[Federal Register Volume 60, Number 31 (Wednesday, February 15, 1995)]
[Notices]
[Pages 8741-8766]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 95-3629]



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


NUCLEAR REGULATORY COMMISSION

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

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 January 20, 1995, through February 3, 1995. 
The last biweekly notice was published on February 1, 1995 (60 FR 
6296).

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 March 17, 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 [[Page 8742]] 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.

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: January 19, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Surveillance Requirement 4.0.3 and 
its associated bases to provide for a delay period of up to 24 hours in 
which to perform a surveillance which has been discovered not to have 
been performed within its specified frequency. This change would adopt 
the requirements of NUREG-1431, ``Standard Technical Specifications, 
Westinghouse Plants.''
    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 will reduce the requirement to unnecessarily 
manipulate and challenge plant systems and equipment. The most probable 
result of performing a surveillance during the delay period will be to 
verify its conformance with Technical Specification requirements. Since 
this change does not affect plant design, operation, or the manner in 
which testing is performed, the consequences of accident scenarios 
postulated in the Final Safety Analysis Report will not increase. 
Therefore, there would be no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously evaluated.
    The proposed change does not introduce any new equipment, nor does 
it require existing systems to perform a different type of function 
than they are currently designed to perform. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment does not involve a significant reduction 
in the margin of safety.
    The margin of safety is neither described or prescribed for this 
specification. The proposed change simply provides additional time to 
perform a surveillance and verify that the operability of equipment is 
in conformance with the Technical Specification requirements. 
Therefore, the proposed change does not involve a significant reduction 
in [the] margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: December 23, 1994.
    Description of amendment request: The proposed amendments would 
increase the allowable enrichment of new fuel stored in the new fuel 
storage vault (NFSV), revise the enrichment description of fuel in the 
reactor core, and include references to documents previously approved 
by the staff in the [[Page 8743]] Technical Specifications that provide 
analytical methods used to determine core operating limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    A.1. The proposed change does not involve a significant increase in 
the probability of occurrence or consequences of any accident 
previously evaluated.
    The Updated Final Safety Analysis Report (UFSAR) does not consider 
any accidents involving the NFSV. The Fuel Handling Accidents that are 
analyzed (Section 15.7.4) include dropping of a spent fuel assembly 
onto the spent fuel pool floor and breaking of all fuel rods, and 
dropping of a fuel assembly inside containment onto the top of the 
core.
    The proposed change to increase the NFSV fuel enrichment limit from 
4.0 to 4.65 weight percent U-235 does not affect any of the initiators 
or precursors of any accident previously evaluated. The proposed change 
will not increase the likelihood that a transient initiating event will 
occur because transients are initiated by equipment malfunction and/or 
catastrophic system failure. Since the proposed change does not involve 
the introduction of new or redesigned plant equipment, failure 
mechanisms are not affected. As a result, the probability of occurrence 
of accidents previously evaluated is not significantly increased.
    A new criticality analysis for the proposed change to increase the 
NFSV fuel enrichment limit from 4.0 to 4.65 weight percent U-235 was 
performed for the NFSV. It was determined that even in worst case 
conditions the acceptance criteria was met since the maximum Keff 
was determined to be well below the 0.95 limit with a 95/95 
probability/confidence level. The consequences of any accident, 
including a fuel handling accident involving the NFSV, are not 
significantly increased.
    A.2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed change to the Technical Specifications does not 
involve the addition of any new or different types of safety related 
equipment, nor does it involve the operation of equipment required for 
safe operation of the facility in a manner different from those 
addressed in the safety analysis. No safety related equipment or 
function will be altered as a result of the proposed changes. Also, the 
procedures governing normal plant operation and recovery from an 
accident are not changed by the proposed Technical Specification 
changes. Since no new failure modes or mechanisms are added by the 
proposed changes, the possibility of a new or different kind of 
accident is not created.
    A.3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant safety margins are established through LCOs, limiting safety 
system settings, and safety limits specified in the Technical 
Specifications. There will be no changes to either the physical design 
of the plant or to any of these settings and limits as a result of 
increasing the NFSV fuel enrichment limit. The change does not involve 
a significant increase in the probability of occurrence or consequences 
of any accident previously evaluated or create the possibility of a new 
or different kind of accident from any previously analyzed. 
Additionally, the revised criticality analysis demonstrates that the 
maximum Keff under all postulated conditions remains below the 
acceptance value of 0.95. Therefore, the change will not result in a 
significant reduction in a margin of safety.
    B.1. The proposed change does not involve a significant increase in 
the probability of occurrence or consequences of any accident 
previously evaluated.
    The proposed change to increase the reactor core fuel enrichment 
range discussed in the Design Features section of Technical 
Specifications from ``between 2.2 to 4.0'' to ``up to 4.65'' weight 
percent U-235 is administrative in nature and does not affect any of 
the initiators or precursors of any accident previously evaluated. The 
proposed change will not increase the likelihood that a transient 
initiating event will occur because transients are initiated by 
equipment malfunction and/or catastrophic system failure. Since the 
proposed change does not involve the introduction of new or redesigned 
plant equipment, failure mechanisms are not affected. As a result, the 
probability of occurrence of accidents previously evaluated is not 
significantly increased.
    The fuel enrichment limit of each core is determined by the core 
specific design and is determined to be acceptable with respect to the 
accident analysis by the reload analysis and is not impacted by the 
value specified in the description in the Design Features section of 
Technical Specifications. This value is only provided as the highest 
expected core fuel enrichment in the Design Features section discussion 
of the reactor core. This change is administrative in nature and does 
not affect the consequences of any accident previously evaluated.
    B.2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed change in the reactor core fuel enrichment description 
contained in the Design Features section of Technical Specifications 
does not involve the addition of any new or different types of safety 
related equipment, nor does it involve the operation of equipment 
required for safe operation of the facility in a manner different from 
those addressed in the safety analysis. No safety related equipment or 
function will be altered as a result of the proposed change. Also, the 
procedures governing normal plant operation and recovery from an 
accident are not changed by the proposed Technical Specification 
change. Since no new failure modes or mechanisms are added by the 
proposed change, the possibility of a new or different kind of accident 
is not created.
    B.3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant safety margins are established through LCOs, limiting safety 
system settings, and safety limits specified in the Technical 
Specifications. There will be no changes to either the physical design 
of the plant or to any of these settings and limits as a result of 
increasing reactor core fuel enrichment value given in the Design 
Features section of Technical Specifications. The change does not 
involve a significant increase in the probability of occurrence or 
consequences of any accident previously evaluated or create the 
possibility of a new or different kind of accident from any previously 
analyzed.
    Based on the above discussion, the ability to safely shutdown the 
operating unit and mitigate the consequences of all accidents 
previously evaluated will be maintained. Therefore, the margin of 
safety is not significantly affected.
    C.1. The proposed change does not involve a significant increase in 
the probability of occurrence or consequences of any accident 
previously evaluated.
    The proposed change to add three documents to the list of documents 
that provide the analytical methods to determine core operating limits 
is administrative in nature and does not affect any of the initiators 
or precursors of any accident previously evaluated. The proposed change 
will not increase the likelihood that a transient initiating event will 
occur because transients are initiated by equipment malfunction 
[[Page 8744]] and/or catastrophic system failure. Since the proposed 
change does not involve the introduction of new or redesigned plant 
equipment, failure mechanisms are not affected.
    The documents have been previously reviewed and approved by the NRC 
and it was determined that they provide an acceptable means to 
determine core operating limits. As a result, the probability of 
occurrence of accidents previously evaluated is not significantly 
increased. Since the documents provide NRC approved methodologies for 
determining core operating limits, the addition of the documents to 
Technical Specifications or use of the documents to determine core 
operating limits will not significantly increase the consequences of 
any accident previously evaluated.
    C.2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    The proposed change to add three documents to the list of documents 
that provide the analytical methods to determine core operating limits 
is administrative in nature and does not involve the addition of any 
new or different types of safety related equipment, nor does it involve 
the operation of equipment required for safe operation of the facility 
in a manner different from those addressed in the safety analysis. No 
safety related equipment or function will be altered as a result of the 
proposed changes. Also, the procedures governing normal plant operation 
and recovery from an accident are not changed by the proposed Technical 
Specification changes. Since no new failure modes or mechanisms are 
added by the proposed changes, the possibility of a new or different 
kind of accident is not created.
    C.3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Plant safety margins are established through LCOs, limiting safety 
system settings, and safety limits specified in the Technical 
Specifications. There will be no changes to either the physical design 
of the plant or to any of these settings and limits as a result of 
adding references to the new documents. The ability to mitigate the 
consequences of all accidents previously evaluated will be maintained. 
Therefore, the margin of safety is not significantly affected.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690.
    NRC Project Director: Robert A. Capra.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York Date 
of amendment request: September 19, 1994.

    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) Section 4.4.A.3 to reference the 
testing frequency requirements of 10 CFR Part 50, Appendix J, and to 
state that NRC approved exemptions to the applicable regulatory 
requirements are permitted. This proposed administrative revision 
simply deletes the paraphrased language and directly references 
Appendix J.
    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--Does Not Involve a Significant Increase in the Probability 
or Consequences of an Accident Previously Evaluated
    The proposed change will provide a one-time exemption from the 10 
CFR [Part] 50, Appendix J Section III.D.1.(a) leak rate test schedule 
requirement. This change will allow for a one-time test interval for 
Type A Integrated Leak Rate Tests (ILRTs) of approximately 70 months.
    Leak rate testing is not an initiating event in any accident, 
therefore this proposed change does not involve a significant increase 
in the probability of a previously evaluated accident.
    Type A tests are capable of detecting both local leak paths and 
gross containment failure paths. The history at IP-2 [Indian Point 2] 
demonstrates that Type B and C Local Leak Rate Tests (LLRTs) have 
consistently detected any excessive local leakages.
    Administrative controls govern the maintenance and testing of 
containment penetrations such that the probability of excessive 
penetration leakage due to improper maintenance or valve misalignment 
is very low. Following maintenance on any containment penetration, an 
LLRT is performed to ensure acceptable leakage levels, following any 
LLRT on a containment isolation valve, an independent valve alignment 
check is performed. Therefore, Type A testing is not necessary to 
ensure acceptable leakage rates through containment penetrations.
    While Type A testing is not necessary to ensure acceptable leakage 
rates through containment penetrations, Type A testing is necessary to 
demonstrate that there are no gross containment failures. Structural 
failure of the containment is considered to be a very unlikely event, 
and in fact, since IP-2 has been in operation it has never failed a 
Type A ILRT. Therefore, a one-time exemption increasing the interval 
for performing an ILRT should not result in a significant decrease in 
the confidence in the leak tightness of the containment structure.
    The proposed change also revises Technical Specification 4.4.A.3 to 
reference the testing frequency requirements of 10 CFR [Part] 50, 
Appendix J, and to state that NRC approved exemptions to the applicable 
regulatory requirements are permitted. The current language of TS 
4.4.A.3 paraphrases the requirements of Section III.D.1.(a) of Appendix 
J. The proposed administrative revision simply deletes the paraphrased 
language and directly references Appendix J. 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 Consolidated Edison 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 TS changes if 
different exemptions are approved in the future.
    The proposed change, in itself, does not affect reactor operations 
or accident analysis 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, the 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 exemption request does not affect normal plant 
operations or configuration, nor does it affect leak rate test methods. 
The proposed change allows a one-time test interval of 
[[Page 8745]] approximately 70 months for the ILRT. Given the test 
history of IP-2 of no Type A test failures during plant lifetime, the 
relaxation in schedule should not significantly decrease the confidence 
in the leak tightness of the containment.
    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 purpose of the existing schedule for ILRTs is to ensure that 
the release of radioactive materials will be restricted to those leak 
paths and leak rates assumed in accident analyses. The relaxed schedule 
for ILRTs does not allow for relaxation of Type B and C LLRTs. 
Therefore, methods for detecting local containment leak paths and leak 
rates are unaffected by this proposed change. Given that the test 
history for ILRTs shows no failure during plant life, a one-time 
increase of the test interval does not lead to a significant 
probability of creating a new leakage path or increased leakage rates, 
and the margin of safety inherent in existing accident analyses is 
maintained.
    The proposed Technical Specification change is administrative and 
clarifies the relationship between the requirements of TS 4.4.A.3, 
Appendix J and any approved exemptions to Appendix J. It does not, in 
itself, change a safety limit, an LCO [limiting condition for 
operation], or a surveillance requirement on equipment required to 
operate the plant. The NRC will directly approve any proposed change or 
exemption to [Section] III.D.1.(a) of Appendix J prior to 
implementation.
    Therefore, this change does not involve a significant reduction in 
the margin of safety.
    Based on the Safety Analysis, it is concluded that: (1) The 
proposed change does not constitute a significant hazards consideration 
as defined by 10 CFR 50.92 and (2) there is reasonable assurance that 
the health and safety of the public will not be endangered by the 
proposed change. Moreover, because this action does not involve a 
significant hazards consideration, it will also not result in a 
condition which significantly alters the impact of the station on the 
environment as described in the NRC Final Environmental Statement.
    Although the licensee has included an evaluation of a proposed 
exemption to 10 CFR part 50, Appendix J requirements in the above 
determination of no significant hazards consideration, only the part 
related to the amendment is pertinent to this notice of proposed 
amendment. The exemption request will be considered as a separate 
matter on its own merits. 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.
    Attorney for licensee: Brent L. Brandenburg, Esq., 4 Irving Place, 
New York, New York 10003.
    NRC Project Director: Ledyard B. Marsh

