[Federal Register Volume 60, Number 234 (Wednesday, December 6, 1995)]
[Notices]
[Pages 62485-62503]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X95-31206]



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UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 10, 1995, through November 24, 
1995. The last biweekly notice was published on November 27, 1995 (60 
FR 58395).

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 

[[Page 62486]]
    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 January 5, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, 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). 

[[Page 62487]]

    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.

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

    Date of amendments request: May 2, 1995
    Description of amendments request: The proposed change revises the 
large- break loss-of-coolant accident (LOCA) dose consequences. The 
large-break LOCA dose calculation is being changed to include an 
additional release path through allowable steam generator tube leakage 
to the atmospheric dump valves (ADVs) or turbine bypass valves (TBVs).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The probability or consequences of an accident previously 
evaluated are not significantly increased by this change to the 
large break LOCA dose consequences. This change has no effect on the 
LOCA safety analysis for emergency core cooling system performance, 
which demonstrates conformance to the acceptance criteria of 10 CFR 
50.46, as described in the PVNGS Updated Final safety Analysis 
Section 6.3.3. This change has no effect on structures, systems or 
components prior to a LOCA or any other accident. The new 
radiological consequences of the revised large break LOCA dose 
calculation are below 10 CFR 100 limits for the exclusion area 
boundary (EAB) and low population zone (LPZ), and the 10 CFR 50, 
Appendix A, GDC 19 limits for the control room, as shown in Table 1-
1, Column C. The NRC has previously approved changes to the PVNGS 
LOCA dose consequences with the acceptance criteria that the doses 
are still within the guidelines set forth in 10 CFR 100 and GDC 19. 
This acceptance criteria is described in the Safety Evaluation 
related to amendment Nos. 64, 50, and 37 to PVNGS Units 1, 2, and 3 
respectively, dated September 8, 1992.
    The LOCA dose calculation is being changed to include an 
additional release path through allowable steam generator tube 
leakage to the ADVs or TBVs. This change is necessary to reflect a 
revised calculation assumption that, following a large break LOCA, 
the secondary system pressure would fall below reactor coolant 
system pressure and containment pressure when operators cooldown the 
steam generators by using ADVs or the TBVs (in accordance with the 
safety analysis and EOPs [emergency operating procedures]). It is 
desirable to use the ADVs or TBVs to vent secondary system steam and 
thus reduce heat input to the reactor coolant system following a 
large break LOCA. No other LOCA analysis assumptions are being 
changed, and no changes are being made to structures, systems, 
components or procedures.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This change has no impact on any structures, systems, 
components, or procedures. The only impact is the revised 
radiological consequences of a large break LOCA to include an 
additional release path, as discussed in the response to Standard 1 
above. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    This change to the large break LOCA dose consequences does not 
involve a significant reduction in a margin of safety. The new 
radiological consequences of the revised large break LOCA dose 
calculation are below 10 CFR 100 limits for the EAB and LPZ, and the 
10 CFR 50, Appendix A, GDC 19 limits for the control room, as 
described in the response to Standard 1 above. The NRC has 
previously approved changes to the PVNGS LOCA dose consequences with 
the acceptance criteria that the doses are still within the 
guidelines set forth in 10 CFR 100 and GDC 19. This acceptance 
criteria is described in the Safety Evaluation related to amendment 
Nos. 64, 50, and 37 to PVNGS Units 1, 2, and 3 respectively, dated 
September 8, 1992. No equipment qualification is affected by the new 
assumption of a release path through the secondary system following 
a large break LOCA, and no post LOCA radiation zones will be 
changed. This change has no impact on any structures, systems, 
components, or procedures.
    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman

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

    Date of amendment request: November 22, 1995
    Description of amendment request: The current Technical 
Specifications (TS) Section 3.3.4.2 describes the limiting condition 
during which components in the Service Water (SW) system may be 
inoperable. The TS Section 3.3.4.2 states, in part, ``During power 
operation, the requirements of 3.3.4.1 may be modified to allow any one 
of the following components to be inoperable provided the remaining 
systems are in continuous operation.'' The proposed change will delete 
the qualifying statement,''... provided the remaining systems are in 
continuous operation,'' from TS Section 3.3.4.2. Currently, this 
statement requires the ``remaining systems to be in continuous 
operation'' while allowing one SW loop header, or one SW pump, or one 
SW booster pump to be inoperable for a period of 24 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change would remove the requirement for the 
remaining SW system components to be in continuous operation while 
one TS-required component is inoperable. Rather, the remaining 
components would remain operable, and no change would be made in 
normal system operation. The SW system provides an accident 
mitigation function and is not involved in accident initiation 
sequences. Therefore, the proposed change would not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The capacity of the SW system is such that its accident 
mitigation function can be performed by operation of a maximum of 
two SW pumps, one SW booster pumps, and one SW header. While a TS-
required component is inoperable, sufficient accident mitigation 
capability is provided by the remaining operable components, rather 
than requiring the remaining systems to be in continuous operation. 
Therefore, the proposed change would not cause a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change would remove the requirement for the 
remaining SW system 

