 
to any plant system, and would not be a change in the method by 
which any safety-related system performs its function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    c. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change to the OPDT constant K6 will not affect any 
accident analysis assumptions, initial conditions, or results.
    The proposed changes to the RCS Loop Delta-T function do not 
affect any accident analysis assumptions, initial conditions, or 
results.
    The proposed changes to revise Table 3.3-4, Functional Unit 
3.c.4), and to delete cycle-specific TS, TS references to RM-14A and 
RM-14B, and the word ``analog'' from the analog channel operation 
test are administrative and clarify the TS. These proposed changes 
have no effect on current operating methodologies or actions that 
govern plant performance.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests [[Page 14027]] involve no significant hazards 
consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: Theodore R. Quay

Philadelphia Electric Company, Public Service Electric and Gas 
Company,Delmarva Power and Light Company, and Atlantic City 
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: September 26, 1994
    Description of amendment request: The proposed TS changes extend 
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.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed TS changes increase the STIs and AOTs for actuation 
instrumentation based on analyses described and justified in 
Licensing Topical Reports (References 2 through 8) [see licensee's 
September 26, 1994 application for reference information] which have 
been evaluated in associated Safety Evaluation Reports. These 
changes were incorporated into PBAPS Technical Specifications 
consistent with NUREG-1433. TS requirements that govern Operability 
or routine testing of plant instruments are not assumed to be 
initiators of any analyzed event because these instruments are 
intended to prevent, detect or mitigate accidents. Therefore, these 
changes will not involve an increase in the probability of 
occurrence of an accident previously evaluated. Additionally, these 
changes will not increase the consequences of an accident previously 
evaluated because the proposed change will not involve any physical 
changes to plant systems, structures, or components (SSC), or the 
manner in which these SSC are operated, maintained, modified, or 
inspected. The changes will not alter the operation of equipment 
assumed to be available for the mitigation of accidents or 
transients by the plant safety analysis or licensing basis. As 
justified in References 1 through 8, the proposed changes establish 
or maintain adequate assurance that components are operable when 
necessary for the prevention or mitigation of accidents or 
transients and that plant variables are maintained within limits 
necessary to satisfy the assumptions for initial conditions in the 
safety analyses. These changes establish or modify time limits 
allowed for operation with inoperable instrument channels based on 
the analyses in References 1 through 8 and will not allow continuous 
plant operation with plant conditions such that a single failure 
will result in a loss of any safety function. Therefore, these 
changes will not increase the consequences of an accident previously 
evaluated.
    2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    These proposed changes will not involve any physical changes to 
SSC, or the manner in which these SSC are operated, maintained, 
modified, tested, or inspected. Therefore, these changes will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. The changes in methods governing 
normal plant operation are consistent with the current safety 
analysis assumptions. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3) The proposed changes do not involve a significant reduction 
in a margin of safety.
    The proposed TS changes increase the STIs and AOTs for actuation 
instrumentation based on analyses described and justified in 
Licensing Topical Reports (References 2 through 8) which have been 
evaluated in associated Safety Evaluation Reports. These changes 
were incorporated into PBAPS Technical Specifications consistent 
with NUREG-1433. These changes can be classified into one of the 
following three categories:
    a. Changes to the minimum STIs and AOTs for the testing and/or 
repair of instrumentation based on the results of generic analyses 
in References 1 through 8;
    b. Changes to conditions, required actions, and completion times 
needed to make PBAPS TS requirements consistent with the assumptions 
used in the analyses in References 1 through 8; and,
    c. Changes that reformat, renumber, and/or reword existing 
requirements to incorporate the changes above.
    All of the proposed changes will be incorporated into the PBAPS 
custom Technical Specifications using the same approach and specific 
requirements used in Reference 12.
    There is no significant reduction in the margin of safety 
resulting from changes to the STIs and AOTs for the testing and/or 
repair of instrumentation based on the results of the analyses in 
References 1 through 8. These analyses determined that there is no 
significant change in the availability and/or reliability of 
instrumentation as a result of this change in STIs and AOTs. PECO 
Energy performed reviews that confirmed these analyses are 
applicable to PBAPS and that there would be no effect on the 
identification of excessive instrument setpoint drift as a result of 
increasing from monthly to quarterly the minimum interval between 
instrument functional tests. The proposed required actions ensure 
that actions to mitigate loss of single failure tolerance is 
initiated within 24 hours (12 hours for RPS) in accordance with the 
results of the analyses in References 1 through 8 and action to 
mitigate a loss of instrument function is initiated within 1 hour.
    The proposed changes which replace the shutdown actions 
associated with inoperable instrumentation with actions to declare 
the supported system inoperable does not involve a reduction in a 
margin of safety. The proposed changes ensure that appropriate 
compensatory measures are taken commensurate with approved TS 
Actions for the affected systems and the safety analyses. In 
addition, the proposed changes provide the benefit of avoiding an 
unnecessary shutdown transient when appropriate measures are 
available to compensate for the inoperable instrumentation.
    There is no significant reduction in the margin of safety 
resulting from changes that reformat, renumber, and/or reword 
existing requirements to incorporate the changes above.
    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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas 
Company,Delmarva Power and Light Company, and Atlantic City 
Electric Company,Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station,Units Nos. 2 and 3, York County, Pennsylvania