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

    Date of amendment request: October 31, 1994
    Description of amendment request: The requested amendments would 
remove the stroke times for the steam generator power operated relief 
valves (PORVs) from Technical Specification (TS) Tables 3.6-2a and 3.6-
2b. The PORVs are part of the main steam vent to atmosphere system. The 
PORV actuators have difficulty developing enough closing thrust to 
adequately overcome all of the friction loads within the valves; 
therefore, difficulty exists in consistently meeting the present 5-
second closing stroke time requirement. The licensee requests the 
proposed change on the basis that the PORVs do not receive an actual 
containment isolation signal; therefore, it is justified to remove the 
stroke times from TS Tables 3.6-2a and 3.6-2b.
    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 48 FR 14870, the Commission has set forth examples of amendments 
that are considered not likely to involve significant hazards 
considerations. Example (vi) describes a change which either may result 
in some increase to the probability or consequences of a previously-
analyzed accident or may reduce in some way a safety margin, but where 
the results of the change are clearly within all acceptable criteria 
with respect to the system or component specified in the Standard 
Review Plan. In this case, the proposed amendment is similar to example 
(vi) in that it removes the required isolation time of the steam 
generator PORVs from TS Tables 3.6-2a and 3.6-2b; however, no adverse 
impact upon accident analyses is created as a result.
Criterion 1
    The requested amendments will not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The effects of the delays in isolation times on the various transients 
affected have been analyzed and found to be acceptable. Since these 
valves do not receive a containment isolation signal, and no credit is 
taken for operation of these valves in the dose analysis for a 
containment isolation function, a maximum stroke time does not apply 
for containment isolation.
Criterion 2
    The requested amendments will not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
SV PORV closure (provided the valves are not already closed at the 
start of the transient) is a response to a transient already in 
progress. The possibility of a spurious SV PORV opening will not be 
affected by the requested amendments. No equipment or component 
reconfiguration will occur as a result of this change. Finally, no 
changes to plant procedures are being made which would affect any 
accident causal mechanisms.
Criterion 3
    The requested amendments will not involve a significant reduction 
in a margin of safety. The isolation times which are applicable to 
these valves are specified in TS Table 3.3-5, Engineered Safety 
Features Response Times. The effects of the isolation of these valves 
were evaluated based on their ESF function, not a containment isolation 
function, and determined to be acceptable.
    Based upon the preceding analyses, Duke Power Company concludes 
that the requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
[[Page 8746]] amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: June 13, 1994, as supplemented August 
15, 1994.
    Description of amendment request: The proposed changes would 
increase the initial fuel enrichment limit from a current maximum of 
4.0 weight % to 4.75 weight % and establish new loading patterns for 
new and irradiated fuel in the spent fuel pool to accommodate this 
increase. These changes would also increase the efficiency of fuel 
storage cell use in the spent fuel pools and provide additional 
flexibility to the reload design efforts at Duke Power Company, while 
at the same time maintaining sufficient criticality safety margin and 
decay heat removal capabilities.
    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.
    There is no increase in the probability or consequences of an 
accident in the new fuel vault since the only credible accidents for 
this area are criticality accidents and it has been shown that 
calculated, worst case Keff for this area is 0.95 under 
all conditions.
    There is no increase in the probability of a fuel drop accident in 
the Spent Fuel Storage Pool since the mass of an assembly will not be 
affected by the increase in fuel enrichment. The likelihood of other 
accidents, previously evaluated and described in Section 9.1.2 of the 
FSAR [Final Safety Analysis Report], is also not affected by the 
proposed changes. In fact, it could be postulated that since the 
increase in fuel enrichment will allow for extended fuel cycles, there 
will be a decrease in fuel movement and the probability of an accident 
may likewise be decreased. There is also no increase in the 
consequences of a fuel drop accident in the Spent Fuel Pool since the 
fission product inventory of individual fuel assemblies will not change 
significantly as a result of increased initial enrichment. In addition, 
no change to safety related systems is being made. Therefore, the 
consequences of a fuel rupture accident remain unchanged. Also, it has 
been shown that keff is 0.95, under all conditions 
therefore, the consequences of a criticality accident remain unchanged 
as well.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of a new or 
different kind of accident since fuel handling accidents (fuel drop and 
misplacement) are not new or different kinds of accidents. Fuel 
handling accidents are already discussed in the FSAR for fuel with 
enrichments up to 4.1 weight %. As described in Section VI.9 of 
Attachment IV, additional analyses have been performed for fuel with 
enrichment up to 4.75 weight %. Worst case misloading accidents 
associated with the new loading patterns were evaluated. For all 
possible misloading accidents the negative reactivity provided by 
soluble boron maintains keff 0.95. of safety.
    3. The proposed changes do not involve a significant reduction in 
the margin of safety.
    The proposed change does not involve a significant reduction in the 
margin of safety since, in all cases, a keff 0.95 is 
being maintained. Criticality analyses have been performed which show 
that the new fuel storage vault will remain subcritical under a variety 
of moderation conditions, from fully flooded to optimum moderation. As 
discussed above, the Spent Fuel Pool will remain sufficiently 
subcritical during any fuel misplacement accident.
    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: November 11, 1994, as supplemented 
January 30, 1995.
    Description of amendment request: The amendments would revise the 
Technical Specifications Design Features section to establish 
restricted loading patterns and associated burnup criteria for placing 
fuel in the Oconee Spent Fuel Pools. These changes are necessary to 
address two new fuel designs which have increased initial fuel 
enrichment and therefore cannot be stored in the spent fuel pools under 
existing Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Standard 1. The proposed amendments will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    Each accident analysis addressed in the Oconee Final Safety 
Analysis Report (FSAR) has been examined with respect to changes in 
Cycle 15 parameters to determine the effect of the Cycle 16 reload and 
to ensure that the acceptance criteria of the FSAR safety analyses 
remain satisfied. The transient evaluation of Cycle 16 is considered to 
be bounded by previously accepted analyses. Section 7 of the Reload 
Report addresses ``Accident and Transient Analysis'' for this core 
reload.
    There is no increase in the probability or consequences of an 
accident due to the spent fuel storage restrictions proposed in this 
amendment request. It has been shown that the calculated, worst case 
keff for this area is [less than or equal to] 0.95 under all 
conditions. There is no increase in the probability of a fuel drop 
accident in the SFP [spent fuel pool] since the mass of the new 
assemblies is not significantly different from the mass of the old 
assemblies. The likelihood of other accidents, previously evaluated and 
described in the FSAR, is also not affected by the proposed changes. In 
fact, it could be postulated that since the increase in fuel enrichment 
will allow for extended fuel cycle lengths, there will be a decrease in 
fuel movement and the probability of an accident may actually be 
reduced. There is also no increase in the consequences of a fuel rod 
drop accident in the SFP since the fission product inventory of 
[[Page 8747]] individual fuel assemblies will not change significantly 
as a result of increasing the initial enrichment. In addition, no 
change to safety related systems is being made. Therefore, the 
consequences of a fuel rupture accident remain unchanged. In addition, 
it has been shown that keff is [less than or equal to] 0.95 under 
all conditions. Therefore, the consequences of a criticality accident 
in the SFP remain unchanged as well. The above analysis ensures that 
the proposed reload amendment request will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The analyses performed in support of this reload are in accordance 
with the NRC approved methods delineated in Specification 6.9.2. The 
predicted operating characteristics of Oconee 3 Cycle 16 are similar to 
previously licensed designs. The Mark B10T and Mark B11 fuel assembly 
designs remain mechanically compatible with all fuel handling 
equipment. Therefore, no new or different kind of fuel handling 
accident is created by the proposed amendment request.
    Section 15.11 of the Oconee FSAR states that the refueling boron 
concentration is maintained such that a criticality accident during 
refueling is not considered credible. The proposed amendment request 
continues to assure that a criticality accident in the SFP or during 
refueling is not credible. The double contingency principle discussed 
in ANSI N-16.1-1975 and the April 1978 NRC letter allows credit for 
soluble boron under other abnormal or accident conditions, since only a 
single accident need be considered at one time. Thus, by requiring a 
minimum boron concentration in the SFP, a criticality accident caused 
by violating the SFP storage restrictions is not considered credible. 
Therefore, the proposed amendment request does 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 
the margin of safety.
    The Oconee 3 Cycle 16 design was performed using the NRC approved 
methods given in Specification 6.9.2. The safety limits for Oconee 3 
Cycle 16 are unchanged from previous cycles. The limits and margins 
summarized in the Oconee 3 Cycle 16 Reload Report are well within the 
allowable limits and requirements, and reflect no reductions to any 
margins of safety.
    The proposed change does not involve a significant reduction in the 
margin of safety related to SFP criticality. In all cases, a keff 
[less than or equal to] 0.95 is maintained. Criticality analyses have 
been performed which show that the SFP will remain sufficiently 
subcritical during any fuel misplacement accident. In summary the 
proposed changes do not involve a significant reduction in the margin 
of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

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

    Date of amendment request: January 6, 1995
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 3/4.8.1.1, ``AC Sources-
Operating,'' and 3/4.8.1.2, ``AC Sources-Shutdown,'' to (1) revise the 
minimum quantity of fuel oil required in the day tanks and the storage 
tanks, (2) add specific actions to be taken if the storage tank levels 
fall below minimum requirements, (3) revise and relocate to the 
associated Bases the fuel oil sampling and testing criteria, and (4) 
add specific actions to be taken if the fuel oil properties do not meet 
specified limits. The proposed amendment would also revise TS 6.8.4, 
``Programs,'' to add a requirement for a diesel fuel oil testing 
program. The licensee stated that the proposed changes are consistent 
with the NRC's Improved Standard Technical Specifications (NUREG-1434).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The diesel generators are not initiators or precursors to an 
accident previously evaluated. The diesel generators are required to 
provide onsite power to safe shutdown loads as assumed in the accident 
analysis. Therefore, the proposed changes to the diesel generator fuel 
oil specifications cannot significantly affect the probability of a 
previously evaluated accident.
    The proposed change to the minimum required diesel generator fuel 
oil levels is based on updated calculations of fuel consumption rates. 
Because the updated calculations assume a lower consumption rate, the 
new minimum fuel oil levels are lower but still assure that a seven-day 
fuel oil capacity is available. Accordingly, the proposed change has no 
effect on the operation of the diesel generator. The proposed change to 
allow 48-hours to restore diesel generator fuel oil to the minimum 
required level does not affect short-term diesel generator operability 
and is acceptable based on the remaining fuel oil capacity (>6 days), 
initiating the process for procuring additional fuel and the low 
probability of an event requiring a diesel generator during this 
interval. Also, the proposed allowance of a limited time to restore 
diesel fuel oil properties to required limits will not affect the 
short-term operability of the diesel generator. Even with minor 
degradation of the fuel oil properties, the diesels will start and 
perform their intended function. Relocation of the testing requirements 
to the bases and adding a description of the Diesel Fuel Oil Testing 
Program to the Administrative Control section are administrative 
changes. The diesel fuel oil will continue to be sampled and tested in 
a manner to assure its quality. In summary, the changes will not 
adversely affect the performance or the ability of the diesel 
generators to perform their intended function. Therefore, the proposed 
changes will not significantly increase the consequences of an accident 
previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes will revise the minimum required diesel 
generator fuel oil levels and requirements associated with diesel 
generator fuel oil properties. [[Page 8748]] The changes do not 
introduce any new accident precursors and do not involve any 
alterations to plant configurations which could initiate a new or 
different kind of accident. The proposed changes do not affect the 
short-term operability of the diesel generator. In addition, the 
operability of the diesel generators is assured by periodic testing and 
preventive maintenance. Therefore, the proposed changes will not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    Safety margins are established through safety analyses. These 
analyses assume that at least one diesel generator will start and load 
whenever offsite power is lost. The proposed change to the minimum 
required diesel generator fuel oil levels is based on updated 
calculations of fuel consumption rates. The updated calculations use 
the guidance delineated in Regulatory Guide 1.137 which is based on 
time-dependent loads of the diesel-generators during design basis 
events. Calculations based on time dependent loads result in new 
minimum fuel oil levels which are lower. This change has no effect on 
the operation of the diesel generator or on a margin of safety. The 
allowance of a limited time to restore the fuel oil levels, or to 
analyze and restore fuel oil properties to required limits, is 
justified since the short term operability of the diesel generators is 
not affected. Relocation of the fuel oil testing requirements to the 
Bases does not affect the quality of the fuel oil. The 10CFR50.59 
process will assure that future changes to the Bases will maintain the 
current margins of safety, and that the diesel fuel oil will continue 
to be sampled and tested in such a manner as to assure its quality. 
Adding a description of the Diesel Fuel Oil Testing Program to the 
Administrative Control section of Technical Specifications are 
administrative. Therefore, the diesel generator will continue to 
operate as analyzed and there will not be a significant reduction in a 
margin of safety.
    The proposed changes are further justified in that they are 
consistent with the requirements of the Improved Standard Technical 
Specifications (NUREG-1434).
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Ledyard B. Marsh.