[[Page 62488]]
components to be in continuous operation while one TS-required 
component is inoperable. Rather, the remaining components would 
remain operable. The proposed change would not change the normal 
operation of the system, nor would any physical modifications result 
from the change. The function and capability of the SW systems would 
remain unchanged. Therefore, the proposed change would not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change would remove the requirement for the 
remaining SW system components to be in continuous operation while 
one allowed TS-required component is inoperable. Rather, the 
remaining TS-required components would remain operable. Adequate 
assurance of operability is maintained by performance of regular 
surveillance testing. Maintaining operable status rather than 
placing equipment in continuous operation does not result in a 
change in the ability of the SW system to perform its intended 
function, since the system provides an automatic response to 
accident conditions, and the system possesses adequate capacity to 
perform its normal operating function with one allowed TS-required 
component inoperable. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: David B. Matthews

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

    Date of amendment request: October 24, 1995
    Description of amendment request: The proposed amendment will 
increase the trip setpoints and allowable values for the low power 
block (P-7).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    In accordance with 10CFR50.92, CYAPCO has reviewed the proposed 
change and has concluded that it does not involve a significant 
hazards consideration (SHC). The basis for this conclusion is that 
the three criteria of 10CFR50.92(c) are not compromised. The 
proposed change does not involve an SHC because the change would 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will relax the power level values for the P-
7 interlock by 2 percent. This change affects both the P-7 and P-7N 
interlocks. The P-7 interlock affects reactor trips on 1) low flow 
in more than one reactor coolant loop, 2) reactor coolant pump bus 
under voltage, 3) more than one reactor coolant pump breaker open, 
4) main steam line isolation valve closure, 5) turbine trip, and 6) 
variable low pressure. The P-7 interlock automatically blocks these 
reactor trips on decreasing power and automatically unblocks these 
reactor trips on increasing power. The P-7N interlock affects the 
reactor trip on wide range, neutron flux, high startup rate. P-7N 
automatically enables this reactor trip on decreasing power level 
and automatically blocks this reactor trip on increasing power 
level. The Applicable Modes requirement and Action Statements for 
the P-7 interlock and the reactor trips associated with both P-7 and 
P-7N in the Instrumentation Channel and Surveillance Requirements of 
Technical Specification 3/4.3.1 are being changed by 2 percent to be 
consistent with the change to P-7. The interlock setpoint cannot 
cause an accident. Also, the proposed 2 percent increase in the 
power level still results in a power level well below the power 
level at which the P-7 interlocked reactor trips are required for 
accident mitigation, as well as maintaining the high startup rate 
trip enabled at a higher power level. This proposed power level is 
consistent with the technical specification requirement prior to the 
conversion to standard format technical specifications and is also 
consistent with the Standard Westinghouse technical specification 
value. Therefore, the proposed change can neither increase the 
consequences of the design basis accident nor the probability of 
occurrence of the design basis accidents.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change only modifies the power level for the P-7 
and P-7N interlocks. The proposed setpoint is a power level at which 
stable plant conditions are easier to maintain while transferring 
the power supply for the reactor coolant pumps between offsite power 
and the main generator. The setpoint is also well below the power 
level for which the reactor protection afforded by the trips that 
are bypassed by P-7 is needed. This cannot create the possibility of 
a new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed change maintains the power level for the P-7 
interlock below the power level for which the reactor trips that are 
blocked by the P-7 interlock are required. It also raises the power 
level to a value at which it is easier to maintain stable plant 
conditions. This will reduce the likelihood of an automatic reactor 
trip during the transferring of power for the reactor coolant pumps 
between offsite power and the main generator. The proposed change 
will result in the high startup rate reactor trip being enabled at a 
higher power level. This is conservative since it expands the range 
of coverage for the trip. Therefore, the proposed change does not 
impact 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: November 1, 1995
    Description of amendment request: The proposed amendment will 
modifiy Surveillance Requirement 4.6.3.2, ``Containment Isolation 
Valves,'' (CIVs) to change the surveillance interval from at least once 
per 18 months to at least once per refueling interval.
    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:
    CYAPCO has reviewed the proposed change in accordance with 
10CFR50.92 and concluded that the change does not involve a 
significant hazards consideration (SHC). The basis for this 
conclusion is that the three criteria of 10CFR50.92(c) are not 
compromised. The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Surveillance Requirement 4.6.3.2 of the 
Haddam Neck Plant Technical Specifications extends the frequency for 
verifying that each CIV actuates to its required position in 
response to a safety injection actuation test signal. The proposal 
would extend the frequency from at least once per 18 months to at 
least once per 