    Date of application for amendments: November 17, 1994
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) are being requested to support 
modifications 5384 and 5386 which upgrade the Main Stack and Vent Stack 
Radiation Monitoring Systems. [[Page 14028]] 
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Neither the Main Stack nor the Vent Stack Radiation Monitoring 
Systems serve as an initiator or contributor to any accidents 
previously evaluated. The systems provide indication and detection 
of radioactivity and effluent release in the main and vent stacks. 
The new systems perform the same function as the old, and have equal 
or better performance characteristics. Installation and operation of 
the new radiation monitoring systems do not degrade any active or 
passive equipment that responds to an accident.
    The proposed increase in the surveillance test interval of the 
subject radiation monitoring systems from 12 to 18 months is 
consistent with vendor recommendations, and is based on operating 
experience with instrumentation of a similar design.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    Both modifications replace obsolete radiation monitoring 
equipment and have the same failure modes as the existing equipment. 
The upgraded systems are considered enhancements to the existing 
systems and are considered neither a contributor nor initiator of 
any accidents previously evaluated.
    Based on the above, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    Neither the accuracy nor the responsiveness of the existing 
radiation monitoring equipment will be degraded as a result of the 
installation of modifications 5384 and 5386. Revisions to the 
calibration and surveillance frequencies are based on vendor 
information and experience with instrumentation of similar design. 
The changes associated with setpoints and the lower limit of 
detection are in the conservative direction. The upgraded main stack 
system continues to provide a non-safety related trip signal to 
Group III isolation valves during purging of the containment through 
the SBGTS [standby gas treatment system]. The revisions to parameter 
descriptions and instrument designation are considered 
administrative.
    Therefore, based on the above, 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: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

    Date of amendment request: February 3, 1994, supplemented September 
19, 1994, and November 23, 1994
    Description of amendment request: The proposed amendment revises 
the Technical Specifications to reflect a reduction in the Reactor 
Coolant System flow.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    No component modification, system realignment, or change in 
operations will occur which could affect the probability of any 
accident or transient. The proposed reduction in RCS loop and total 
flow rates will not change the probability of a challenge to any 
Engineered Safeguard Feature or other device. The consequences of 
previously analyzed accidents have been found to remain within 
acceptable licensing basis limits when the reduced flow rates are 
assumed. The system transient response is not affected by the 
initial RCS flow assumption, unless the initial assumption is so low 
as to impair the steady-state core cooling capability or steam 
generator heat transfer capability. This is clearly not the case 
with a 1% reduction in RCS flow. The proposed change to the wording 
of the parameter title on Table 3.2-1 is editorial for clarity. 
Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously analyzed.
    2. Create the possibility of a new or different kind of 
accident.
    No component modification, system realignment, or change in 
operating procedure will occur which could create the possibility of 
a new event not previously considered. The proposed reduction in RCS 
loop and total flow rates will not initiate any new events. 
Therefore, the proposed changes would not create the possibility of 
a different or new kind of accident.
    3. Involve a significant reduction in a margin of safety.
    The proposed decrease in RCS loop and total flow rates has been 
analyzed and found to have an insignificant effect on the applicable 
transient analyses found in the FSAR. The proposed change to the 
wording of the parameter title on Table 3.2-1 is editorial for 
clarity. Therefore, the proposed changes would not involve a 
significant reduction in any margin of safety.
    Therefore, based on the information presented above, PSE&G has 
concluded there is no significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

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

    Date of amendment request: January 30, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.6.1.2.a and associated Bases for 
3/4.6.1.2 to state that Type A tests for overall integrated containment 
leakage rate shall be conducted in accordance with the requirements 
specified in Appendix J of 10 CFR 50, as modified by NRC-approved 
exemptions. Additionally, TS 4.6.1.2.b would be revised to eliminate 
the reference to the schedule contained in TS 4.6.1.2.a.
    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:
    Toledo Edison has reviewed the proposed change and determined 
that a significant hazards consideration does not exist because 
operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in 
accordance with these changes would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because no accident initiators, 
[[Page 14029]] 
    conditions or assumptions are significantly affected by the 
proposed changes.
    The proposed change would revise Technical Specification (TS) 
Surveillance Requirement (SR) 4.6.1.2.a to allow overall integrated 
containment leakage rate (Type A) testing to be scheduled in 
accordance with 10 CFR 50 Appendix J, as modified by approved 
exemptions, and would make associated administrative changes to TS 
SR 4.6.1.2.b and to TS Bases 3/4.6.1.2. As stated above, none of 
these proposed changes involve accident initiators, conditions, or 
assumptions.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because no accident conditions or 
assumptions are affected by the proposed changes.
    The results of the previous Type A testing demonstrate a high 
degree of containment integrity. The Type B and C testing performed 
since the last Type A test provides confidence that the high degree 
of containment integrity will be maintained during the interval to 
the next Type A test. Therefore, the proposed changes do not alter 
the source term, containment isolation, or allowable releases, and 
will not increase the radiological consequences of a previously 
evaluated accident.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because no new or 
different accident initiators or assumptions are introduced by the 
proposed changes. The proposed changes do not affect the design or 
operation of any plant system, structure, or component. The proposed 
changes do not affect any accident initiators and are not initiators 
themselves. The proposed changes do not alter any accident 
scenarios.
    3. Not involve a significant reduction in a margin of safety. 
The initial conditions and methodologies used in the accident 
analyses remain unchanged. As described above, the proposed changes 
do not significantly reduce or adversely affect the confidence that 
the present high degree of containment integrity will be maintained.
    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 Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm

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

    Date of amendment request: January 30, 1995
    Description of amendment request: The proposed amendment would 
provide new Reactor Coolant Pressure Boundary (RCPB) pressure-
temperature limit curves that are applicable up to 21 effective full 
power years (EFPY).
    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:
    Toledo Edison had reviewed the proposed change and determined 
that a significant hazards consideration does not exist because 
operation of Davis-Besse Nuclear Power Station, Unit 1, in 
accordance with this change would:
    1a. Not involve a significant increase in the probability of an 
accident previously evaluated because: (1) revision of the pressure-
temperature curves and the extended applicability of the pressurizer 
level/RCS pressure limit curves for periods when relief valve DH4849 
is inoperable will continue to provide the same level of protection 
of the RCPB as was previously evaluated, and (2) the revision to 
License Condition 2.C(3)(d) is administrative to reflect the 
validity of the present analyses to 21 EFPY and (3) the revision to 
the Technical Specification Bases
    to reflect the extension to 21 EFPY is administrative and does 
not affect any previously analyzed accidents.
    1b. Not involve a significant increase in the consequences of an 
accident previously evaluated because: (1) revision of the pressure-
temperature curves and the extended applicability of the pressurizer 
level/RCS pressure limit curves for periods when relief valve DH4849 
is inoperable will continue to provide the same level of protection 
of the RCPB as was previously evaluated, and (2) the revision to 
License Condition 2.C(3)(d) is administrative to reflect the 
validity of the present analyses to 21 EFPY and (3) the revision to 
the Technical Specification Bases to reflect the extension to 21 
EFPY is administrative and does not affect any previously analyzed 
accidents.
    2. Not create the possibility of a new or different kind of 
accident from any accident previously evaluated because: (1) 
revision of the pressure-temperature curves and the extended 
applicability of the pressurizer level/RCS pressure limit curves 
will continue to provide protection against reactor vessel failure 
due to brittle fracture concerns under all postulated circumstances, 
and (2) the revision to License Condition 2.C(3)(d) is 
administrative to reflect the validity of the present analyses to 21 
EFPY and (3) the revision to the Technical Specification Bases to 
reflect the extension to 21 EFPY is an administrative change and 
does not affect any activities or equipment in plant operation.
    3. Not involve a significant reduction in a margin of safety 
because: (1) revision of the pressure-temperature curves and the 
extended applicability of the pressurizer level/RCS pressure limit 
curves maintains the present margin of safety from reactor vessel 
brittle fracture as required by 10 CFR 50, Appendix G, and (2) the 
revision to License Condition 2.C(3)(d) and the Bases revision are 
administrative and do not affect any analyses which provide the 
basis for the Technical Specifications.
    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 Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia

    Date of amendment request: February 14, 1995
    Description of amendment request: The proposed change revises 
Technical Specification 4.4.D to reference the testing requirements of 
10 CFR Part 50, Appendix J, and to state that the Nuclear Regulatory 
Commission-approved exemptions to the applicable regulatory 
requirements are permitted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Virginia Electric and Power Company has performed an evaluation 
of ... the proposed administrative Technical Specification change, 
in accordance with 10 CFR 50.91(a)(1) regarding no significant 
hazards considerations using the standards in 10 CFR 50.92(c). A 
discussion of these standards as they relate to this ... amendment 
request follows.
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The proposed change ... revises Technical Specification 4.4.D to 
reference the testing frequency requirements of 10 CFR 50 Appendix J 
and to state that NRC approved exemptions to the applicable 
regulatory [[Page 14030]] requirements are permitted. The current 
Technical Specification requires retests in accordance with Section 
III.D.1(a) of Appendix J. The proposed administrative change simply 
includes the statement ``as modified by NRC approved exemptions.'' 
No new requirements are added, nor are any existing requirements 
deleted. Any specific changes to the requirements of Section 
III.D.1(a) will require a submittal from Virginia Electric and Power 
Company under 10 CFR 50.12 and subsequent review and approval by the 
NRC prior to implementation. The proposed change is stated 
generically to avoid the need for further Technical Specification 
changes if different exemptions are approved in the future.
    The proposed change, in itself, does not affect reactor 
operations or accident analyses and has no radiological 
consequences. The change provides clarification so that future 
Technical Specifications changes will not be necessary to correspond 
to applicable NRC approved exemptions from the requirements of 
Appendix J.
    Therefore, this proposed change does not involve a significant 
increase in the probability or consequences of any accident 
previously evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    The proposed Technical Specification amendment provides 
clarification to a specification that paraphrases a codified 
requirement.
    Since the proposed change would not change the design, 
configuration or method of operation of the plant, it would not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed Technical Specification change is administrative 
and clarifies the relationship between the requirements of TS 4.4.D, 
Appendix J, and any approved exemptions to Appendix J. It does not, 
in itself, change a safety limit or [a] Limiting Condition for 
Operation. The NRC will directly approve any proposed change or 
exemption to III.D.1(a) of Appendix J prior to implementation.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, 951 E. Byrd Street, Richmond, Virginia 
23219.
    NRC Project Director: David B. Matthews

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

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

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of amendment request: February 14, 1995
    Brief description of amendment request: The amendment request 
proposes changes to Technical Specification 3.8.2, ``AC Sources-
Shutdown;'' 3.8.5, ``DC Sources-Shutdown;'' and 3.8.8, ``Inverters-
Shutdown.'' The proposed changes would revise the operability 
requirements for the Division 3 diesel generator and the Division 3 and 
4 batteries, battery chargers, and inverters to apply only when the 
high pressure core spray system is required to be operable.Date of 
publication of individual notice in Federal Register: February 17, 1995 
(60 FR 9412).
    Expiration date of individual notice: March 20, 1995
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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

    Date of amendment request: September 8, 1994
    Brief description of amendment request: The amendment request 
proposes changes to Technical Specification Section 3/4.9.1 to 
establish administrative controls to address a possible boron dilution 
event directly from the reactor makeup water system.
    Date of publication of individual notice in Federal Register: March 
1, 1995 (60 FR 11151).
    Expiration date of individual notice: March 31, 1995
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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 No. STN 50-529, Palo 
Verde Nuclear Generating Station, Unit 2, Maricopa County, Arizona

    Date of application for amendment: November 30, 1994, as 
supplemented by letter dated January 27, 1995
    Brief description of amendment: The amendment changed the 
pressurizer code safety valve lift setting from 2500 
[[Page 14031]] psia to 2475 psia. The lift setting is being changed to 
permit Unit 2 to operate with up to 1500 plugged tubes in each steam 
generator.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995
    Amendment No.: 78
    Facility Operating License No. NPF-74: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
496) The additional information contained in the January 27, 1995, 
supplemental letter was clarifying in nature and thus within the scope 
of the initial notice and did not affect the NRC staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 1, 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: September 6, 1994
    Brief description of amendment: The proposed amendment relocates 
the alarms for the drywell to suppression chamber vacuum breaker to a 
different annunicator panel.
    Date of issuance: February 16, 1995 Effective date: To be 
implemented prior to startup from refueling outage 10.
    Amendment No.: 158
    Facility Operating License No. DPR-35: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53839) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated No significant hazards 
consideration comments received: No
    Local Public Document Room location: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.