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

    Date of amendment request: January 6, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs) 3/4.3.7.5, ``Accident Monitoring 
Instrumentation,'' and TS 3/4.4.2, ``Safety/Relief Valves.'' TS 3/
4.3.7.5 would be revised to delete certain instruments not classified 
as Category 1 (Type A or non-Type A) as defined in Regulatory Guide 
1.97 and to delete the requirement that accident monitoring 
instrumentation be operable in Operational Condition 3. The ACTIONS of 
TS Table 3.3.7.5-1 would be revised to allow 30 days to restore one 
inoperable channel and 7 days to restore two inoperable channels. TS 
3.3.7.5 would be revised to add an exception to the requirements of TS 
3.0.4. In addition, editorial changes would be made to TS Tables 
3.3.7.5-1 and 4.3.7.5-1 for consistency and clarity.
    The proposed amendment would also revise TS 3/4.4.2 to remove 
requirements related to safety/relief valve acoustic monitors to be 
consistent with the proposed changes to TS Tables 3.3.7.5-1 and 
4.3.7.5-1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of NMP2 [Nine Mile Point Nuclear Station Unit 2] in 
accordance with the proposed amendment, will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated.
    PAM [Post-Accident Monitoring] instruments are used to help guide 
operator response to postulated accidents. Thus, the status or 
operability of PAM instrumentation does not affect the probability of 
previously analyzed accidents. The non-Category 1 PAM instruments being 
removed from the Technical Specifications do not meet any of the 
Commission's screening criteria and are not of controlling importance 
to safety or necessary to obviate the possibility of an abnormal 
situation or event giving rise to an immediate threat to public health 
and safety. The operability of critical parameters necessary to assure 
proper response to previously analyzed accidents (i.e., Category 1 
instruments) is still controlled by the Technical Specifications. Thus, 
deleting non-Category 1 instruments will not increase the consequences 
of any accident previously evaluated.
    PAM instruments are related to the diagnosis and preplanned actions 
required to mitigate DBAs [Design Basis Accidents] assumed to occur in 
Operational Conditions 1 and 2. A DBA during Operational Condition 3 is 
extremely unlikely. The requirement to maintain the Reactor Water 
Level, Suppression Pool Water Level and Drywell High Range Radiation 
Monitor instrumentation operable in Operational Condition 3 will be 
deleted. Because Suppression Pool Water Level indication will no longer 
be required in Operational Condition 3, its ACTION requirement was 
revised to delete the requirement to place the plant in COLD SHUTDOWN, 
Operational Condition 4. This is consistent with ITS [Improved Standard 
Technical Specifications] which requires that the plant be brought to 
an operational condition in which the LCO [Limiting Condition for 
Operation] does not apply if a required action cannot be met. 
Therefore, deleting the requirement that PAM instruments be operable 
during Operational Condition 3 and changing the ACTION requirement for 
Suppression Pool Water Level Monitoring does not affect the probability 
or consequences of an accident.
    The passive nature of the Category 1 PAM instruments (i.e., those 
instruments that initiate no critical automatic action) and the 
alternate means available to obtain the required information assure an 
acceptable level of safety is maintained during operation with 
instrument channels out of service. Since an acceptable level of safety 
is maintained with inoperable channels, plant startup or operation with 
inoperable channels will not alter plant response to analyzed 
accidents. Thus, the proposed changes to the required ACTIONS and the 
proposed exemption to Specification 3.0.4 will not increase the 
consequences of analyzed events.
    The proposed changes to the requirements for PCIV [Primary 
[[Page 8749]] Containment Isolation Valve] indication are consistent 
with the proposed required ACTIONS. Position indication will still be 
required for each operable PCIV and penetrations without adequate PCIV 
indication status will be isolated, thus assuring containment integrity 
in the event of an accident. Deletion of the ``Minimum Required 
Actions'' column in Table 3.3.7.5-1 is consistent with the proposed 
ACTIONS for LCO 3.3.7.5, since compensatory actions are based on 
compliance with the ``Required Number of Channels.'' Deleting the 
``Applicable Operating Conditions'' column is consistent with the 
proposed changes and other NMP2 Technical Specifications sections. 
Finally, referencing Specification 4.0.5 is an administrative change 
which does not alter any existing surveillance requirements for the 
safety relief valves.
    In aggregate, the proposed changes do not affect the plant in a way 
that could directly contribute to causing or mitigating the effects of 
an accident. Therefore, the operation of NMP2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The operation of NMP2, in accordance with the proposed amendment, 
will not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not represent a physical change to the 
plant as described in the NMP2 USAR [Updated Safety Analysis Report]. 
The proposed changes do not modify any plant equipment and the initial 
conditions used for the design basis accident analysis are still valid. 
Thus, no potential initiating events are created which would cause any 
new or different kinds of accidents. PAM instrumentation is used to 
guide operator response during postulated accidents. Those PAM 
instruments considered of controlling importance to safety are retained 
in the Technical Specifications. Thus, plant response to previously 
analyzed events is not altered so as to create any new or different 
kinds of accidents. Therefore, operation of Nine Mile Point Unit 2 in 
accordance with the proposed change will not create the possibility of 
a new or different kind of accident from any previously assessed.
    The operation of NMP2, in accordance with the proposed amendment, 
will not involve a significant reduction in a margin of safety.
    The non-Category 1 PAM instruments being removed from the Technical 
Specifications do not meet any of the Commission's screening criteria. 
That is, the instruments being proposed for removal are not of 
controlling importance to safety or necessary to obviate the 
possibility of an abnormal situation or event giving rise to an 
immediate threat to public health and safety. Thus, they are not 
critical to any margin of safety.
    PAM instruments are related to the diagnosis and preplanned actions 
required to mitigate DBAs assumed to occur in Operational Conditions 1 
and 2. A DBA during Operational Condition 3 is extremely unlikely. The 
requirement to maintain the Reactor Water Level, Suppression Pool Water 
Level and Drywell High Range Radiation Monitor instrumentation operable 
in Operational Condition 3 will be deleted. Because Suppression Pool 
Water Level indication will no longer be required in Operational 
Condition 3, its ACTION requirement was revised to delete the 
requirement to place the plant in COLD SHUTDOWN, Operational Condition 
4. This is consistent with the ITS, which requires that the plant be 
brought to an operational condition in which the LCO does not apply if 
a required action cannot be met. Therefore, deleting the requirement 
that PAM instruments be operable during Operational Condition 3 and 
changing the ACTION requirement for Suppression Pool Water Level 
Monitoring does not significantly reduce a margin of safety.
    Since the Category 1 PAM instruments are passive in nature (i.e., 
no critical automatic action is assumed to occur from these 
instruments) and alternate means exist to obtain the required 
information, an acceptable level of safety is assured when instrument 
channels are out of service. Also, the probability of an event 
requiring PAM instrumentation is low. Continued operation with one 
channel out of service, and limited plant operation with two channels 
out of service, does not compromise plant safety margins. An acceptable 
level of safety is maintained during plant startups and operation with 
instrument channels out of service. Thus, the proposed changes to the 
required ACTIONS and the proposed exemption to Specification 3.0.4 will 
not significantly reduce a margin of safety.
    The proposed changes to PCIV indication will assure correct 
implementation of the ACTIONS discussed above. Isolating the flow path 
associated with one or two inoperable PCIV indication channels is 
conservative since the subject valve will be positioned as required to 
assure primary containment integrity. The remaining editorial changes 
are administrative in nature and by definition do not affect safety 
margins. Deleting the ``Minimum Operable Channels'' and ``Applicable 
Operating Conditions'' columns is consistent with the proposed changes. 
Finally, referencing the requirements of Specification 4.0.5 is an 
administrative change and by definition does not reduce the margin of 
safety.
    Therefore, the operation of NMP2 in accordance with the proposed 
change will not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Ledyard B. Marsh.

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

    Date of amendment request: December 9, 1994.
    Description of amendment request: The proposed changes incorporate 
NRC recommendations contained in Generic Letter 93-05 related to the 
diesel generator (DG) surveillance requirements and other DG 
surveillance requirements related to the cold starts. The proposed 
changes to the DG operability testing surveillance requirements are 
consistent with the intent of GL 93-05 however vary in some 
particulars, because of circumstances specific to Millstone 3. The 
proposed changes will modify the requirement for the DG operability 
testing when the other DG is inoperable, delete the requirement for DG 
operability testing when one or both offsite AC sources are inoperable, 
eliminate fast loading of DGs except for the 18-month test, and modify 
the hot restart test from the 24-hour loaded test run for the DGs.
    Basis for proposed no significant hazards consideration 
determination: [[Page 8750]] As required by 10 CFR 50.91(a), the 
licensee has provided its analysis of the issue of no significant 
hazards consideration (SHC), which is presented below:

* * * The proposed changes do not involve a SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The proposed changes revise the action requirements regarding 
operability testing of a non-affected DG when the other DG is 
inoperable, delete the requirement for operability testing of the DGs 
when one or both offsite AC sources are inoperable and eliminate the 
fast loading of DGs except for the 18-month test. These changes will 
improve DGs performance by reducing the number of unnecessary quick 
starts and by requiring more appropriate testing of the DGs when there 
is a potential for common mode failure. The proposed change, to revise 
the method of verifying DG hot restart capability after a 24-hour run 
without loading the DG with LOP/SI [loss of offsite power/safety 
injection] load, meets an intent of Regulatory Guide 1.108, Position 
C.2.a.5, which states the purpose of the test as to ``demonstrate 
functional capability at full load temperature conditions.'' Functional 
capability of the DG can be adequately demonstrated by manually or 
automatically restarting the DG within five minutes after a 24-hour 
test run without loading it with LOP/SI loads, provided that a full 
load temperature condition is maintained prior to restart. The proposed 
DG restart method does not reduce the effectiveness of the test. The 
proposed revisions of the DG surveillance requirements will not 
increase the probability of an accident and it will not change the 
response of the DG to a LOP as described in the Millstone Unit No. 3 
FSAR. Since the plant response to an accident will not change, there is 
no change in the potential for an increase in the consequences of an 
accident previously analyzed.
    2. Create the possibility of a new or different kind of an accident 
previously evaluated.
    The proposed changes of the DG surveillance requirements and 
operability testing requirements do not affect the operation or 
response of any plant equipment or introduce any new failure 
mechanisms. The proposed changes do not affect the test results and the 
DGs will be verified to be operable and their response to a loss of 
voltage will be unchanged. The plant equipment will respond per the 
design and analyses and there will not be a malfunction of a new or any 
type introduced by the revision to the DG surveillance requirements. As 
such, the changes do not create the possibility of a new or different 
kind of accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The bases of Technical Specification 3/4.8, ``Electrical Power 
Systems,'' state that the operability of the AC and DC power systems 
and associated distribution systems ensure that sufficient power will 
be available to supply the safety-related equipment required for safe 
shut down and mitigation and control of accident conditions. The bases 
also state that the surveillance requirements for determining the 
operability of the DGs are in accordance with the recommendations of 
Regulatory Guide 1.108, Revision 1. The revisions of the surveillance 
requirements establishes tests that will continue to verify that the 
DGs are operable and the testing will still meet the intent of 
Regulatory Guide 1.108, Revision 1. Operable DGs ensure that the 
assumptions in the bases of the Technical Specifications are not 
affected and ensure that the margin of safety is not reduced. 
Therefore, the assumptions in the bases of the technical specifications 
are not affected and these changes do not result in a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

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

    Date of amendment request: December 14, 1994.
    Description of amendment request: The proposed amendment would 
revise the Millstone Unit No. 3 Technical Specifications by:
    1. Increasing the upper bound of the overall containment integrated 
leakage rate required by Technical Specification 3.6.1.2.a from 0.3 wt. 
% per day to 0.65 wt. % per day of the containment air per 24 hours at 
design basis pressure.
    2. Revising Technical Specification 4.6.6.1.d.3 by providing more 
margin with respect to the drawdown time for secondary containment 
vacuum.
    3. Revising Bases Section 3/4.7.9 to reflect the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

* * * The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.