[[Page 62489]]
refueling interval (24 months + 25% as allowed by Technical 
Specification 4.0.2).
    The proposed change to Surveillance Requirement 4.6.3.2 does not 
alter the intent or method by which the surveillance is conducted, 
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.
    Additional assurance of CIV operability is provided by 
Surveillance Requirement 4.6.3.3. Surveillance Requirement 4.6.3.3 
requires the confirmation of the mechanical operability of the CIVs 
by the inservice inspection program. The proposed change does not 
modify these requirements.
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
Surveillance Requirement 4.6.3.2. This evaluation included a review 
of surveillance results, preventive maintenance records, and 
corrective maintenance records. It has been concluded that the CIVs 
are highly reliable, and that there is no indication that the 
proposed extension could cause deterioration in valve condition or 
performance.
    As such, the proposed change to the frequency of Surveillance 
Requirement 4.6.3.2 will not degrade the ability of the CIVs to 
perform their safety function.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.3.2 of the Haddam Neck Plant Technical 
Specifications does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed change to Surveillance Requirement 4.6.3.2 of the 
Haddam Neck Plant Technical Specifications extends the frequency for 
verifying that each CIV actuates to its required position in 
response to a safety injection actuation test signal. The proposal 
would extend the frequency from at least once per 18 months to at 
least once per refueling interval (24 months + 25% as allowed by 
Technical Specification 4.0.2).
    The proposed change does not alter the intent or method by which 
the surveillance is conducted, 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 in the frequency of 
Surveillance Requirement 4.6.3.2 will not degrade the ability of the 
CIVs to perform their safety function.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.3.2 of the Haddam Neck Plant 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 a margin of safety.
    The proposed change to Surveillance Requirement 4.6.3.2 of the 
Haddam Neck Plant Technical Specifications extends the frequency for 
verifying that each CIV actuates to its required position in 
response to a safety injection actuation test signal. The proposal 
would extend the frequency from at least once per 18 months to at 
least once per refueling interval (24 months + 25% as allowed by 
Technical Specification Section 4.0.2).
    The proposed change does not alter the intent or method by which 
the surveillance is conducted, 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 in the frequency of 
Surveillance Requirement 4.6.3.2 will not degrade the ability of the 
CIVs to perform their safety function.
    Additional assurance of the operability of the CIVs is provided 
by Surveillance Requirement 4.6.3.3.
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
Surveillance Requirement 4.6.3.2. This evaluation included a review 
of surveillance results, preventive maintenance records, and 
corrective maintenance records. It has been concluded that the CIVs 
are highly reliable, and that there is no indication that the 
proposed extension could cause deterioration in valve condition or 
performance.
    Based on the above, the proposed change to Surveillance 
Requirement 4.6.3.2 of the Haddam Neck Plant Technical 
Specifications 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: November 8, 1995, as supplemented 
November 17, 1995
    Description of amendment request: The proposed amendment would 
remove the prescriptive Type A containment leakage test rate frequency 
of 40 plus or minus 10 months and add a reference to perform 
containment leakage rate tests in accordance with the criteria 
specified in Appendix J of 10 CFR Part 50 as amended by approved 
exemptions. In addition, the proposed amendment would revise the test 
pressure for Type B and C testing to correct a typographical error.
    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:
    Leakage test rate frequency
    1) The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    This change is administrative in nature and does not impact 
plant systems, structures or components. The proposed change will 
allow the facility's technical specifications to be revised to allow 
containment sphere leakage testing in accordance with Appendix J to 
10 CFR Part 50 as modified by approved exemptions.
    2) The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    This change is administrative in nature and does not impact 
plant syst ems, structures or components. The proposed change will 
allow the facility's technical specifications to be revised to allow 
containment sphere leakage testing in accordance with Appendix J to 
10 CFR Part 50 as modified by approved exemptions.
    3)The proposed change does not involve a significant reduction 
in a margin of safety.
    This change is administrative in nature and does not impact 
plant systems, structures or components. The underlying purpose of 
Appendix J is still achieved. Appendix J states that the leakage 
test requirements provide for periodic verification testing of the 
leak tightness integrity of the primary reactor containment. The 
appendix further states that the purpose of the tests is to assure 
that leakage through the primary containment shall not exceed the 
allowable leakage rate values as specified in the technical 
specifications or associated bases. As stated previously, for Big 
Rock Point and a large percentage of other plants, the Appendix J 
Type B and C testing programs provide the most significant and 
meaningful assessment of containment leak tightness. The testing 
history and structural capability of the containment establish that 
there is significant assurance that the extended interval between 
Type A tests will not adversely impact the integrity of the 
containment.
    Test pressure revision
    As stated in the technical specification change request, this 
revision is being performed to be consistent with accident pressure, 
Pa, used for Big Rock Point. 20 psig is a typographical error. 
23 psig has always been used for these tests.
    The proposed change does not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This change is administrative in nature and does not impact 
plant systems, structures or components. The proposed change will 
allow the facility's technical specifications to be revised to 
reflect current containment sphere leakage testing in accordance 
with Appendix J to 10 CFR Part 50. 

[[Page 62490]]

    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    This change is administrative in nature and does not impact 
plant systems, structures or components. The proposed change will 
allow the facility's technical specifications to be revised to 
reflect current containment sphere leakage testing in accordance 
with Appendix J to 10 CFR Part 50.
    3) involve a significant reduction in a margin of safety.
    This change is administrative in nature and does not impact 
plant systems, structures or components.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: Brian E. Holian, Acting