Commonwealth Edison Company, Docket Nos. STN 50-456 and STN 50-457, 
Braidwood Station, Unit Nos. 1 and 2, Will County, Illinois

    Date of application for amendments: January 5, 1994, as 
supplemented by letters dated April 26, 1994, September 30, 1994, and 
January 12, 1995.
    Brief description of amendments: The amendments change the 
Braidwood Technical Specifications to remove the requirement to verify, 
every 18 months, that the control room ventilation can be manually 
isolated.
    Date of issuance: February 28, 1995
    Effective date: February 28, 1995
    Amendment Nos.: 60 and 60
    Facility Operating License Nos. NPF-72 and NPF-77: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 25, 1995 (60 FR 
4930). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 28, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Wilmington Township Public 
Library, 201 S. Kankakee Street, Wilmington, Illinois 60481.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, IllinoisDocket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, 
Will County, Illinois

    Date of application for amendments: August 31, 1993, as 
supplemented July 19, 1994.
    Brief description of amendments: The amendments revise the 
technical specifications by increasing the allowed outage time for an 
inoperable chiller only in MODES 1 through 4, adding an optional ACTION 
statement in MODES 5 and 6, and adding a surveillance requirement for 
the control room ventilation system.
    Date of issuance: March 2, 1995
    Effective date: March 2, 1995
    Amendment Nos.: 70, 70, 61 and 61
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 25, 1995 (60 FR 
4932). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 2, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481.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, IllinoisDate 
of application for amendments: July 29, 1992, as supplemented January 
14, 1993, and February 16, 1993
    Brief description of amendments: Dresden and Quad Cities Technical 
Specification Upgrade Program. Date of issuance: February 16, 
1995Effective date: Immediately, to be implemented by December 31, 
1995.
    Amendment Nos.:  131, 125, 152, and 148
    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: June 23, 1993 (58 FR 
34071) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 16, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: For Dresden, The Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450; For Quad Cities, 
The Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 61021.

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

    Date of application for amendments: July 29, 1992, as supplemented 
January 14, 1993, February 16, 1993 and January 27, 1995
    Brief description of amendments: The July 29, 1992, application, is 
one of twelve applications which have been submitted by Commonwealth 
Edison Company (ComEd) in an effort to upgrade the existing custom 
Technical Specifications (TS) to the Boiling Water Reactor (BWR) 
Standard Technical Specifications (STS). Dresden has recently 
rescheduled the Unit 2 refueling outage from March 4, 1995, until June 
1995. Currently, the surveillance frequency for certain Inservice 
Testing (IST) requirements expires on February 21, 1995. The current 
TSs do not make provisions for a grace period for surveillance 
frequencies of the IST program. In accordance with BWR STS guidance, 
the TSs regarding IST proposed in the July 29, 1992, application, allow 
the flexibility to perform these tests appropriately during refueling 
outages (where applicable) by providing a 25 percent extension to IST 
surveillance intervals. The January 27, 1995, supplement requested the 
staff to review and approve just that portion of the July 29, 1992, 
application dealing with the implementation of the IST program in 
Section 3.0/4.0 of the proposed TS. [[Page 14032]] 
    Date of issuance: February 22, 1995Effective date: February 22, 
1995
    Amendment Nos.:  132 and 126
    Facility Operating License Nos. DPR-19 and DPR-25: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34071) The January 27, 1995, letter 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 datedFebruary 22, 1995.No significant hazards consideration 
comments received: No
    Local Public Document Room location: Morris Public Library, 604 
Liberty Street, Morris, Illinois 60450.

Connecticut Yankee Atomic Power Company and Northeast Nuclear 
Energy Company, Docket Nos. 50-213 and 50-245, Haddam Neck Plant 
and Millstone Nuclear Power Station, Unit 1, Middlesex County and 
New London County, Connecticut

    Date of application for amendments: October 31, 1994, as 
supplemented February 14, 1995.
    Brief description of amendments: The amendments renew the existing 
license conditions for both plants to implement and maintain Integrated 
Implementation Schedule Program Plans (the Program Plans). The Program 
Plans provide a methodology to be followed for scheduling plant 
modifications and engineering evaluations.
    Date of issuance: February 23, 1995
    Effective date:  February 23, 1995
    Amendment Nos.:  183 for Haddam Neck, 80 for Millstone 1
    Facility Operating License Nos. DPR-61 and DPR-21. Amendments 
revise the Licenses.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63117)The February 14, 1995, letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated February 23, 
1995.No significant hazards consideration comments received: No.
    Local Public Document Room locations: Russell Library, 123 Broad 
Street, Middletown, CT 06457, for the Haddam Neck Plant, and the 
Learning Resource Center, Three Rivers Community-Technical College, 
Thames Valley Campus, 574 New London Turnpike, Norwich, CT 06360, for 
Millstone Unit 1.

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

    Date of application for amendment: October 5, 1994, as supplemented 
February 10, 20, and 22, 1995.
    Brief description of amendment: The amendment revises primary 
coolant system (PCS) pressure-temperature limits, power-operated relief 
valve setting limits, and primary coolant pump starting limits to 
accommodate reactor vessel fluence for an additional 4 effective full 
power years. The amendment also revises the emergency core cooling 
system technical specifications to render two high-pressure safety 
injection pumps incapable of injecting into the PCS when the PCS is 
below 300 deg.F rather than rendering both inoperable below 260 deg.F. 
In addition, it revises the pressurizer heatup to achieve consistency 
between design assumptions and technical specifications limits.
    Date of issuance: March 2, 1995
    Effective date: March 2, 1995
    Amendment No.: 163
    Facility Operating License No. DPR-20. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
501) The February 10, 20, and 22, 1995, submittals provided 
clarifyinginformation which was within the scope of the initial 
application and did not affect the staff's initial proposed no 
significant hazards consideration findings. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
March 2, 1995.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423.