* * * There is a reasonable assurance that the modified criteria for 
the negative pressure in the secondary containment boundary proposed 
via the proposed change (i.e., a negative pressure of 0.1 inches in one 
minute and a negative pressure of 0.4 inches within the next two 
minutes), can be accomplished in the prescribed time.
    Extension of the time allowed to achieve the final drawdown of 
secondary containment from 120 seconds to 180 seconds (these times 
include the diesel generator start and load time of approximately 11 
seconds) will have a negligible impact on heating and cooling. Plant 
experience has shown that heatup and cooldown of thick-walled concrete 
structures, such as the Millstone Unit No. 3 auxiliary building, is a 
relatively slow process. Also, natural convection within the auxiliary 
building tends to stabilize temperatures. Following an accident signal, 
ventilation equipment is restarted promptly. Therefore, heatup or 
cooldown, during short periods while ventilation fans and/or heaters 
are inactive, is insignificant and can be neglected.
    The proposed change to reinstate the containment integrated leakage 
rate at the design basis pressure from 0.3 wt % per day to 0.65 wt % 
per day has been evaluated to determine the impact to the Appendix J 
requirements for Type A, B and C Testing. In addition, the radiological 
consequence evaluation also addressed the increase in La (i.e., 
from 0.3 wt % per day to 0.65 wt % per day).
    On October 12, 1993, Millstone Unit No. 3 successfully conducted 
the second [[Page 8751]] Type A test in the first 10-year service 
period. Test results indicated that the ``As-Found'' and ``As-Left'' 
ILRTs [integrated leakage rate tests] passed the technical 
specification acceptance criteria. The ``As-Found'' value was 0.1327 
weight percent per day and the ``As-Left'' value was 0.1313 weight 
percent per day. These values represent 27.2% and 26.9% of the 
technical specification criterion of 0.4875 wt % per day (0.75 
La), based on La equal to 0.65 wt % per day, respectively. In 
addition, as of October 9, 1993, the total Type B and C ``As-Found'' 
and ``As-Left'' leakage results were 0.099 wt % per day, and 0.084 wt % 
per day, respectively. These values represent approximately 25.3% and 
21.5% of the technical specification limit of 0.39% wt % per day (0.6 
La), based on La equal of 0.65 wt % per day, respectively. 
Correspondingly, the 1993 Type A, B, and C test results indicate that 
the ``As-Found'' and ``As-Left'' result in each test case was below the 
existing Technical Specification limit of 0.3 wt % per day. This 
further demonstrates the overall leakage integrity of the containment 
and its boundaries.
    Based on the relatively low ``As-Left'' ILRT leakage rate (i.e., 
0.1313 wt % per day is well below the existing technical specification 
limit of 0.225 wt % per day (0.75 La), based on La equal to 
0.3 wt % per day), which represents the overall containment integrated 
leakage rate for the containment prior to start-up, there is reasonable 
assurance that containment integrity will be maintained below the 
allowable leakage rate limit of 0.65 wt % per day. In addition, the 
total Type B and C ``As-Left'' leakage result of 0.084 wt % per day 
(this is well below the existing technical specification limit of 0.18 
wt % per day (0.6 La), based on La equal to 0.3 wt % per 
day), provides further assurance that leakage, based on individual 
penetration, will be maintained within sufficient margin of the leakage 
limits.
    Because the last Type A, B, and C tests were performed under the 
technical specification limit of 0.65 wt % per day, the proposed change 
to restore La to 0.65 wt % per day has no impact to these systems 
from a leakage allowance perspective. As indicated above, the previous 
test results met the technical specification leakage limits (based on 
0.65 wt % per day) within sufficient margin and, therefore, would not 
present any challenge to these leakage limits.
    NNECO has evaluated the proposed changes to Surveillance 
Requirement 4.6.6.1.d.3 that increase the time to draw a final required 
negative pressure as measured at the 24'-6'' elevation of the auxiliary 
building in conjunction with the proposed change to reinstate the 
containment integrated leakage rate of 0.65 wt % per day to determine 
the impact on the offsite doses following a LOCA. The calculated 
radiological doses are, in most cases, less than the previously 
calculated doses (i.e., EAB [exclusion area boundary] and LPZ [low-
population zone] doses) and are within the 10CFR100 limits. Previously, 
the EAB thyroid and whole body doses as documented in the November 4, 
1993, submittal were calculated to be 141 REM and 9.4 REM respectively, 
while the previously docketed (i.e., the November 4, 1993, submittal) 
LPZ doses to the thyroid and whole body were calculated to be 29.8 REM 
and 1.7 REM respectively. Utilizing the revised application of 
containment recirculation spray DF, the EAB thyroid and whole body 
doses were calculated to be 61 REM and 16.7 REM, respectively, and the 
LPZ thyroid and whole body doses were calculated to be 10.9 REM and 2.8 
REM respectively. The assumptions used in the above radiological dose 
calculations are provided in Attachment 1. It is noted that a LOCA at 
Millstone Unit No. 3 is also one of the bounding accidents for the 
Millstone Unit No. 3 control room, Millstone Unit No. 2 control room, 
and the Millstone Technical Support Center habitability analysis. 
Therefore, the doses for these areas were recalculated and are 
presented in the Safety Assessment section above. The Millstone Unit 
No. 1 control room and the Emergency Operating Facility doses are 
bounded by the Millstone Unit No. 1 LOCA calculations.
    The Millstone Unit Nos. 2 and 3 control rooms and Millstone 
Technical Support Center doses were not recalculated in 1993 (i.e., 
November 4, 1993, submittal) since EAB/LPZ doses proved that the 
releases were less than the 1990 submittal. In summary, all control 
room and Technical Support Center doses are within the guidelines of 
GDC 19. Therefore, the proposed changes do not result in an increase in 
consequences of an accident (i.e., a LOCA) previously analyzed.
    The proposed changes to Bases Section 3/4.6.6 do not have any 
safety impact since they only reflect the changes proposed to 
Surveillance Requirement 4.6.6.1.d.3.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed changes do not compromise the ability of the SLCRS 
[supplementary leak collection and release system] and ABFS [auxiliary 
building filter system] to mitigate the consequences of an accident. 
The proposed changes do not make any physical or operational changes to 
existing plant structures, systems or components. The proposed changes 
do not introduce any new or unique operational modes or accident 
precursors. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    NNECO has evaluated the proposed changes to Surveillance 
Requirement 4.6.6.1.d.3 that increase the time to draw a final required 
negative pressure as measured at the 24'-6'' elevation of the auxiliary 
building in conjunction with the proposed change to reinstate the 
containment integrated leakage rate of 0.65 wt % per day to determine 
the impact on the offsite doses following a LOCA. The calculated 
radiological doses are, in most cases, less than the previously 
calculated doses and these doses are within the 10CFR100 limits. All 
control rooms and technical support center doses are within the 
guidelines of GDC 19. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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.

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

    Date of amendment request: December 23, 1994.
    Description of amendment request: The proposed amendment would 
change the acceptance criteria for the peak transient generator voltage 
from 4784 volts to 5000 volts during full load rejection tests of the 
diesel generator (DG), and delete the 10-year surveillance requirement 
to perform a [[Page 8752]] 110% pressure test of the DG fuel oil 
system.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

* * * The proposed changes do not involve a SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
DG Full-Load Rejection Test
    NNECO is proposing to modify Surveillance Requirement 4.8.1.1.2.g.3 
of the Millstone Unit No. 3 Technical Specifications by changing the 
acceptable transient voltage to 5000 volts from 4784 volts. This change 
will permit the DG full load rejection tests to be performed at 
realistic plant conditions using a power factor that will envelope the 
calculated power factor during the worst kW loading conditions. The 
transient voltage of 5000 volts is within the normal design limits of 
the DGs.
    The proposed change does not alter the intent of the surveillance, 
does not involve any physical changes to the plant, does not alter the 
way any structure, system, or component functions, and does not modify 
the manner in which the plant is operated. As such, the proposed change 
to Surveillance Requirement 4.8.1.1.2.g.3 will not degrade the 
capability of the DGs to perform their intended safety function, and 
will not reduce the availability of the DGs. Actually, the proposed 
change will increase the effectiveness of the full load rejection 
tests, because the DGs will be tested in a configuration that is closer 
to the design basis conditions.
Pressure Test of the DG Fuel Oil System
    The DG fuel oil system is classified as an ASME Code Class 3 system 
in accordance with the guidance of Regulatory Guide 1.26, ``Quality 
Group Classification and Standards for
Water-, Steam-, and Radioactive-waste Components of Nuclear Power 
Plants.'' Surveillance Requirement 4.0.5 requires the testing of ASME 
Class 1, 2, and 3 components in accordance with Section XI of the ASME 
Code. Surveillance Requirement 4.8.1.1.2.i.2 is redundant to the ASME 
Section XI pressure test requirements of Surveillance Requirement 
4.0.5. Additionally, the DG fuel oil tank cannot be tested in the 
configuration required by Surveillance Requirement 4.8.1.1.2.i.2, 
because the tanks are vented to the atmosphere and the vent cannot be 
isolated. Therefore, NNECO is proposing to delete Surveillance 
Requirement 4.8.1.1.2.i.2.
    The proposed change does not modify the manner in which the DGs 
respond to an accident. Also, the proposed change does not reduce the 
reliability of the DGs.
Conclusion
    Based on the above, the proposed changes to Surveillance 
Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No. 
3 Technical Specifications 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 
from any previously analyzed.
DG Full-Load Rejection Test
    The DGs are required to operate in response to a loss of offsite 
power. Their failure cannot initiate an accident. Additionally, the 
proposed change to Surveillance Requirement 4.8.1.1.2.g.3 does not 
affect the operation or response of any plant structure, system, or 
component, and it does not introduce any new failure mechanisms.
Pressure Test of the DG Fuel Oil System
    The proposed change to Surveillance Requirement 4.8.1.1.2.i.2 does 
not affect the design or function of the DG fuel oil system. Failure of 
the DG fuel oil system would not initiate an accident.
Conclusion
    Based on the above, the proposed changes to Surveillance 
Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No. 
3 Technical Specifications will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. Involve a significant reduction in the margin of safety.
DG Full-Load Rejection Test
    NNECO is proposing to modify Surveillance Requirement 4.8.1.1.2.g.3 
of the Millstone Unit No. 3 Technical Specifications by changing the 
acceptable transient voltage to 5000 volts from 4784 volts. The intent 
of the proposal is to permit the DG full load rejection tests to be 
conducted at conditions which simulate design basis conditions.
    The proposed change does not alter the intent of the surveillance, 
does not involve any physical changes to the plant, does not alter the 
way any structure, system, or component functions, and does not modify 
the manner in which the plant is operated. As such, the proposed change 
to Surveillance Requirement 4.8.1.1.2.g.3 will not degrade the ability 
of the DGs to perform their intended safety function, and will not 
reduce the availability of the DGs.
    The bases of Technical Specification 3/4.8, ``Electrical Power 
Systems,'' state that the operability of the AC and DC power systems 
and associated distribution systems ensure that sufficient power will 
be available to supply the safety related equipment required for safe 
shutdown and for the mitigation of transients. The proposed change to 
the surveillance requirement will increase the effectiveness of the 
full load rejection tests.
    This will ensure the operability of the DGs. Operable DGs ensure 
that the assumptions for the bases of the Millstone Unit No. 3 
Technical Specifications are not affected.
Pressure Test of the DG Fuel Oil System
    NNECO is proposing to delete Surveillance Requirement 4.8.1.1.2.i.2 
from the Millstone Unit No. 3 Technical Specifications. This 
surveillance requirement is redundant to the requirements of 
Surveillance Requirement 4.0.5 which invokes ASME Section XI. 
Additionally, the fuel oil system cannot be tested to the requirements 
of Surveillance Requirement 4.8.1.1.2.i.2 because the DG fuel oil tanks 
are vented to the atmosphere and this vent path cannot be isolated.
    Millstone Unit No. 3 will include the DG fuel oil system pressure 
test as an augmented inspection within the Inservice Inspection 
program. Inspections will be performed in compliance with the 
requirement of the 1983 Edition of ASME Section XI, Table IWD-2500-1, 
``Test and Examination Categories.'' Testing (i.e., a system 
hydrostatic test) in accordance with ASME Section XI will provide 
equivalent assurance of tank and piping integrity.
Conclusion
    Based on the above, the proposed changes to Surveillance 
Requirements 4.8.1.1.2.g.3 and 4.8.1.1.2.i.2 of the Millstone Unit No. 
3 Technical Specifications do not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
[[Page 8753]] 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.