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 amendment request: October 16, 1995
    Description of amendment request: Appendix J of 10 CFR Part 50, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' has recently been revised to include Option B. This option 
allows the implementation of a performance based Type B and C testing 
program. The proposed change will add a footnote to Technical 
Specification (TS) 4.6.1.2.d stating that the Type B and C tests 
scheduled for Unit 1 refueling outage Cycle 6 (1R6) will be conducted 
in accordance with Option B and using the guidance of Regulatory Guide 
1.163, Revision 0. This option is being incorporated into the 
licensee's request to implement the improved TS. However, the improved 
TS are not scheduled to become effective until after the Unit 1 
refueling outage 1R6.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The change does not involve a significant increase in 
probability or consequences of an accident previously evaluated. The 
proposed change does not involve a change to structures, systems, or 
components which would affect the probability or consequences of an 
accident previously evaluated in the Vogtle Electric Generating 
Plant (VEGP) Final Safety Analysis Report (FSAR). The proposed 
change only provides a mechanism within the Technical Specifications 
for implementing a performance-based method of determining the 
frequency for leak rate testing which has been approved by the NRC 
via a revision to 10 CFR 50, Appendix J.
    2. The proposed change will not create the possibility of a new 
or different kind of accident from any accident previously analyzed. 
The amendment will not change the design, configuration, or method 
of plant operation. It only allows for the implementation of Option 
B of 10 CFR 50, Appendix J for Unit 1 refueling outage 1R6 without 
violating the plant Technical Specifications.
    3. Operation of VEGP, Unit 1, in accordance with the proposed 
change will not involve a significant reduction in the margin of 
safety. The proposed change does not affect a safety limit, an LCO 
[limiting condition for operation], or the way plant equipment is 
operated. The NRC is aware that changes similar to this proposed 
change are required in order to implement Option B of 10 CFR 50, 
Appendix J. In fact, the staff indicates in Paragraph V.B. of 
Appendix J that Option B or parts thereof may be adopted by a 
licensee 30 days after the rule becomes effective by submitting 
notification of its implementing plan and a request for revision to 
Technical Specifications. Since the NRC has approved the provision 
for performance-based testing and must approve this Technical 
Specification[] change before the performance-based Option B can be 
implemented, the margin of safety will not be significantly reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: May 30, 1995 (noticed in the Federal 
Register July 5, 1995, (60 FR 35080) as supplemented by letter dated 
November 20, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications as follows:
    1. The Surveillance Frequency for the drywell bypass test is 
changed from 18 months to 10 years with an increased testing frequency 
required if performance degrades.
    2. The following changes are requested for the drywell air lock 
testing: (a) the leakage rate surveillance is moved from the air lock 
Limiting Condition for Operation (LCO) to the drywell LCO, (b) the 
requirement for the air lock to meet a specific overall leakage limit 
is deleted, (c) the Note that an inoperable air lock door does not 
invalidate the previous air lock leakage test is deleted, (d) the Note 
which required that the air lock leakage test at 3 psid be preceded by 
pressurizing the air lock to 19.2 psid is moved to the bases, and (e) 
the Surveillance Frequency for the air lock leakage test and interlock 
test is changed from 18 months to 24 months.
    3. The Actions Notes in the drywell air lock LCO and the drywell 
isolation valve LCO that identifies that the Actions required by the 
drywell LCO must be taken when the drywell bypass leakage limit is not 
met is deleted.
    4. The requirement for the drywell air lock seal leakage rate to 
meet a specific leakage limit is deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration for River Bend Station (RBS) and Grand Gulf Nuclear 
Station (GGNS), which is presented below:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The requested changes are either administrative changes which 
clarify the format of the requirement or change the requirement to 
match the design bases of the plant, a change which relocates the 
requirement to the Technical Specification Bases, or a change in 
surveillance interval. Each of these types of change are discussed 
below:
    1. The administrative changes clarify the format of the 
requirement or change therequirement to match the design bases of 
the plant. Clarifying administrative format of 