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

    Date of application for amendments: January 10, 1994, as 
supplemented March 21 and September 15, 1994, and January 5, 1995
    Brief description of amendments: The amendments revised Technical 
Specification Table 2.2-1 and TS 4.2.5 to allow a change in the method 
for measuring reactor coolant system (RCS) flow rate from the 
calorimetric heat balance method to a method based on a one-time 
calibration of the RCS cold leg elbow differential pressure taps.
    Date of issuance: February 17, 1995
    Effective date: To be implemented within 30 days from the date of 
issuance
    Amendment Nos.: 128 and 122
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 26, 1994 (59 FR 
3743) for Unit 1; and March 1, 1994 (59 FR 9785) for Unit 2
    The March 21 and September 15, 1994, and January 5, 1995, letters 
provided additional information that did not change the initial scope 
of the January 10, 1994, application and the initial proposed no 
significant hazards consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 17, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of amendment request: August 30, 1994
    Brief description of amendment: The amendment revised the Technical 
Specifications to address the installation of two battery chargers on 
each 125 vdc power train in lieu of the ``swing'' battery charger that 
is currently used.
    Date of issuance: February 17, 1995
    Effective date: February 17, 1995
    Amendment No.: 176
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 17, 1995 (60 FR 
3439) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 17, 1995.No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801

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

    Date of amendment request: June 22, 1994.
    Brief description of amendment: The amendment extends the allowable 
outage time for one inoperable train of emergency feedwater from 36 
hours to 72 hours, clarifies the specifications and their associated 
bases, and relocates information within the specifications.
    Date of issuance: March 1, 1995
    Effective date: 30 days following the date of 
issuance. [[Page 14033]] 
    Amendment No.: 177
    Facility Operating License No. DPR-51. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994, (59 FR 
42339) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801

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

    Date of amendment request: January 19, 1995
    Brief description of amendment: The amendment changed the Appendix 
A technical specifications (TSs) by adding TS 3.0.5 and its associated 
Bases. This new specification will allow equipment removed from service 
or declared inoperable to comply with ACTIONS to be returned to service 
under administrative controls soley to perform testing required to 
demonstrate its OPERABILITY or the OPERABILITY of other equipment.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995
    Amendment No.: 101
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 27, 1995 (60 FR 
5441) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated March 1, 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.

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

    Date of amendment request: August 11, 1994, as supplemented by 
letter dated December 2, 1994.
    Brief description of amendment: The amendment revised the Technical 
Specifications for the Waterford Steam Electric Station, Unit 3, by 
modifying the specifications having cycle-specific parameter limits by 
replacing the values of those limits with a reference to a core 
operating limits report for the values of those limits. These changes 
are in accordance with the requirements of Generic Letter 88-16.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995
    Amendment No.: 102
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65812) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 1, 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.

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

    Date of amendment request: August 19, 1994, as supplemented by 
letter dated October 14, 1994.
    Brief description of amendment: The amendment changed the Appendix 
A technical specification (TSs) by removing the Limiting Condition For 
Operation (LCO) 3/4.3.4, the associated surveillance requirements, and 
Bases information from the TSs. This information and requirements will 
be incorporated into the Waterford 3 Updated Final Safety Analysis 
Report (UFSAR) and maintained under the provisions of 10 CFR 50.59.
    Date of issuance: March 2, 1995
    Effective date: March 2, 1995
    Amendment No.: 103
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45023) The additional information contained in the supplemental letter 
dated October 14, 1994, was clarifying in nature and thus, within the 
scope of the initial notice and did not affect the staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated March 2, 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.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: April 21, 1993
    Brief description of amendment: The amendment revised the 
requirement for control rod testing to increase the ``notch testing'' 
surveillance interval for partially withdrawn control rods from once 
per 7 days to once per 31 days. The change is consistent with the 
format and content of the Improved Standard Technical Specifications 
(NUREG-1434, Revision 0).
    Date of issuance: February 16, 1995
    Effective date: February 16, 1995
    Amendment No: 115
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1993 (58 FR 
28055) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 16, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: July 14, 1993
    Brief description of amendment: The amendment revised technical 
specification requirements for the hydrogen ignition system (HIS). The 
amendment also removed several tables related to the HIS in accordance 
with guidance contained in Generic Letter 91-08, ``Removal of Component 
Lists From Technical Specifications.''
    Date of issuance: February 16, 1995
    Effective date: February 16, 1995
    Amendment No: 116
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46232) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 16, 1995. No 
significant hazards consideration comments received: No [[Page 14034]] 
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce at Washington, Natchez, Mississippi 39120.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: August 11, 1993
    Brief description of amendment: The amendment deleted the 
requirements of Limiting Condition for Operation (LCO) 3.3.3.9 and 
Surveillance Requirement 4.3.3.9 related to loose-part detection 
instrumentation. The deleted requirements will be relocated to 
documents that are controlled by the licensee under the provisions of 
10 CFR 50.59. The change is consistent with the format and content of 
the Improved Standard Technical Specifications (NUREG-1434, Revision 
0).
    Date of issuance: February 16, 1995
    Effective date: February 16, 1995
    Amendment No: 117
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46232) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 16, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company,Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: August 11, 1993
    Brief description of amendment: The amendment deleted certain 
accident monitoring instruments from Technical Specification Table 
3.3.7.5-1 ``Accident Monitoring Instrumentation'' and deleted the 
corresponding Surveillance Requirements from Table 4.3.7.5-1, 
``Accident Monitoring Instrumentation Surveillance Requirements.'' The 
deleted requirements will be relocated to documents that are controlled 
by the licensee under the provisions of 10 CFR 50.59. The change is 
consistent with the format and content of the Improved Standard 
Technical Specifications (NUREG-1434, Revision 0).
    Date of issuance: February 16, 1995
    Effective date: February 16, 1995
    Amendment No: 118
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46234) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 16, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: October 22, 1993, as 
supplemented by letters dated February 10, and 14, 1995.
    Brief description of amendment: The amendment modified the testing 
frequencies for the drywell bypass test and the airlock test, relocated 
certain drywell airlock tests from the technical specifications to 
administrative procedures, and incorporates various improvements from 
the Improved Standard Technical Specifications (NUREG-1434, Revision 
0).
    Date of issuance: February 16, 1995
    Effective date: February 16, 1995
    Amendment No: 119
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1993 (58 FR 
64607) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 16, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, Mississippi 39120.