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

    Date of amendment request: January 18, 1995.
    Description of amendment request: The proposed changes to the 
technical specifications will increase the minimum required boron 
concentration in the boric acid tank (BAT) from 6300 ppm to 6600 ppm.
    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 (SHC), which is presented below:

* * * The proposed changes do not involve an SHC because the changes 
would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The change affects the minimum required boron concentration in the 
BAT. Changes in the tank's boron concentration will not affect the 
probability of any plant accident.
    An increase in the minimum BAT concentration of 6600 ppm was 
recommended by Westinghouse based on their Cycle 6 BORDER evaluation. 
The BORDER evaluation conservatively determines the ability to maintain 
shutdown margin when the plant is taken from an initial operating 
condition of Mode 1 or 2 to a final condition of Mode 5 or 6 using an 
assumed minimum BAT concentration. Therefore, the ability to maintain 
shutdown margin is assured and the change will not adversely affect the 
consequences of any plant accident.
    2. Create the possibility of a new or different kind of accident 
from any Previously Analyzed.
    The change conservatively increases the minimum required boron 
concentration in the BAT from 6300 ppm to 6600 ppm. There is no impact 
on the operability of plant systems or equipment. Therefore, the change 
does not create a malfunction that is different from those previously 
evaluated.
    3. Involve a significant reduction in the margin of safety.
    The proposed increase in the minimum boron concentration in the BAT 
provides conservatism in the calculated shutdown margin for Millstone 
Unit No. 3. The change does not adversely affect any equipment credited 
in the safety analysis. Also, the change does not adversely affect the 
probability or consequences of any plant accident, including the 
calculated PCT [peak clad temperature] or offsite doses. Therefore, 
there is no impact on the margin of safety as specified in 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: 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 10, 1995.
    Description of amendment requests: The proposed amendments would 
revise the Prairie Island Event Monitoring Instrumentation Technical 
Specifications and associated Bases to conform to Standard Technical 
Specifications for post-accident monitoring.
    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 primary purpose of post accident monitoring instrumentation is 
to display plant variables that provide information to the control room 
operators during accident situations. Plant instrumentation was 
evaluated for importance for this function when Regulatory Guide 1.97 
[''Instrumentation for Light Water Cooled Nuclear Power Plants to 
Assess Plant Conditions During and Following an Accident''] 
classifications were determined. The Prairie Island Regulatory Guide 
1.97 classification of instruments was previously approved by the NRC 
on October 18, 1985. This amendment request proposes to base Prairie 
Island Technical Specifications on the results of the Regulatory Guide 
1.97 evaluation in accordance with the guidance of the industry 
standard.
    Revising the allowed outage time for these instruments will not 
significantly increase the probability or consequences of an accident 
since these instruments do not initiate automatic actions, there are 
available backup indications and the probability of an event requiring 
these instruments to be operable is very low.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected by any of the proposed 
amendments.
    2. The proposed amendment[s] will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The license amendment request proposes to add instruments to the 
Technical Specifications which have been previously determined to be 
important for post accident monitoring, and to remove instruments from 
Technical Specifications which have been previously determined to be 
less important for post accident monitoring. This amendment ensures the 
control room operators are provided with the instrumentation required 
to properly manage an accident situation.
    Therefore, based on the above considerations, the possibility of a 
new or different kind of accident from any accident previously 
evaluated would not be created.
    3. The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The post accident monitoring functions do not initiate any 
automatic actions. The instrumentation to be added to the Event 
Monitoring Instrumentation Table was previously recognized through the 
Regulatory Guide 1.97 evaluation process as important for post accident 
monitoring and would be relied upon if there were an event without this 
license amendment. Instrumentation to be removed from Technical 
Specifications was previously recognized to be less 
[[Page 8754]] important and would not be relied upon very much in an 
event. Overall, with the trade-off of adding and deleting 
instrumentation, the margin of safety will not be significantly 
affected.
    The proposed license amendment will increase the allowed outage 
time for most of the instruments. Again, these instruments do not 
provide automatic actions, they provide indications for monitoring post 
accident conditions. All of the instruments have backup or 
corroborating indications which could be relied upon if the Technical 
Specifications instruments were inoperable. Also, an event requiring 
use of these instruments has a very low probability. For these reasons 
the proposed changes in allowed outage time will not result in a 
significant reduction in the margin of safety.
    For these same reasons, the proposed changes in radiation 
instrument surveillance requirements will not significantly reduce the 
margin of safety.
    Overall, a significant reduction in the margin of safety would not 
result from this license 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: John N. Hannon.

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

    Date of amendment request: January 25, 1995.
    Description of amendment request: The proposed Technical 
Specification change would replace a specific requirement for the 
frequency of Type A tests with a general requirement to perform Type A 
tests. The proposed amendment would change Surveillance Requirement 
4.6.1.2.a. Specifically, the change would require the performance of 
Type A tests (overall containment integrated leak rate tests (ILRTs)) 
at intervals as specified in 10 CFR 50, Appendix J, instead of on a 
specific schedule for performance of ILRTs of ``40 plus or minus 10 
months.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92(c). The NRC staff's review is 
presented below.
    A. The change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated (10 CFR 
50.92(c)(1)) because the proposed change merely replaces a prescriptive 
schedule for performing ILRTs with a requirement to conduct the ILRTs 
on a schedule consistent with the Commission's regulations. The change 
does not alter the methodology, frequency, or acceptance criteria for 
ILRTs, does not affect the design basis of the containment, and does 
not change the post-accident response of the containment.
    B. The change does not create the possibility of a new or different 
kind of accident from any accident previously evaluated (10 CFR 
50.92(c)(2)) because the change does not affect the manner by which the 
facility is operated and does not make any changes to existing plant 
structures, systems, or components. The proposed change merely replaces 
a prescriptive schedule for performing ILRTs with a requirement to 
conduct the ILRTs on a schedule consistent with the Commission's 
regulations.
    C. The change does not involve a significant reduction in a margin 
of safety (10 CFR 50.92(c)(3)) because the proposed change does not 
affect the manner by which the facility is operated or involve changes 
to equipment or features which affect the operational characteristics 
of the facility.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, NH 03833.
    Attorney for licensee: Thomas Dignan, Esquire, Ropes & Gray, One 
International Place, Boston, MA 02110-2624.
    NRC Project Director: Phillip F. McKee.

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 7, 1994.
    Description of amendments request: The amendments would provide a 
permanent voltage-based steam generator tube repair criteria for both 
units. This criteria is based on the guidance contained in the NRC 
Proposed Generic Communication (Generic Letter 94-XX), ``Voltage-Based 
Repair Criteria for the Repair of Westinghouse Steam Generator Tubes 
Affected by Outside Diameter Stress Corrosion Cracking,'' that was 
issued for public comment in the Federal Register (59 FR 41520) on 
August 12, 1994. The licensee's submittal also includes responses to 
and identifies exceptions taken to the draft Generic Letter. The 
significant exceptions are: (1) The requirement to reinspect all tubes 
if bobbin probe wear exceeds 15%; (2) the 1 x 10-2 limit on the 
calculated conditional burst probability; and (3) the need to pull 
additional steam generator tubes to evaluate the current condition of 
the steam generator tubes. In addition, the operational leakage 
requirement for Unit 2 will be modified to reduce the total allowable 
primary-to-secondary leakage for any steam generator from 500 gallons 
per day (gpd) to 150 gallons per day.
    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 Farley units in accordance with the proposed 
license amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Testing of model boiler specimens for free standing tubes at room 
temperature conditions shows burst pressures as high as approximately 
5000 psi for indications of outer diameter stress corrosion cracking 
with voltage measurements as high as 26.5 volts. Burst testing 
performed on pulled tubes with up to 7.5 volt indications show burst 
pressures in excess of 5900 psi at room temperature. As stated earlier, 
tube burst criteria are inherently satisfied during normal operating 
conditions by the presence of the tube support plate. Furthermore, 
correcting for the effects of temperature on material properties and 
minimum strength levels (as the burst testing was [[Page 8755]] done at 
room temperature), tube burst capability significantly exceeds the R.G. 
[Regulatory Guide] 1.121 criterion requiring the maintenance of a 
margin of 1.43 times the steam line break pressure differential on tube 
burst if through-wall cracks are present without regard to the presence 
of the tube support plate. Considering the existing data base, this 
criterion is satisfied with bobbin coil indications with signal 
amplitudes over twice the 2.0 volt voltage-based repair criteria, 
regardless of the indicated depth measurement. This structural limit is 
based on a lower 95% confidence level limit of the data. The 2.0 volt 
criterion provides an extremely conservative margin of safety to the 
structural limit considering expected growth rates of outside diameter 
stress corrosion cracking at Farley. Alternate crack morphologies can 
correspond to a voltage so that a unique crack length is not defined by 
a burst pressure to voltage correlation. However, relative to expected 
leakage during normal operating conditions, no field leakage has been 
reported from tubes with indications with a voltage level of under 7.7 
volts for 3/4 inch tube which correlates to 10 volts for 7/8 inch 
tubing (as compared to the 2.0 volt proposed voltage-based tube repair 
limit). Thus, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident.
    Relative to the expected leakage during accidents (sic) condition 
loadings, the accidents that are affected by primary-to-secondary 
leakage and steam release to the environment are Loss of External 
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station 
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator Tube 
Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a Control 
Rod Drive Mechanism Housing. Of these, the Major Secondary System Pipe 
Failure is the most limiting for Farley in considering the potential 
for off-site doses. The offsite doses analyses for the other events 
which model primary-to-secondary leakage and steam releases from the 
secondary side to the environment assume that the secondary side 
remains intact. The steam generator tubes are not subjected to a 
sustained increase in differential pressure, as is the case following a 
steam line break event. This increase in differential pressure is 
responsible for the postulated increase in leakage and associated 
offsite doses following a steam line break event. In addition, the 
steam line break event results in a bypass of containment for steam 
generator leakage. Upon implementation of the voltage-based repair 
criteria, it must be verified that the expected distributions of 
cracking indications at the tube support plate intersections are such 
that primary-to-secondary leakage would result in site boundary dose 
within the current licensing basis. Data indicate that a threshold 
voltage of 2.8 volts could result in through-wall cracks long enough to 
leak at steam line break conditions. Applications of the proposed 
repair criteria requires that the current distribution of a number of 
indications versus voltage be obtained during the refueling outages. 
The current voltage is then combined with the rate of change in voltage 
measurement and a voltage measurement uncertainty to establish an end 
of cycle voltage distribution and, thus, leak rate during steam line 
break pressure differential. The leak rate during a steam line break is 
further increased by a factor related to the probability of detection 
of the flaws. If it is found that the potential steam line break 
leakage for degraded intersections planned to be left in service 
coupled with the reduced specific activity levels allowed result in 
radiological consequences outside the current licensing basis, then 
additional tubes will be plugged or repaired to reduce steam line break 
leakage potential to within the acceptance limit. Thus, the 
consequences of the most limiting design basis accident are constrained 
to present licensing basis limits.
    (2) The proposed license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Implementation of the proposed voltage-based tube support plate 
elevation steam generator tube repair criteria does not introduce any 
significant changes to the plant design basis. Use of the criteria does 
not provide a mechanism that could result in an accident outside of the 
region of the tube support plate elevations. Neither a single or 
multiple tube rupture event would be expected in steam generator in 
which the repair criteria have been applied during all plant 
conditions. The bobbin probe signal amplitude repair criteria are 
established such that operational leakage or excessive leakage during a 
postulate steam line break condition is not anticipated. Southern 
Nuclear has previously implemented a maximum leakage limit of 140/150 
gpd (Unit 1/Unit 2) per steam generator. The R.G. 1.121 criterion for 
establishing operational leakage limits that require plant shutdown are 
based upon leak-before-break considerations to detect a free span crack 
before potential tube rupture. The 140/150 gpd limit provides for 
leakage detection and plant shutdown in the event of the occurrence of 
an unexpected single crack resulting in leakage that is associated with 
the longest permissible crack length. R.G. 1.121 acceptance criteria 
for establishing operating leakage limits are based on leak-before-
break considerations such that plant shutdown is initiated if the 
leakage associated with the longest permissible crack is exceeded. The 
longest permissible crack is the length that provides a factor safety 
of 1.43 against bursting at steam line break pressure differential. A 
voltage amplitude of approximately 9 volts for typical outside diameter 
stress corrosion cracking corresponds to meeting this tube burst 
requirement at the 95% prediction interval on the burst correlation. 
Alternate crack morphologies can correspond to a voltage so that a 
unique crack length is not defined by the burst pressure versus voltage 
correlation. Consequently, typical burst pressure versus throughwall 
crack length correlations is used below to define the ``longest 
permissible crack'' for evaluating operating leakage limits.
    The single through-wall crack lengths that results in tube burst at 
1.43 times steam line break pressure differential and steam line break 
conditions are about 0.53 inch and 0.84 inch, respectively. Normal 
leakage for these crack lengths would range from about 0.4 gallons per 
minute to 4.5 gallons per minute, respectively, while lower 95% 
confidence level leak rates would range from about 0.06 gallons per 
minute to 0.6 gallons per minute, respectively.
    An operating leak rate of 140/150 gpd per steam generator has been 
implemented. This leakage limit provides for detection of 0.4 inch long 
cracks at nominal leak rates and 0.6 inch long cracks at the lower 95% 
confidence level leak rates. Thus, the 140/150 gpd limit provides for 
plant shutdown prior to reaching critical crack lengths for steam line 
break conditions at leak rates less than 95% confidence level and for 
three times normal operating pressure differential at less than nominal 
leak rates.
    Considering the above, the implementation of voltage-based plugging 
criteria will not create possibility of a new or different kind of 
accident from any previously evaluated.
    (3) The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the voltage-based tube support plate elevation repair 
criteria is demonstrated to maintain steam [[Page 8756]] generator tube 
integrity commensurate with the requirements of R.G. 1.121. R.G. 1.121 
describes a method acceptable to the NRC staff for meeting GDCs 
[General Design Criteria] 2, 14, 15, 31, and 32 by reducing the 
probability of the consequences of steam generator tube rupture. This 
is accomplished by determining the limiting conditions of degradation 
of steam generator tubing, as established by inservice inspection, for 
which tubes with unacceptable cracking should be removed from service. 
Upon implementation of the criteria, even under the worst case 
conditions, the occurrence of outside diameter stress corrosion 
cracking at the tube support plant elevations is not expected to lead 
to a steam generator tube rupture event during normal or faulted plant 
conditions. The most limiting effect would be a possible increase in 
leakage during a steam line break event. Excessive leakage during a 
steam line break event, however, is precluded by verifying that, once 
the criteria are applied, the expected end of cycle distribution of 
crack indications at the tube support plate elevations would result in 
minimal, and acceptable primary to secondary leakage during the event 
and, hence, help to demonstrate radiological conditions are less than 
an appropriate fraction of the 10 CFR 100 guideline.
    The margin to burst for the tubes using the voltage-based repair 
criteria is comparable to that currently provided by existing technical 
specifications.
    In addressing the combined effects of LOCA [loss-of-coolant 
accident] + SSE [safe shutdown earthquake] on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is the 
case as the tube support plates may become deformed as a result of 
lateral loads at the wedge supports at the periphery of the plate due 
to either the LOCA rarefaction wave and/or SSE loadings. Then, the 
resulting pressure differential on the deformed tubes may cause some of 
the tubes to collapse.
    There are two issues associated with steam generator tube collapse. 
First, the collapse of steam generator tubing reduces the RCS [reactor 
coolant system] flow area through the tubes. The reduction in flow area 
increases the resistance to flow of steam from the core during a LOCA 
which, in turn, may potentially increase Peak Clad Temperature (PCT). 
Second, there is a potential the partial through-wall cracks in tubes 
could progress to through-wall cracks during tube deformation or 
collapse or that short through-wall indications would leak at 
significantly higher leak rates than included in the leak rate 
assessments.
    Consequently, a detailed leak-before-break analysis was performed 
and it was concluded that the leak-before-break methodology (as 
permitted by GDC 4) is applicable to the Farley reactor coolant system 
primary loops and, thus, the probability of breaks in the primary loop 
piping is sufficiently low that they need not be considered in the 
structural design basis of the plant. Excluding breaks in the RCS 
primary loops, the LOCA loads from the large branch line breaks were 
analyzed at Farley and were found to be of insufficient magnitude to 
result in steam generator tube collapse or significant deformation.
    Regardless of whether or not leak-before-break is applied to the 
primary loop piping at Farley, any flow area reduction is expected to 
be minimal (much less than 1%) and PCT margin is available to account 
for this potential effect. Based on analyses' results, no tubes near 
wedge locations are expected to collapse or deform to the degree that 
secondary to primary in-leakage would be increased over current 
expected levels. For all other steam generator tubes, the possibility 
of secondary-to-primary leakage in the event of a LOCA + SSE event is 
not significant. In actuality, the amount of secondary-to-primary 
leakage in the event of a LOCA + SSE is expected to be less than that 
previously allowed, i.e., 500 gpd per steam generator. Furthermore, 
secondary-to-primary in-leakage would be less than primary-to-secondary 
leakage for the same pressure differential since the cracks would tend 
to tighten under a secondary-to-primary pressure differential. Also, 
the presence of the tube support plate is expected to reduce the amount 
of in-leakage.
    Addressing the R.G. 1.83 considerations, implementation of the tube 
repair criteria is supplemented by 100% inspection requirements at the 
tube support plate elevations having outside diameter stress corrosion 
cracking indications, reduced operating leakage limits, eddy current 
inspection guidelines to provide consistency in voltage normalization, 
and rotating pancake coil inspection requirements for the larger 
indications left in service to characterize the principle degradation 
mechanism as outside diameter stress corrosion cracking.
    As noted previously, implementation of the tube support plate 
elevation repair criteria will decrease the number of tubes that must 
be taken out of service with tube plugs or repaired. The installation 
of steam generator tube plugs or tube sleeves would reduce the RCS flow 
margin, thus implementation of the voltage-based repair criteria will 
maintain the margin of flow that would otherwise be reduced through 
increased tube plugging or sleeving.
    Considering 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 Final Safety Analysis Report or any 
bases of the plant 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: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: William H. Bateman.