[[Page 62491]]
the Technical Specifications does not result in any changes to the 
Technical Specification requirements and, as a result, does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated. Also, changing the requirements of 
the Technical Specifications to more closely match the design bases 
of the plant will continue to assure that the plant will respond as 
assumed in the accident analyses and, as a result, does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes relocate information to the Technical 
Specification Bases. In the Technical Specifications Bases the 
relocated information will be maintained in accordance with 10 CFR 
50.59 and subject to the change control provisions in Chapter 5 of 
Technical Specifications. Since any changes to the Technical 
Specifications Bases will be evaluated per the requirements of 10 
CFR 50.59, no increase (significant or insignificant) in the 
probability or consequences of an accident previously evaluated will 
be allowed. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    3. The proposed changes in frequency for the drywell bypass 
leakage and drywell air lock surveillances will continue to ensure 
that no paths exist through passive drywell boundary components that 
would permit gross leakage from the drywell to the primary 
containment air space and result in bypassing the primary 
containment pressure-suppression feature beyond the design basis 
limit. The Mark III primary containment system satisfies General 
Design Criterion 16 of Appendix A to 10 CFR Part 50. Maximum drywell 
bypass leakage was determined previously by reviewing the full range 
of postulated primary system break sizes. The limiting case was a 
primary system small break loss of coolant accident (LOCA) and 
yielded a design allowable drywell bypass leakage rate limit of 
approximately 35,000 scfm for GGNS and 46,000 scfm (the Technical 
Specification limit is based on a lower limit of 40,110 scfm) for 
RBS. The Technical Specifications acceptable limit for the bypass 
leakage following a surveillance is less than 10% of this design 
basis value. The most recent bypass leakage value was approximately 
2.5% for GGNS and .91% for RBS of the design allowable leakage rate 
limit for the limiting event. EOI is committed to maintaining 
programmatic and oversight controls that ensure that drywell bypass 
leakage remains a small fraction of the design allowable leakage 
limit.
    The drywell is typically exposed to essentially 0 psig during 
normal plant operation and 3 psig during drywell bypass leak rate 
testing. These pressures are considerably lower than the structural 
integrity test pressure and are less likely to initiate a crack or 
cause an existing crack to grow. Visual inspections of the 
accessible drywell surfaces that have been performed since the 
structural integrity tests have not revealed the presence of 
additional cracking or other abnormalities. Therefore, additional 
cracking of the drywell structure is not expected due to testing or 
operation and, similar to the justification for the ten year 10 CFR 
50 Appendix J Type A test interval, it is not considered credible 
for the passive drywell structure to begin to leak sufficiently to 
impact the design drywell bypass leakage limit.
    The primary containment's ability to perform its safety function 
is fairly insensitive to the amount of drywell leakage, thereby 
providing a margin to loss of the drywell safety function that is 
not normally available for safety systems. This insensitivity is 
demonstrated by the extremely high limiting event design basis 
allowable leakage for the drywell (e.g., 35,000 scfm for GGNS and 
46,000 scfm for RBS). The limiting leakage is almost an order of 
magnitude higher for other events. Additionally, an even higher 
allowable leakage can be realistically accommodated by the primary 
containment due to the margins in the containment design. Because of 
the margins available, it will take valves in multiple penetration 
flow paths leaking excessively to cause the primary containment to 
fail as a result of overpressurization, the probability that drywell 
isolation valve leakage will result in primary containment failure 
due to excessive drywell leakage is not considered significant and 
this drywell/primary containment failure mode is not considered 
credible.
    The proposed Technical Specification changes have no significant 
impact on the GGNS Individual Plant Examination (IPE) or the RBS IPE 
conducted per NRC Generic Letter 88-20. The IPEs considered 
overpressurization failure of primary containment as part of the 
primary containment performance assessment. Due to the magnitude of 
acceptable drywell leakage and the extremely low probabilities of 
achieving such leakage, primary containment failure due to 
preexisting excessive drywell leakage was considered a non 
significant contributor to primary containment failure. Primary 
containment overpressurization failure can occur with or without 
preexisting excessive drywell leakage in a severe accident. This is 
due to physical phenomena associated with potentially extreme 
environmental conditions inside primary containment following a 
severe accident. However, the calculated frequency of such extreme 
conditions is very small. The proposed changes do not impact the IPE 
evaluated phenomena causing primary containment overpressurization 
failure nor significantly increase the probability that the drywell 
has preexisting excessive leakage and therefore would not contribute 
to these accident scenarios.
    For the reasons discussed above, the proposed changes do not 
have any significant risk impact to accidents previously evaluated 
and do not significantly increase the consequences of an accident 
previously evaluated. Additionally, drywell leakage is not the 
initiator of any accident evaluated; therefore, changes in the 
frequency of the surveillance for drywell leakage does not increase 
the probability of any accident evaluated.
    Therefore, the proposed changes do not significantly increase 
the probability or consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested changes are either administrative changes which 
clarify the format of the requirement or change the requirement to 
match the design bases of the plant, a change which relocates the 
requirement to the Technical Specification Bases, or a change in 
surveillance interval. Each of these types of change are discussed 
below:
    1. The administrative changes in the Technical Specification 
requirements do not involve a physical alteration of the plant (no 
new or different type of equipment will be installed) nor does it 
change the methods governing normal plant operation. Thus, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    2. The proposed relocation of requirements does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) nor does it change the methods 
governing normal plant operation. The proposed change will not 
impose or eliminate any requirements. Adequate control of the 
information will be maintained in the Technical Specification Bases. 
Thus, the change proposed does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed change modifies the surveillance frequency for 
drywell bypass leakage and drywell air lock surveillances. The 
changes only impact the test frequency and do not result in any 
change in the response of the equipment to an accident. The changes 
do not alter equipment design or capabilities. The changes do not 
present any new or additional failure mechanisms. The drywell is 
passive in nature and the surveillance will continue to verify that 
its integrity has not deteriorated. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    The requested changes are either administrative changes which 
clarify the format of the requirement or change the requirement to 
match the design bases of the plant, a change which relocates the 
requirement to the Technical Specification Bases, or a change in 
surveillance interval. Each of these types of changes are discussed 
below:
    1. The administrative changes in the Technical Specification 
requirements do not involve a physical alteration of the plant (no 
new or different type of equipment will be installed) nor does it 
change the methods governing normal plant operation. Thus, this 
change does not cause a significant reduction in the margin of 
safety. 

[[Page 62492]]