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

    Date of application for amendment: May 17, 1993, as supplemented on 
December 23, 1994
    Brief description of amendment: The amendment changes the action 
statement for inoperable degraded grid and loss of voltage relays and 
their associated auxiliary relays and timers.
    Date of issuance: January 31, 1995
    Effective date: January 31, 1995
    Amendment No.: 193
    Facility Operating License No. DPR-50. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59750). The December 23, 1994, letter provided additional 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of this amendment is contained in a Safety Evaluation dated 
January 31, 1995. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, PA 17105. The above Notice was to be 
published in the Federal Register of February 15, 1995. The notice that 
was inadvertently published at 60 FR 8762 relates to a licensing action 
which has not been completed.

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: November 7, 1994, as supplemented by 
letters dated December 20, 1994, and January 23, 1995.
    Brief description of amendments: The amendments changed the number 
of standby diesel generators (SDGs) (emergency power source) required 
to be operable during Mode 6 with greater than or equal to 23 feet of 
water above the reactor vessel flange, from two to one. The amendment 
also allows limited substitution of an alternate onsite emergency power 
source for one of the two required SDGs, in Mode 5, and in Mode 6 with 
less than 23 feet of water. In addition, certain system specifications 
that are affected by the changes for the emergency power source were 
also changed.
    Date of issuance: February 14, 1995
    Effective date: February 14, 1995, to be implemented within 31 days 
of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 34; Unit 2 - Amendment No. 
20
    Facility Operating License Nos. NPF-76 and NPF-80. The amendments 
revised the Technical [[Page 14035]] Specifications.Public comments 
requested as to proposed no significant hazards consideration: Yes (60 
FR 5739, dated January 30, 1995). The notice provided an opportunity to 
submit comments on the Commission's proposed no significant hazards 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by March 1, 1995, 
but stated that, if the Commission makes a final no significant hazards 
consideration determination, any such hearing would take place after 
issuance of the amendments.
    The Commission's related evaluation of the amendments, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration is contained in a Safety Evaluation dated 
February 14, 1995.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, Texas 77488

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

    Date of application for amendment: August 15, 1994, as supplemented 
on December 21, 1994, and January 20, 1995. The licensee's submittals 
of December 21, 1994, and January 20, 1995, provided clarification and 
did not change the original no significant hazards consideration.
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications by increasing the allowable main steam 
isolation valve (MSIV) leakage and deleting the requirements applicable 
to the MSIV leakage control system.
    Date of issuance: February 22, 1995
    Effective date: February 22, 1995 and to be implemented within 90 
days.
    Amendment No.: 207
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47169) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 22, 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.

Illinois Power Company and Soyland Power Cooperative, Inc., Docket 
No. 50-461, Clinton Power Station, Unit No. 1, DeWitt County, 
Illinois

    Date of application for amendment: August 12, 1994, as supplemented 
on October 14, 1994 and February 6, 1995.
    Brief description of amendment: The amendment modifies Clinton 
Power Station Technical Specification 3.6.5.1, ``Drywell,'' to permit a 
one-time only change to forego performance of the drywell bypass 
leakage rate test during the fifth refueling outage scheduled to begin 
in March 1995.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995
    Amendment No.: 96
    Facility Operating License No. NPF-62. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49427). The October 14, 1994, and February 6, 1995, submittals 
consisted of revisions and clarifications which did not change the 
staff's initial proposed no significant hazards consideration 
determination or expand the scope of the original notice.The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated March 1, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location:  The Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727.

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

    Date of application for amendments:  November 18, 1994
    Brief description of amendments: The amendments revise Technical 
Specification 4.0.5 to delete the wording ``except where specific 
written relief has been granted by the Commission pursuant to 10 CFR 
50, Section 50.55a(g)(6)(i).'' This change allows the licensee to 
implement certain 10 CFR 50.55a relief requests while the relief 
requests are being reviewed by the NRC at the beginning of an updated 
interval.
    Date of issuance: February 23, 1995
    Effective date: February 23, 1995
    Amendment Nos.: 190/176
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65817) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 23, 1995. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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: May 18, 1994
    Brief description of amendment: The amendment modifies the 
operability requirements for the fuel building exhaust filter system. 
The amendment will result in modifications to the applicability, 
surveillance requirement, and bases sections of Technical Specification 
3/4.9.12, ``Fuel Building Exhaust Filter System.''
    Date of issuance: February 22, 1995
    Effective date: As of the date of issuance to be implemented 
within30 days.
    Amendment No.: 105
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: June 22, 1994 (59 FR 
32234) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 22, 1995.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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