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

    Date of amendments request: January 9, 1995.
    Description of amendments request: The requested changes to the 
Technical Specifications (TS) would implement the recommended changes 
from Generic Letter 93-05, ``Line Item Technical Specification 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation.'' Specifically, the amendments would implement TS 
changes corresponding to the following GL 93-05 line-item improvement 
issues: Control Rod Movement Test for Pressurized Water Reactors, 
Radiation Monitors, Surveillance of Boron Concentration in the 
Accumulator/Safety Injection/Core Flood Tank, Containment Spray System, 
Hydrogen Recombiner, and Special Test Exemptions.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes do not involve any change to the configuration or 
method [[Page 8757]] of operation of any plant equipment used to 
mitigate the consequences of an accident. The changes to the 
surveillance requirements will result in an overall improvement in 
plant safety by reducing the likelihood of plant trips and subsequent 
challenges to safety systems, decreasing equipment degradation due to 
excessive testing, reducing radiation exposure to plant personnel, 
increasing the availability of safety related equipment, and 
eliminating an unnecessary burden on plant personnel. Therefore, the 
proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated. The 
proposed changes do not involve any change to the configuration or 
method of operation of any plant equipment used to mitigate the 
consequences of an accident. The relaxation of surveillance tests 
curtails the excessive amount of testing that increases wear on the 
equipment and reduces the likelihood of plant trips and subsequent 
challenges to safety systems. The relaxation also increases the 
availability of safety related equipment. Accordingly, no new failure 
modes have been defined for any plant system or component important to 
safety nor has any new limiting failure been identified as a result of 
the proposed changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes eliminate an unnecessary burden 
without compromising protection for public health and safety. The 
proposed changes were generically analyzed by the NRC as part of a 
comprehensive study and presented in NUREG-1366 ``Improvement to 
Technical specifications (sic) Surveillance Requirements.'' The NRC 
concluded that while some testing at power is essential to verify 
equipment and system operability, safety can be improved, equipment 
degradation decreased, and unnecessary personnel burden relaxed by 
reducing the amount of testing at power. SNC has analyzed plant 
operations and made a comparison with the criteria stated in NUREG-1366 
for the line-item improvements contained in this request and has found 
the NUREG-1366 basis to be consistent with the Farley design and 
operation experience. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302.
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201.
    NRC Project Director: William H. Bateman.

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: December 6, 1994.
    Description of amendment request: The proposed change to Technical 
Specification 3/4.1.3.2 will delete Surveillance Requirement (SR) 
4.1.3.2.2, that presently requires, every 31 days, the movement of at 
least 2% of its height for each Axial Power Shaping Rod not fully 
withdrawn. The proposed amendment would also change the surveillance 
intervals for the following Technical Specifications (TS) in accordance 
with the guidance of Generic Letter 93-05, ``Line Item Technical 
Specifications Improvements to Reduce Surveillance Requirements For 
Testing During Power Operation,'' and NUREG-1366, ``Improvements to 
Technical Specifications Surveillance Requirements:''
    1. TS 4.1.3.2 for the Movable Control Assemblies ``Group Height--
Safety and Regulating Rod Groups,'' will relax testing requirements 
from at least once every 31 days to every 92 days.
    2. TS 4.4.6.2, for ``Operational Leakage,'' relaxes the requirement 
to leakage test RCS pressure isolation valves prior to MODE 2 whenever 
the plant has been in COLD SHUTDOWN for 72 hours to whenever the plant 
has been in COLD SHUTDOWN for 7 days.
    3. SR 4.5.2.c.2 for TS 4.5.2, ``ECCS Subsystems--Tavg equal to or 
greater than 280 deg. F,'' relaxes the inspection requirements for 
ensuring no debris in containment from ``at the completion of each 
containment entry'' to ``at least once daily.''
    4. TS 4.6.2.1.d, for the ``Containment Spray System,'' relaxes the 
SR to perform an air or smoke flow test through the spray header and 
nozzles from once per 5 years to once per 10 years.
    5. TS 4.10.4.2 for ``Special Test Exceptions Shutdown Margin'' 
relaxes the SR interval for testing rod insertion capability prior to 
reducing shutdown margin from 24 hours to 7 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC 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 change does not involve a significant increase in the 
probability of an accident previously evaluated nor does it involve a 
significant increase in the consequences of an accident previously 
evaluated because no change is being made to any accident initiator and 
no accident conditions or assumptions used in evaluating the 
radiological consequences of an accident are changed. Relaxation of 
surveillance requirements is in accordance with GL 93-05, NUREG-1366, 
and is compatible with plant operating experience. Deletion of SR 
4.1.3.2 is consistent with NUREG-1430, ``Improved Standard Technical 
Specifications for B&W Plants.'' No credit is taken in any accident 
analysis or mitigation requirements for the Axial Power Shaping Rod 
Group.
    (2) The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not create the possibility of any new or 
different kind of accident from any accident previously evaluated 
because no new accident initiators or assumptions are introduced by 
these proposed changes. Relaxation of SRs as discussed in GL 93-05 was 
evaluated as reducing equipment degradation with no increase in safety 
consequences consistent with the maintenance of plant specific 
reliability of the equipment and systems affected. Deletion of the SR 
to move the Axial Power Shaping Rod Group does not affect the 
requirement to verify rod position, and there is no credit taken for 
movement of these rods to mitigate an accident.
    (3) The proposed changes do not result in a significant reduction 
in the margin of safety. [[Page 8758]] 
    The changes do not involve a significant reduction in the margin of 
safety, because the proposed changes affect only surveillance 
requirements, do not affect the function of the components and systems 
involved, and do not decrease the estimated equipment or system 
reliability.
    Based on the NRC staff analysis, 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: December 6, 1994.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.0.5, ``Applicability'' and its 
associated Bases; TS 3/4.1.2.3, ``Reactivity Control Systems--Makeup 
Pump--Shutdown; TS 3/4.1.2.4, ``Reactivity Control Systems--Makeup 
Pump--Operating; TS 3/4.1.2.6, Reactivity Control Systems--Boric Acid 
Pump--Shutdown; and TS 3/4.1.2.7, ``Reactivity Control System--Boric 
Acid Pumps--Operating.'' The proposed change would replace the specific 
monthly surveillance requirements associated with the makeup pumps and 
boric acid pumps with a surveillance requirement referencing TS 4.0.5, 
which references Section XI of the American Society of Mechanical 
Engineers Boiler and Pressure Vessel Code for quarterly pump testing 
requirements. The proposed change to TS 4.0.5 and its associated Bases 
would revise the requirement regarding the NRC's approval of relief 
requests to be in accordance with the NRC Staff's recommendation 
contained in NUREG-1482, ``Guidelines for Inservice Testing at Nuclear 
Power Plants.'' Additionally, TS 4.0.5.a.2 which describes historical 
requirements for inservice inspection and testing would be deleted and 
TS 4.0.5.a.1 would be renumbered as TS 4.0.5.a.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the NRC Staff has 
performed an 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.
    Operation of the Davis-Besse Nuclear Power Station, Unit No. 1, in 
accordance with these changes, would not involve a significant increase 
in the probability of an accident previously evaluated because no 
accident initiators, conditions, or assumptions are affected by the 
proposed changes to replace the specific monthly surveillance 
requirements for the makeup and boric acid pumps with surveillance 
requirements referencing TS 4.0.5 (ASME Boiler and Pressure Vessel Code 
Section XI requirements) and to delete wording regarding NRC approval 
of relief requests. The changes do not involve a significant increase 
in the consequences of an accident previously evaluated, because no 
accident conditions or assumptions are affected that would increase the 
radiological consequences of a previously evaluated accident.
    (2) The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed changes do not result in any new accident initiators 
nor do they alter any accident scenarios. The changes do not create the 
possibility of a different kind of accident from any accident 
previously evaluated, because the surveillance requirements for the 
makeup and boric acid pumps only affect the testing of existing 
components, systems, and functions, and do not introduce any new 
requirements.
    (3) The proposed changes do not result in a significant reduction 
in the margin of safety.
    The proposed changes do not reduce or adversely affect the 
capabilities or reliability of any plant structures, systems or 
components. Relaxation of the surveillance testing interval for the 
boric acid and makeup pumps and modifying the testing requirements is 
consistent with previous NRC guidance.
    Based on this NRC staff evaluation, 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.