    2. The relocation of requirements will not reduce a margin of 
safety because it has no impact on any safety analysis assumptions. 
In addition, the requirements to be transferred from the Technical 
Specifications to the Technical Specifications Bases are the same as 
the existing Technical Specifications. Since any future changes to 
these requirements in the Technical Specifications Bases will be 
evaluated per the requirements of 10 CFR 50.59, no reduction 
(significant or insignificant) in a margin of safety will be 
allowed.
    3. The proposed change modifies the surveillance frequency for 
drywell bypass leakage and associated air lock surveillances. 
Reliability of drywell integrity is evidenced by the measured 
leakage rate during past drywell bypass leakage surveillances. 
Appropriate design basis assumptions will be upheld, even when 
combined with the complementary bypass leakage surveillances as 
proposed. Drywell integrity will continue to be tested by means of 
the proposed periodic drywell bypass leakage test, performance of 
the drywell air lock door latching and interlock mechanism 
surveillance, and performance of additional surveillances including 
excercising of drywell isolation valves. The combination of these 
surveillances will provide adequate assurance that drywell bypass 
leakage will not exceed the design basis limit. Margins of safety 
would not be reduced unless leakage rates exceeded the design 
allowable drywell bypass leakage limit. Therefore, the proposed 
change does not cause 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: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: October 26, 1995
    Description of amendment request: The proposed amendment would 
revise the technical specifications for sixteen editorial changes and 
would delete the requirement for a program to prevent and detect 
Asiatic Clams (Corbicula) in the service water system (SWS). The 
editorial changes covers such things as removing systems or components 
that do not exist in the River Bend Station, correcting typographical 
errors, correcting to be consistent with the writers guide for Improved 
Technical Specifications, adding descriptions for systems to make them 
clear, and wording changes to be consistent with approved facility 
operations. The Corbicula program is no longer needed because the 
facility has been modified and SWS no longer takes water from the 
Mississippi River; source of the larvae and infestation.
    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:
    EDITORIAL CHANGES
    The purposed changes involves reformatting, renumbering and 
rewording of the existing Technical Specifications. The 
reformatting, renumbering and rewording process involves no 
technical changes to existing Technical Specifications. As such, 
these changes are administrative in nature and do not impact 
initiators of analyzed events or assumed mitigation of accident or 
transient events. Therefore, these changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes do not involve a physical alteration of the 
plant (no new or different type of equipment will be installed) or 
changes in methods governing normal plant operation. The proposed 
changes will not impose or eliminate any new or different 
requirements. Thus, these changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes will not reduce a margin of safety because 
they have no impact on any safety analysis assumptions. These 
changes are administrative in nature. As such, no question of safety 
is involved, and the changes do not involve a significant reduction 
in a margin of safety.
    CORBICULA PROGRAM
    The proposed change deletes the program associated with the 
prevention and detection of Asiatic Clams (Corbicula) based upon 
improvements to the non-safety related Normal Service Water System 
(SWS). The source of makeup water to the SWS is no longer the 
Mississippi River, which is the source of Asiatic Clams. 
Demineralized water or well water is used eliminating the source of 
asiatic clams. To prevent biofouling SWS is treated with chlorine/
bromine. This program is not considered as an initiator for any 
previously evaluated accident. Therefore, the proposed change will 
not increase the probability or consequences of any accident 
previously evaluated.
    The proposed change introduces no new mode of plant operation 
and it does not involve a physical modification to the plant. The 
possibility of the SES becoming contaminated by any other means is 
highly unlikely since it is a ``closed-loop'' system. Therefore it 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Prevention of Asiatic Clam infestation in the SWS and associated 
safety-related equipment is ensured by the ``closed-loop'' design of 
the SWS. Post Refuel Outage (RF-4) inspections of the safety-related 
heat exchangers that interface with the ``closed-loop'' SWS have 
shown no evidence of clam infestations. Therefore, the change does 
not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 20, 1995
    Description of amendment request: The proposed amendment would 
revise the technical specifications to eliminate the response time 
testing requirements for selected Reactor Protection System 
Instrumentation.
    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 purpose of the proposed Technical Specification (TS) change 
is to eliminate response time testing requirements for selected 
components in the Reactor Protection System (RPS). The Boiling Water 
reactors Owners' Group (BWROG) has completed an evaluating which 
demonstrates that response time testing is redundant to the other 
TS-required testing. These other tests, in conjunction with actions 
taken in response to NRC Bulletin 90-01, ``Loss of Fill-Oil in 
Transmitters Manufactured by Rosemount,'' and Supplement 1, are 
sufficient to identify failure modes or degradation in instrument 
response times and ensure operation of the associated systems within 
acceptable limits. There are no known failure modes that can be 
detected by response time testing that cannot also be detected by 
the other TS-required testing. This evaluation was 