    Date of application for amendments: July 9, 1992
    Brief description of amendments: The amendments extend the 
operating licenses for the Diablo Canyon Nuclear Power Plant, Units 1 
and 2 to recover or recapture the construction period of the reactors. 
Specifically, the amendments extend the expiration date of the Unit 1 
license from April 23, 2008, to September 22, 2021, and the expiration 
date of the Unit 2 license from December 9, 2010, to April 26, 2025.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995
    Amendment Nos.: 97 and 96
    Facility Operating License Nos. DPR-80 and DPR-82: The amendments 
revised the license.
    Date of initial notice in Federal Register: July 22, 1992 (57 FR 
32575) [[Page 14036]] The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated March 1, 1995.No 
significant hazards consideration comments received: Yes. Comments from 
the San Luis Obispo Mothers for Peace (MFP) and their contentions were 
admitted into this proceeding. These contentions concern the adequacy 
of the licensee's maintenance and surveillance program and interim 
corrective actions in lieu of Thermo-Lag. The Atomic Safety and 
Licensing Board, in its initial decision dated November 4, 1994 (LBP-
94-35), authorized the staff to extend the DCPP operating license 
expiration dates. Because a hearing was held prior to license issuance, 
the staff does not need to make a final no significant hazards 
consideration determination.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407

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

    Date of application for amendment: July 27, 1994, as supplemented 
October 27, 1994 and February 3, 1995
    Brief description of amendment: The amendment raises the authorized 
Power Level from 3293 MWt to a new limit of 3441 MWt.
    Date of issuance: February 22, 1995
    Effective date: As of date of issuance and is to be implemented 
prior to startup in Cycle 9, currently scheduled to occur in May 1995.
    Amendment No.: 143
    Facility Operating License No. NPF-14: This amendment revised the 
Technical Specifications and license.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47171) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 22, 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.

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: June 23, 1994
    Brief description of amendments: The amendment revises Technical 
Specification 4.0.5, which provides the requirements for inservice 
inspection and testing of ASME Code components, to conform to Standard 
Technical Specifications (NUREG-1433).
    Date of issuance: February 28, 1995
    Effective date: February 28, 1995
    Amendment Nos.: 144 and 113
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39595) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 28, 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.

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

    Date of application for amendments: October 28, 1994, and 
supplemented by letter dated December 29, 1994
    Brief description of amendments: These amendments change the 
Technical Specifications (TS) for the two units by adding reference 
20 (Unit 1) and reference 18 (Unit 2) to Section 
6.9.3.2 as ``PL-NF-90-001, Supplement 1, 'Application of Reactor 
Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating 
Changes and Use of RETRAN MOD 5.1,' September 1994.'' These additions 
reflect changes to the methodology that the licensee is using to 
perform its nuclear fuel reload analysis for the two units.
    Date of issuance: February 28, 1995
    Effective date: February 28, 1995
    Amendment Nos.: 145 and 114
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65819) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated February 28, 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 31, 1994
    Brief description of amendments: These amendments address Section 
5, ``Remove Temperature Requirement for Operational Condition 5 (TSCR 
94-44-0), by revising TS Table 1.2 and TS Bases 3/4.9.11 to remove the 
average reactor coolant temperature requirement in Operational 
Condition (OPCON) 5, Refueling.
    Date of issuance: January 27, 1995
    Effective date: January 27, 1995Amendment Nos. 88 and 50
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55884) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 27, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Power Authority of The State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment:  November 16, 1994
    Brief description of amendment: The amendment revises Technical 
Specifications Section 3.10.8 and the associated Bases, to reduce the 
maximum allowable control rod drop time from 2.4 to 1.8 seconds.
    Date of issuance: February 21, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 160
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 20, 1995 (60 FR 
4203) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 21, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610. [[Page 14037]] 

Saxton Nuclear Experimental Corporation, Docket No. 50-146, Saxton 
Nuclear Reactor Facility

    Date of application for amendment: August 8, 1994, as supplemented 
on October 28, 1994, and January 12, 1995.
    Brief description of amendment: The amendment adds characterization 
as an authorized activity at Saxton and improves the wording of the 
technical specifications.
    Date of issuance: February 22, 1995
    Effective date: February 22, 1995
    Amendment No.: 12Amended Facility License No. DPR-4: Amendment 
changed the Technical Specifications
    Date of initial notice in Federal Register: November 9, 1994. The 
Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated February 22, 1995.No significant hazards 
consideration comments received: No
    Local Public Document Room location: Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678

Southern California Edison Company, et al., Docket Nos. 50-361 and 
50-362, San Onofre Nuclear Generating Station, Unit Nos. 2 and 3, 
San Diego County, California

    Date of application for amendments: December 30, 1993, as 
supplemented by letters dated June 3, 1994, August 25, 1994, and 
January 3, 19, and 30, 1995.
    Brief description of amendments: These amendments revise TS 3.9.4, 
``Containment Building Penetrations,'' and the associated bases to 
allow both doors of the containment personnel airlock to be open at the 
same time during refueling operations provided certain conditions are 
met.
    Date of issuance: February 28, 1995
    Effective date: February 28, 1995
    Amendment Nos.: 117 and 106
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49434). The additional information contained in the January 3, 19, 
and 30, 1995, letters were clarifying in nature, within the scope of 
the initial notice and did not affect the NRC staff's proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated February 28, 1995.No significant hazards consideration 
comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713

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

    Date of amendments request: December 19, 1994
    Brief description of amendments: The amendments to Technical 
Specifications include: (1) a revision in Table 3.7-3 to the main steam 
safety valve (MSSV) setpoint tolerance from plus or minus 1 percent to 
plus or minus 3 percent, (2) modification of the bases to 3/4.7.1.1 to 
increase the relieving capacity of the MSSVs to at least 12,984,660 
pounds per hour which corresponds to approximately 112 percent of total 
secondary steam flow at 100 percent rated thermal power, (3) 
modifications to Table 3.7-1 to reduce the allowable power range 
neutron flux high setpoints for multiple inoperable steam generator 
safety valves, and (4) an editorial correction to Bases 3/4.7.1.2 to 
indicate required auxiliary feedwater flow at ``1133 psia'' rather than 
``1133 psig.''
    Date of issuance March 1, 1995
    Effective date: March 1, 1995
    Amendment Nos.: 112 and 103
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
505) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated March 1, 1995No 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 36302The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated March 1, 1995No significant hazards 
consideration comments received: No