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

    Date of amendment request: January 13, 1995.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) by relocating Tables 3.3-2, 
``Reactor Trip System Instrumentation Response Times,'' and 3.3-5, 
``Engineered Safety Features Response Times,'' to FSAR Chapter 16, 
Section 16.3. The Bases discussion specific to Table 3.3-5 would also 
be relocated to FSAR Section 16.3.
    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 revision does not involve a significant hazards 
consideration because operation of Callaway Plant with this change 
would not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Overall protection system performance will remain within the bounds 
of the accident analyses documented in FSAR Chapter 15, WCAP-10961-P, 
and WCAP-11883 since no changes to the response times or measurement 
interval are proposed.
    The RTS and ESFAS will continue to function in a manner consistent 
with the above analysis assumptions and the plant design basis. As 
such, there will be no degradation in the performance of nor an 
increase in the number of challenges to equipment assumed to function 
during an accident situation.
    These Technical Specification revisions do not involve any hardware 
changes nor do they affect the probability of any event initiators. 
There will be no change to normal plant operating parameters or 
accident mitigation capabilities. Therefore, there will be no increase 
in the probability or consequences of any accident occurring due to 
these changes.
    (2) Create the possibility of a new or different kind of accident 
from any previously evaluated. [[Page 8759]] 
    As discussed above, there are no hardware changes associated with 
these Technical Specification revisions nor are there any changes in 
the method by which any safety-related plant system performs its safety 
function. The normal manner of plant operation is unaffected.
    No new accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures are introduced as a result of 
these changes. There will be no adverse effect or challenges imposed on 
any safety-related system as a result of these changes. Therefore, the 
possibility of a new or different type of accident is not created.
    (3) Involve a significant reduction in a margin of safety.
    No response time changes are proposed in this amendment 
application; only the document where these limits are listed will be 
changed. There will be no effect on the manner in which safety limits 
or limiting safety system settings are determined nor will there be any 
effect on those plant systems necessary to assure the accomplishment of 
protection functions. There will be no impact on DNBR limits, FQ, 
F-delta-H, LOCA PCT, peak local power density, or any other margin of 
safety.
    Based upon the preceding information, it has been determined that 
the proposed changes to the Technical Specifications do not involve a 
significant increase in the probability or consequences of an accident 
previously evaluated, create the possibility of a new or different kind 
of accident from any accident previously evaluated, or involve a 
significant reduction in a margin of safety. Therefore, it is concluded 
that the proposed changes meet the requirements of 10CFR50.92(C) [sic] 
and 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: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037.
    NRC Project Director: Leif J. Norrholm.

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

    Date of amendment request: December 8, 1994.
    Description of amendment request: The proposed amendment would 
change Standby Gas Treatment Power Supply Requirements during refueling 
operations.
    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:
SGTS [Standby Gas Treatment System] DURING REFUELING OPERATIONS 
(Specification 3.7.B.1, 3.7.B.3)
    1. The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously evaluated. 
The Standby Gas Treatment System (SGTS) is not the initiator of any 
accident. SGTS may be required to operate for a design basis loss of 
coolant accident or for a refueling accident in order to mitigate the 
consequences of said accident by providing a filtered exhaust path to 
minimize the potential release of radioactive material to the environs. 
The proposed amendment does not reduce or change the operational 
requirements for the SGTS for an accident. The proposed amendment now 
clearly defines the operability requirements during refueling 
conditions. The proposed amendment further requires the availability of 
a second auxiliary power supply in the event that an Emergency Diesel 
Generator (EDG) is out of service during refueling operations, not 
currently required. We conclude, therefore, that the proposed amendment 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
The SGTS is not an accident initiator, therefore, the proposed 
amendment will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant reduction 
in a margin of safety. The proposed amendment requires the availability 
of a second auxiliary power supply in the event that an EDG is out of 
service during refueling operations, not currently required. 
Maintaining availability of a specific reliable auxiliary electrical 
power source as an alternative to an EDG in this mode provides 
assurance that SGTS can, if required, be operated without placing undue 
constraints on EDG availability and represents an enhancement that 
increases a margin of safety. We conclude, therefore, that the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    Based on the above discussion, we have determined that this change 
does not constitute a significant hazards consideration as defined in 
10CFR50.92(c).
LABORATORY CARBON SAMPLE ANALYSIS (Specification 3.7.B.2.b)
    1. The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously evaluated. 
The Standby Gas Treatment System (SGTS) is not the initiator of any 
accident. SGTS may be required to operate for a design basis loss of 
coolant accident or for a refueling accident in order to mitigate the 
consequences of said accident by providing a filtered exhaust path to 
minimize the potential release of radioactive material to the environs. 
The proposed amendment does not reduce or change the operational 
requirements for the SGTS for an accident. The proposed amendment now 
clearly defines the operability requirements during the interval 
between sample removal and completion of laboratory analysis.
    2. The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
The SGTS is not an accident initiator, therefore, the proposed 
amendment will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant reduction 
in a margin of safety. The proposed change does not reduce the 
requirements or acceptance criteria for sampling, testing or analysis. 
The proposed change only incorporates into the specification an 
existing clarification which addresses the determination of operability 
during the time between sample removal and completion of laboratory 
analysis. The change provides an explicit time limit consistent with 
current regulatory criteria for completion of analyses.
    Based on the above discussion, we have determined that this change 
does not constitute a significant hazards [[Page 8760]] consideration 
as defined in 10CFR50.92(c).
TORUS VENT MODE (Specification 4.7 B.2.c)
    1. The proposed amendment will not involve a significant increase 
in the probability or consequences of an accident previously evaluated. 
The Standby Gas Treatment System (SGTS) is not the initiator of any 
accident. SGTS may be required to operate for a design basis loss of 
coolant accident or for a refueling accident in order to mitigate the 
consequences of said accident by providing a filtered exhaust path to 
minimize the potential release of radioactive material to the environs. 
The proposed amendment does not reduce or change the operational 
requirements for the SGTS for an accident.
    2. The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously evaluated. 
The SGTS is not an accident initiator, therefore, the proposed 
amendment will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed amendment will not involve a significant reduction 
in a margin of safety. The proposed change will incorporate into the 
specification an existing clarification. Use of the SGTS filters during 
Torus venting results in an insignificant flow through the filters. 
Further, maintaining humidity control prevents any adsorber 
degradation. Past sample testing on a six month calendar interval when 
720 hours operating time has not accumulated has shown no detectable 
impact.
    Based on the above discussion, we have determined that this change 
does not constitute a significant hazards consideration as defined in 
10CFR50.92(c).
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Brooks Memorial Library, 224 
Main Street, Brattleboro, Vermont 05301.
    Attorney for licensee: John A. Ritsher, Esquire, Ropes and Gray, 
One International Place, Boston, Massachusetts 02110-2624.
    NRC Project Director: Walter R. Butler.

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.

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 19, 1994.
    Brief description of amendment: The proposed amendment would revise 
Section 3.10.8 and the associated Bases of the Indian Point Nuclear 
Generating Unit No. 3 Technical Specifications. Specifically, the 
proposed revision would reduce the maximum allowable control rod drop 
time from 2.4 to 1.8 seconds. The change would remove, for testing 
purposes, the allowance for a seismic event (0.6 seconds), which had 
been integral to the 2.4 second safety analysis basis. Since a seismic 
event cannot be simulated during the rod drop time test, the more 
conservative testing acceptance criteria value of 1.8 seconds is needed 
to ensure that the plant is within its design basis. This proposed 
revision will support control rod testing which is required during 
startup from the current outage.
    Date of publication of individual notice in Federal Register: 
January 20, 1995 (60 FR 4203).
    Expiration date of individual notice: February 21, 1995.
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

Notice of Issuance of Amendments to Facility Operating Licenses

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

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

    Date of application for amendments: November 30, 1994.
    Brief description of amendments: These amendments relocate Table 
3.3-2, ``Reactor Protective Instrumentation Response Times,'' and Table 
3.3-5, ``Engineered Safety Features Response Times,'' of TS 3/4.3.1 and 
3/4.3.2, respectively, to the Palo Verde Updated Final Safety Analysis 
Report (UFSAR) in accordance with the guidance provided in Generic 
Letter 93-08. In addition, the amendments make administrative changes 
to two previous TS amendment requests to maintain consistency with the 
deletion of Tables 3.3-2 and 3.3-5. The amendments also delete an 
obsolete footnote on page 3/4 3-17 of the Palo Verde Unit 2's TS. 
[[Page 8761]] 
    Date of issuance: February 3, 1995.
    Effective date: February 3, 1995.
    Amendment Nos.: 88, 75 and 59.
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
496) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004.

Carolina Power & Light Company, et al., Docket Nos. 50-325 and 50-324, 
Brunswick Steam Electric Plant, Units 1 and 2, Brunswick County, North 
Carolina

    Date of application for amendments: November 16, 1994.
    Brief description of amendments: The proposed amendments change the 
Technical Specifications to revise the wording for the containment 
integrated leakage rate testing in Section 3/4.6.1.2 to make it 
consistent with the requirements of the BWR-4 Improved Standard 
Technical Specifications (NUREG-1433).
    Date of issuance: January 26, 1995.
    Effective date: January 26, 1995.
    Amendment Nos.: 173 and 204.
    Facility Operating License Nos. DPR-71 and DPR-62.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65810).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 26, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.

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

    Date of application for amendments: June 13, 1994, as supplemented 
on October 7, 1994.
    Brief description of amendments: The amendments revise the 
administrative controls in Section 6 of the technical specifications 
(TS). The changes include: (1) a change to the submittal frequency of 
the Radiological Effluent Release Report from semiannually to annually; 
(2) changes to the Shift Technical Advisor (STA) description; (3) a 
clarification of the Shift Engineer responsibilities; and (4) several 
editorial changes.
    Date of issuance: February 2, 1995.
    Effective date: February 2, 1995.
    Amendment Nos.: 69, 69, 59 and 59.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53839).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 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.

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

    Date of application for amendment: May 17, 1993 as supplemented 
October 12, 1994.
    Brief description of amendment: The amendment replaces License 
Condition 2.C.4, relating to the implementation and maintenance of the 
approved Fire Protection Program, in its entirety with a new License 
Condition. In conjunction, with this change, and in accordance with GL 
86-10, Technical Specification provisions related to the Fire 
Protection Program are being deleted and placed in the Updated Final 
Safety Analysis Report.
    Date of Issuance: February 1, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 179.
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 7, 1993 (58 FR 
36432).
    The October 12, 1994, letter provided clarifying 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 February 1, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.

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: August 25, 1994, as supplemented November 
16, 1994.

    Brief description of amendments: The amendments revise Technical 
Specification Table 3.3-4, by revising the ``Trip Setpoint'' and 
``Allowable Value'' for the 4 kV bus undervoltage grid degraded voltage 
relays and the ``Allowable Value'' for the 4 kV undervoltage loss of 
voltage/loss of offsite power relays. This revision was submitted in 
response to a concern identified by the licensee in their Self-
Initiated Technical Audit and during the electrical distribution system 
functional inspection team findings.
    Date of issuance: January 20, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 127 and 121.
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51619).
    The November 16, 1994, letter provided clarifying information that 
did not change the scope of the August 25, 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 January 20, 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.