[[Page 62493]]
documented in NEDO-32291, ``System Analyses for Elimination of Selected 
Response Time Testing Requirements,'' January 1994. Entergy 
Operations, Inc. (EOI) has confirmed the applicability of this 
evaluation to River Bend Station (RBS). In addition EOI will 
complete the actions identified in the NRC staff's safety evaluation 
of NEDO-32291.
    Because of the continued application of other existing TS-
required tests such as channel calibration, channel checks, channel 
functional tests, and logic system functional tests, the response 
time of these systems will be maintained within the acceptance 
limits assumed in plant safety analyses and required for successful 
mitigation of an initiating event. The proposed changes do not 
affect the capability of the associated systems to perform their 
intended function within their required response time, nor do the 
proposed changes themselves affect the operation of any equipment. 
As a result, EOI has concluded that the proposed changes do not 
involve a significant increase in the probability or the 
consequences of an accident previously evaluated.
    The proposed changes only apply to the testing requirements for 
the components identified above and do not result in any physical 
change to these or other components or their operation. As a result, 
no new failure modes are introduced. Therefore, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accidents previously evaluated.
    The current TS-required response times are based on the maximum 
allowable values as assumed in the plant safety analyses. These 
analyses conservatively establish the margin of safety. As described 
above, the proposed changes do not affect the capability of the 
associated systems to perform their intended function within the 
allowed response time used as the basis for the plant safety 
analyses. The potential failure modes for the components within the 
scope of this request were evaluated for impact on instrument 
response time. This evaluation confirmed that, with the exception of 
loss of fill-oil of Rosemount transmitters, the remaining TS-
required testing is sufficient to identify failure modes or 
degradation in instrument response times and ensure operation of the 
instrument within the scope of this request is within acceptable 
limits. The actions taken in response to NRC Bulletin 90-09 and 
Supplement 1 are adequate to identify loss of fill-oil failures of 
Rosemount transmitters. As a result, it has been concluded that 
plant and systems response to an initiating event will remain in 
compliance with the assumptions of the safety analysis.
    Further, although not explicitly evaluated, the proposed changes 
will provide an improvement to plant safety and operation by 
reducing the time safety systems are unavailable, reducing the 
potential for safety system actuations, reducing plant shutdown 
risk, limiting radiation exposure to plant personnel, and 
eliminating the diversion of key personnel resources to conduct 
unnecessary testing. Therefore, EOI has concluded that this request 
will result in an overall increase 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: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: September 22, 1995
    Description of amendment request: The proposed amendment would 
modify a requirement of the Seabrook Station, Unit No. 1 Technical 
Specifications. Specifically, the proposed amendment would change the 
ACTION referenced in Table 3.3-3, Engineered Safety Features Actuation 
System Instrumentation, for Functional Unit 8.b, Automatic Switchover 
to Containment Sump/RWST Level Low-Low. The ACTION requirement would be 
changed to ACTION 15 from ACTION 18. ACTION 15 requires an inoperable 
channel to be placed in bypass (with no time limit specified) while 
ACTION 18 requires an inoperable channel to be placed in the tripped 
condition within 6 hours.
    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 would result in an 
inoperable Functional Unit 8.b. protective channel being placed in 
the bypassed condition vice tripped condition. Functional Unit 8.b. 
is not involved in any accident initiation sequence; therefore, the 
probability of a previously-analyzed accident is not increased. 
Placing an inoperable Functional Unit 8.b. in bypass vice trip 
reduces the probability of premature opening of the containment 
building sump isolation valves thereby reducing the potential for 
increasing the consequences of a previously-analyzed accident. Thus, 
the consequences of a previously-analyzed accident is not increased.
    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 reduce the minimum 
required number of channels of instrumentation to be operable. The 
change does not alter the function of or affect the failure modes of 
Functional Unit 8.b. instrumentation channels. The proposed change 
does not otherwise affect the manner by which the facility is 
operated, and it does not involve any changes to equipment or 
features which affect the operational characteristics of the 
facility.
    C. The change does not involve a significant reduction in a 
margin of safety (10 CFR 50.92(c)(3)) because the change does not 
reduce the minimum required number of channels of instrumentation to 
be operable, and it does not involve any changes to equipment or 
features which affect the operational characteristics of the 
facility. Therefore, the protection previously provided remains 
unchanged.
    Based on this review, it appears that the three standards of 10 CFR 
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine 
that the amendment request involves no significant hazards 
consideration.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.
    Attorney for licensee: Lillian M. Cuoco, Esquire, 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. 50-336, 
Millstone Nuclear Power Station, Unit No. 2, New London, 
Connecticut

    Date of amendment request: May 26, 1995, supplemented and revised 
October 20, 1995.
    Description of amendment request: The proposed changes would modify 
TS 3.8.1.1., ``Electrical Power Systems, A.C. Sources, Operating,'' TS 
3.8.1.2, ``Electrical Power Systems, Shutdown,'' TS 3.8.2.2, 
``Electrical Power Systems, A.C. Distribution - Shutdown,'' and TS 
3.8.2.4, ``Electrical Power Systems, D.C. Distribution - Shutdown,'' to 
provide operational flexibility as well as consistency between action 
statements and to eliminate certain surveillance requirements that are 
not applicable in Modes 5 or 6.
    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: 

[[Page 62494]]