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: September 29, 1993
    Brief description of amendment: The proposed changes increase the 
amount of boron required in the standby liquid control system.
    Date of issuance: February 28, 1995
    Effective date: February 28, 1995
    Amendment Nos.: 217, 233 and 191
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: June 8, 1994 (59 FR 
29635) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 28, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket Nos. 50-259, 50-260 and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: September 30, 1993 (TS 336)
    Brief description of amendment: The proposed changes revise and 
clarify the spent fuel pool water level, temperature, sampling, and 
analysis surveillance requirements.
    Date of issuance: March 2, 1995
    Effective date: March 2, 1995
    Amendment Nos.: 218, 334 and 192
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 22, 1993 (58 
FR 67862) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated March 2, 1995.No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket Nos. 50-259, 50-260, and 50-296, 
Browns Ferry Nuclear Plant, Units 1, 2, and 3, Limestone County, 
Alabama

    Date of application for amendments: March 31, 1994
    Brief description of amendment: For Browns Ferry Units 1 and 3, the 
proposed changes provide for operation in the extended load line limit 
region and revised rod block monitor operability requirements. For all 
three Browns Ferry units, the changes delete a obsolete value for rated 
loop recirculation flow rate, relocate cycle-specific equations to the 
Core Operating Limits report, and provide other miscellaneous changes.
    Date of issuance: February 24, 1995
    Effective date: February 24, 1995
    Amendment Nos.: 216, 232, 190
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49437) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 24, 1995.No significant 
hazards consideration comments received: None [[Page 14038]] 
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendment: October 7, 1994
    Brief description of amendment: Eliminates redundancy in system 
leakage test requirements by revising TS 3/4.5.2 and its associated 
basis for the Emergency Core Cooling System and TS 3/4.6.2 and its 
associated basis for the Containment Spray System.
    Date of issuance: February 27, 1995
    Effective date: February 27, 1995 and to be implemented within 90 
days.
    Amendment No. 195
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 9, 1994 (59 FR 
55893) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 27, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

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: November 18, 1994 (published in Federal 
Register as November 11, 1994)
    Brief description of amendments: The proposed amendments would 
provide for cycle-specific allowances to account for increases in the 
Heat Flux Hot Channel Factor between monthly surveillances.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995, to be implemented within 30 days of 
issuance.
    Amendment Nos.: Unit 1 - Amendment No. 34; Unit 2 - Amendment No. 
20
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63127) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated March 1, 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, Texas 76019.

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington

    Date of application for amendment: October 31, 1994

    Brief description of amendment: The amendment modifies the Techical 
Specifications (TS) to (1) add two action statements that would provide 
allowed outage times for either one or both of the scram discharge 
volume (SDV) vent or drain valves less stringent than the current 
requirements of TS 3.0.3., and (2) change the surveillance requirements 
for the SDV vent and drain valves to conduct the testing during 
shutdown conditions rather than at power as currently required.
    Date of issuance: February 27, 1995
    Effective date: February 27, 1995
    Amendment No.: 134
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65828) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 27, 1995.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

Wisconsin Public Service Corporation, Docket No. 50-305, Kewaunee 
NuclearPower Plant, Kewaunee County, Wisconsin

    Date of application for amendment: February 23, 1994
    Brief description of amendment: The amendment revises Kewaunee 
Nuclear Power Plant (KNPP) Technical Specification (TS) 6.8.c by 
removing the requirement to conduct a biennial review of plant 
procedures in accordance with American National Standards Institute 
(ANSI) N18.7-1976, Section 5.2.15. Alternate programs that are 
described in the KNPP Operational Quality Assurance Program Description 
(OQAPD) will be used to ensure that procedures are reviewed and 
maintained current.
    Date of issuance: February 23, 1995
    Effective date: February 23, 1995 and to be implemented within 30 
days.
    Amendment No.: 115
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14903) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated February 23, 1995.No significant 
hazards consideration comments received: None.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for [[Page 14039]] example, in derating or shutdown of a 
nuclear power plant or in prevention of either resumption of operation 
or of increase in power output up to the plant's licensed power level, 
the Commission may not have had an opportunity to provide for public 
comment on its no significant hazards consideration determination. In 
such case, the license amendment has been issued without opportunity 
for comment. If there has been some time for public comment but less 
than 30 days, the Commission may provide an opportunity for public 
comment. If comments have been requested, it is so stated. In either 
event, the State has been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By April 14, 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 limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, 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 [[Page 14040]] 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

Pennsylvania Power and Light Company, Docket No. 50-388, 
SusquehannaSteam Electric Station, Unit 2, Luzerne County, 
Pennsylvania

    Date of application for amendment: February 7, 1995
    Brief description of amendment: The amendment changed the Technical 
Specifications to allow continued operation with one neutron flux 
monitor system channel (B'' channel) inoperable and should 
the remaining channel become inoperable to allow continued plant 
operation for 7 days to restore one of the two inoperable channels.
    Date of issuance: March 1, 1995
    Effective date: March 1, 1995
    Amendment No.: 115
    Facility Operating License No. NPF-22: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: No. On February 8, 1995, the staff 
issued a Notice of Enforcement Discretion, which was immediately 
effective and remained in effect until this amendment was issued.
    The Commission's related evaluation of the amendments, finding of 
emergency circumstances, consultation with the Commonwealth of 
Pennsylvania and final no significant hazards considerations 
determination are contained in a Safety Evaluation dated March 1, 1995.
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge 2300 N Street NW., Washington, D.C. 20037
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18071.
    NRC Project Director: John F. Stolz
    Dated at Rockville, Maryland, this 8th day of March 1995.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Acting Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 95-6207 Filed 3-14-95; 8:45 am]
BILLING CODE 7590-01-F