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

    Date of application for amendment: July 28, 1994.
    Brief description of amendment: This amendment revises Technical 
Specifications 3/4.4.13 to incorporate Low Temperature Overpressure 
Protection requirements similar to those recommended by the NRC staff 
via Generic Letter 90-06. [[Page 8762]] 
    Date of Issuance: January 27, 1995.
    Effective Date: January 27, 1995.
    Amendment No.: 132.
    Facility Operating License No. DPR-67: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42341).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Indian River Junior College 
Library, 3209 Virginia Avenue, Fort Pierce, Florida 34954-9003.

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

    Date of application for amendments: February 3, 1994.
    Brief description of amendments: The amendments relocate the 
requirements of Technical Specification 3/4.7.10, Area Temperature 
Monitoring, to section 16.3 of the VEGP Final Safety Analysis Report 
(FSAR). With this relocation to the FSAR, GPC plans to clarify the 
basis for areas to be monitored and modify these surveillance 
requirements. This change is in accordance with NUREG-1431, ``Standard 
Technical Specifications, Westinghouse Plants.''
    Date of issuance: January 23, 1995.
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 83 and 61.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 2, 1994 (59 
FR 45735).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830.

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 13, 1992.
    Brief description of amendment: The amendment changes the allowable 
primary-to-secondary leakage rate, as specified in License Condition 
2.c.(8)2, from 0.1 gallons per minute (gpm) to 0.2 gpm.
    Date of Issuance: January 31, 1995.
    Effective date: January 31, 1995.
    Amendment No.: 193.
    Facility Operating License No. DPR-50. Amendment revises a License 
Condition.
    Date of initial notice in Federal Register: October 14, 1992 (57 FR 
47137).
    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, Pennsylvania 17105.

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

    Date of amendment request: September 26, 1994.
    Brief description of amendment: The amendment revised Technical 
Specification 3.5.C.1 and 3.5.C.4 to increase the minimum pressure at 
which the high pressure coolant injection system is required to be 
operable from 113 psig to 150 psig.
    Date of issuance: January 25, 1995.
    Effective date: January 25, 1995.
    Amendment No.: 166.
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53841). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 25, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305.

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

    Date of amendment request: December 22, 1994.
    Brief description of amendment: The amendment revised Technical 
Specification 1.0.J, definition of limiting conditions for operation, 
consistent with the guidance provided in NRC Generic Letter 87-09, 
``Sections 3.0 and 4.0 of the Standard Technical Specifications on the 
Applicability of Limiting Conditions for Operation and Surveillance 
Requirements.''
    Date of issuance: February 3, 1995.
    Effective date: February 3, 1995.
    Amendment No.: 168.
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1995 (60 FR 
153).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 3, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305.

Niagara Mohawk Power Corporation, Docket No. 50-220, Nine Mile Point 
Nuclear Station Unit No. 1, Oswego County, New York.

    Date of application for amendment: July 21, 1994.
    Brief description of amendment: The amendment revises Technical 
Specifications 2.2.2, 3.2.8, 4.2.8, and the associated Bases to reduce 
the number of reactor head safety valves required operable from 16 
valves to 9 valves. The setpoints of the valve groups are unchanged by 
this amendment. The amendment requires testing of the safety valves in 
accordance with the approved NMP-1 Inservice Test Program.
    Date of issuance: January 25, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 152.
    Facility Operating License No. DPR-63: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: August 31, 1994 (59 FR 
45027).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 25, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

Niagara Mohawk Power Corporation, Docket No. 50-410, Nine Mile Point

    Nuclear Station, Unit 2, Oswego County, New York.
    Date of application for amendment: October 28, 1994.
    Brief description of amendment: The amendment revises Technical 
[[Page 8763]] Specification (TS) 1.7, ``CORE ALTERATION,'' to state 
that movement or replacement of incore instrumentation is not 
considered to be a CORE ALTERATION and that movement of control rods is 
not considered a CORE ALTERATION provided there are no fuel assemblies 
in the associated core cell. This amendment includes changes to TS 3/
4.9.3, ``Control Rod Position,'' and associated Bases to be consistent 
with the revision to TS 1.7. TS 3/4.9.3 is being revised to require 
that all control rods be inserted only during loading of fuel 
assemblies into the core rather than during CORE ALTERATIONS. These 
changes are consistent with the NRC's, ``Improved Standard Technical 
Specifications,'' (NUREG-1434).
    This amendment also revises Item 1.i.3) of TS Tables 3.3.2-1 and 
4.3.2.1-1 to delete the requirement for Reactor Water Cleanup isolation 
due to actuation of the Standby Liquid Control System (SLCS) in 
OPERATIONAL CONDITION 5. License Amendment No. 48 issued on September 
30, 1993, deleted the requirement for the SLCS to be OPERABLE in 
OPERATIONAL CONDITION 5; however, due to an oversight, Item 1.i.3) and 
associated notations were not deleted from TS Tables 3.3.2-1 and 
4.3.2.1-1 as part of License Amendment No. 48. This amendment corrects 
that oversight.
    Date of issuance: January 20, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 61.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 23, 1994 (59 
FR 60382).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 20, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: November 14, 1994.
    Brief description of amendment: The amendment revises Technical 
Specification 4.5.1.e.2.e) to reduce the leak rate test pressure for 
the Automatic Depressurization System (ADS) nitrogen receiving tanks 
from 385 psig to 365 psig.
    Date of issuance: January 31, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 62.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: December 21, 1994 (59 
FR 65817).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 31, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of amendment request: January 14, 1994, as modified by letter 
dated October 17, 1994.
    Description of amendment request: The amendment revises the 
Appendix A Technical Specifications (TS) to specify the composition of 
the Station Operation Review Committee (SORC) based on experience and 
expertise vice organizational position, to implement a Station 
Qualified Reviewer Program (SQRP), and to revise the time within which 
the Nuclear Safety Audit Review Committee (NSARC) must issue reports 
and minutes.
    The amendment also incorporated a number of editorial changes to 
delete certain items that are no longer applicable; remove 
inconsistencies involving the names of systems, equipment and NSARC 
function, composition, and use of alternates; and correct the value for 
the reactor coolant system volume. Other editorial changes have been 
incorporated for document format consistency. The amendment affects the 
following: TS Sections 1.31, 3.3.3.6, 3.4.1.2, 4.6.3.2, 3.7.1.2, 3/4 
10.6, 5.4.2, 6.3.1, 6.4, 6.7, and 6.8.1.4, and Table 4.3-1.
    Date of issuance: January 26, 1995.
    Effective date: January 26, 1995.
    Amendment No.: 34.
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27057) The licensee's letter dated October 17, 1994, provided 
clarification and minor revision to the application but does not change 
the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 26, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Exeter Public Library, 47 
Front Street, Exeter, NH 03833.

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

    Date of application for amendment: July 22, 1994.
    Brief description of amendment: The amendment revises the Technical 
Specifications to incorporate a different setpoint and transient 
methodology for determining the maximum allowable power range neutron 
flux setpoint. These changes allow Millstone Unit 3 to operate with a 
reduced number of main steam-line safety valves at a reduced power 
level, as determined by the high flux setpoint.
    Date of issuance: January 31, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 102.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    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 January 31, 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.

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 25, 1994.
    Brief description of amendments: These amendments add to the 
Susquehanna, Units 1 and 2, Technical Specifications, isolation signals 
to Table 3.6.3-1 for the containment isolation valves on the sample 
lines for the containment radiation monitoring and 
[[Page 8764]] wetwell sample lines. This change is based on the 
licensee's design change for installation of a new CRM and wetwell 
sample system.
    Date of issuance: January 31, 1995.
    Effective date: January 31, 1995.
    Amendment Nos.: 141 and 111.
    Facility Operating License Nos. NPF-14 and NPF-22. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1994 (59 FR 
63126). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 31, 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 No. 50-352, Limerick Generating 
Station, Unit 1, Montgomery County, Pennsylvania.

    Date of application for amendment: June 10, 1994, as supplemented 
by letter dated December 19, 1994.
    Brief description of amendment: This amendment involves a one-time 
change affecting the Allowed Outage Time (AOT) for the Emergency Sevice 
Water (ESW) system, Residual Heat Removal Service Water (RHRSW) System, 
the Suppression Pool Cooling, the Suppression Pool Spray, and Low 
Pressure Coolant Injection modes of the Residual Heat Removal System, 
and Core Spray System to be extended from 3 and 7 days to 14 days 
during the Unit 2 refueling outage scheduled to begin in January 1995. 
This proposed extended AOT allows adequate time to install isolation 
valves and cross-ties on the ESW and RHRSW Systems to facilitate future 
inspections or maintenance.
    Date of issuance: January 27, 1995.
    Effective date: January 27, 1995.
    Amendment No. 86.
    Facility Operating License No. NPF-39. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 20, 1994 (59 FR 
37077). The December 19, 1994 letter provided clarifying information 
that did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 27, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric Company, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: June 30, 1994.
    Brief description of amendment: This amendment removes the controls 
for a remote shutdown system control valve and the primary containment 
isolation valves from TS Tables 3.3.7.4-1 and 3.6.3-1 respectively, as 
a result of eliminating the steam condensing mode of the Residual Heat 
Removal system.
    Date of issuance: January 27, 1995.
    Effective date: January 27, 1995.
    Amendment No. 47.
    Facility Operating License No. NPF-85. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42343).
    The Commission's related evaluation of the amendment 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.

Philadelphia Electric Company, Docket No. 50-353, Limerick Generating 
Station, Unit 2, Montgomery County, Pennsylvania

    Date of application for amendment: August 27, 1993, supplemented by 
letter dated November 17, 1993.
    Brief description of amendment: The amendment allows an expanded 
operating domain for the Limerick Generating Station (LGS), Unit 2, 
resulting from the implementation of the Average Power Range Monitor--
Rod Block Monitor Technical Specifications/Maximum Extended Load Line 
Limit Analysis. These improvements are a prerequisite for Power Rerate 
Program implementation at Limerick Generating Station, Unit 2.
    Date of issuance: January 27, 1995.
    Effective date: January 27, 1995.
    Amendment No. 48.
    Facility Operating License No. NPF-85. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52992). 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.

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 22, 1994.
    Brief description of amendments: These amendments revise TS 3.1.5, 
``Standby Liquid Control System,'' to remove the requirement for the 
standby liquid control system to be operable in OPERATIONAL CONDITION 
5, Refueling, when any control rod is withdrawn and the TS definition 
of CORE ALTERATION to exclude control rod movement in a control cell 
that contains no fuel assemblies.
    Date of issuance: January 27, 1995.
    Effective date: January 27, 1995.
    Amendment Nos. 87/49.
    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 
55881).
    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. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: September 28, 1993.
    Brief description of amendment: The amendment revised Technical 
Specification (TS) Section 4.11.D to change the surveillance 
requirements for the Emergency Service Water System pumps. The change 
added pump flow rate requirements and tests the pumps in accordance 
with the licensee's Inservice Testing Program. The respective TS Bases 
were also revised.
    Date of issuance: January 30, 1995.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 223.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62156). [[Page 8765]] 
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 1995.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendments: January 3, 1995 (TS 95-01).
    Brief description of amendments: The amendments add a permissive 
statement to Surveillance Requirement 4.9.7.1 that will allow the 
auxiliary building bridge crane interlocks and physical stops to be 
defeated during implementation of the spent fuel pool storage capacity 
increase modification.
    Date of issuance: January 24, 1995.
    Effective date: January 24, 1995.
    Amendment Nos.: 194 and 185.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 9, 1995 (60 FR 
2404) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 24, 1995.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By March 17, 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 [[Page 8766]] 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 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).

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

    Date of amendment request: July 26, 1994, as supplemented by 
letters dated December 27, 1994, and January 27, 1995.
    Brief description of amendment: The amendment changed the Technical 
Specification Section 3/4.12.A to allow for increased flow capacity of 
the control room emergency filter system. By increasing the maximum 
allowed makeup capacity of this system, additional margin is provided 
for the positive pressurization of the control room envelope.
    Date of issuance: January 27, 1995.
    Effective date: January 27, 1995.
    Amendment No.: 167.
    Facility Operating License No. DPR-46. Amendment revised the 
Technical Specifications.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated
    Local Public Document Room location: Auburn Public Library, 118 
15th Street, Auburn, Nebraska 68305.
    Attorney for licensee: Mr. G.D. Watson, Nebraska Public Power 
District, Post Office Box 499, Columbus, Nebraska 68602-0499.
    NRC Project Director: William D. Beckner.

    Dated at Rockville, Maryland, this 8th day of February 1995.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects--III/IV, Office of 
Nuclear Reactor Regulation.
[FR Doc. 95-3629 Filed 2-14-95; 8:45 am]
BILLING CODE 7590-01-P