    In accordance with 10 CFR 50.92, NNECO has reviewed the proposed 
changes and has concluded that they do not involve an SHC. The basis 
for this conclusion is that the three criteria of 10 CFR 50.92(c) 
are not compromised. The proposed changes do not involve an SHC 
because the change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to Surveillance Requirement 4.8.1.1.1 is 
being made because presently, the surveillance requirement for 
demonstrating offsite sources are operable states that ``two'' 
independent circuits are required. The surveillance requirement is 
referenced for both operating and shutdown modes. While it is 
accurate for operating modes, it is inconsistent with the limiting 
condition for operation for shutdown. The proposed change is safe 
because it renders the surveillance requirement consistent with the 
applicable limiting condition for operation (i.e., operating or 
shutdown) and eliminates a potential source of confusion.
    The change to Surveillance Requirement 4.8.1.2 and Technical 
Specification 3.8.2.2 merely clarifies the diesel generator 
surveillance and operability requirements for Modes 5 and 6 and 
renders action statements for related technical specification 
sections consistent with and appropriate for operational Modes 5 and 
6.
    Regarding diesel generator surveillance requirements, automatic 
A.C. power for LNP events in Modes 5 and 6 is not required. This is 
validated by the fact that the undervoltage sensors are only 
required to be operable in Modes 1, 2 and 3 to meet technical 
specifications. Because the undervoltage sensors provide the logic 
that results in actuation of the sequencer, it follows that the 
sequencer need not be operable in Modes 5 and 6. Accordingly, the 
sequencer is not required to support operability of the available 
diesel generator in Modes 5 and 6. Further, because SIAS is blocked 
in Modes 5 and 6, automatic start of the diesel generator upon 
receipt of a SIAS is similarly not required to support operability 
of the diesel generator in Modes 5 and 6.
    Additionally, operation of the diesel generator in parallel with 
the system during Modes 5 and 6 is not required to perform its 
intended safety function. In fact, such operation may compromise 
both sources as the result of a single event.
    Since automatic A.C. power is not credited in the mitigation of 
Mode 5 and 6 events and accidents, such as fuel handling accidents, 
there is no increase in the probability or consequences of 
previously evaluated accidents.
    The action statement in Technical Specification 3.8.2.2 has been 
revised to cite actions that are more appropriate for Modes 5 and 6 
for Millstone Unit No. 2. This is due to the ability to maintain the 
plant in a safe condition without needing to automatically load the 
diesel generator through the sequencers in Modes 5 and 6. In 
addition, the proposed change is consistent with the CE Owner's 
Group Standard Technical Specification and with other Millstone Unit 
No. 2 action statements. Consequently, there is no increase in the 
probability or consequences of previously evaluated accidents.
    The change to TS 3.8.2.4 merely renders the action statement 
consistent with, and appropriate for, operational Modes 5 and 6.
    Since D.C. power is not credited in the mitigation of Mode 5 and 
6 events and accidents, such as fuel handling accidents, there is no 
increase in the probability or consequences of previously evaluated 
accidents.
    The action statement in TS 3.8.2.4 has been revised to cite 
actions that are more appropriate for Modes 5 and 6 for Millstone 
Unit No. 2. This is due to the ability to maintain the plant in a 
safe condition without D.C. power distribution available in Modes 5 
and 6. In addition, the proposed change is consistent with the CE 
Owner's Group Standard Technical
    Specifications (NUREG-1432) and with other Millstone Unit No. 2 
action statements. Consequently, there is no increase in the 
probability or consequences of previously evaluated accidents.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    The proposed changes do not alter or affect the design, 
function, failure mode, or operation of the plant. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to the technical specifications provides 
greater consistency between the action statements and clarifies 
which surveillance requirements are required in Modes 5 and 6. Since 
the diesel generators are not required to be loaded automatically in 
Modes 5 and 6, and since it is part of our shutdown risk management 
program to assure that adequate cooling is able to be provided, and 
since the diesel will still be verified to start and achieve rated 
speed, the proposed changes to the technical specifications do not 
reduce the margin of safety.
    The proposed change to the TS provides greater consistency among 
action statements during Modes 5 and 6. Since the D.C. distribution 
system is not credited in the mitigation of Mode 5 and 6 events and 
accidents, and since it is part of our shutdown risk management 
program to assure that adequate fuel cooling is able to be provided, 
the proposed change to the TS does not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: 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: June 27, 1995, as supplemented July 21, 
1995
    Description of amendment request: The amendment revises the 
Technical Specifications (TS) to relocate TS requirements for the 
containment purge exhaust and supply valves, and to remove a duplicate 
testing requirement for the safety injection input from engineered 
safety features from the TS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    ... The proposed changes do not involve an SHC [significant 
hazards consideration] because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The first proposed change relocates the operability and 
surveillance requirements for the containment high range radiation 
monitors from Technical Specification Section 3.3.3 to Technical 
Specification Section 3.3.2. The proposed changes are administrative 
in nature. The proposed changes do not alter the way any structure, 
system, or component functions and do not modify the manner in which 
the plant is operated and do not involve any physical changes to the 
plant.
    The second proposed modification will delete the testing 
requirement for functional unit 16, ``Safety Injection Input from 
ESF,'' of Table 4.3-1 because the logic circuitry that processes
    the safety injection signals and produces a reactor trip is 
tested under functional unit 19 ``Automatic Trip and Interlock 
Logic,'' and the testing is performed on a more frequent basis 
(i.e., on a monthly staggered bases versus on an 18-month 
frequency). In addition, the same logic testing is accomplished with 
an 18-month TADOT of functional unit 1.a of Table 4.3-2 and with a 
monthly staggered actuation logic testing of functional unit 16 of 
Table 4.3-2. This testing ensures that operability of the logic 
under functional unit 16 of Table 4.3-1 is verified. The other tests 
will continue to verify the operability of the reactor trip system 
and that a reactor trip will be initiated when required.
    Therefore, 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 accident 
from any previously analyzed. 

[[Page 62495]]

    The proposed changes do not affect the operation or response of 
any plant equipment or introduce any new failure mechanisms. The 
proposed elimination of the testing requirement line item does not 
affect the test results since the logic circuitry that processes the 
safety injection signal and produces a reactor trip will be tested 
and is tested under functional unit 19 of Table 4.3-1. 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 proposed changes do not have any adverse impact on the 
protective boundaries nor do they affect the consequences of any 
accident analyzed. The operability and surveillance requirements, 
although relocated to other technical specifications, will still 
ensure that the system (the radiation monitors) is tested and within 
limits. The proposed elimination of the testing equipment will not 
change the performance or operating conditions of the safety 
systems. The operable reactor trip system instrumentation ensures 
that the assumptions in the Bases of the Technical Specifications 
are not affected and ensures that the margin of safety is not 
reduced. Therefore, the proposed changes do not reduce the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment requests: November 14, 1994
    Description of amendment requests: The proposed amendment would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2, for the slave relay test 
frequency from quarterly (Q) to refueling (R). The request would also 
remove table notation 4 from Table 4.3-2. The associated Bases would 
also be appropriately revised.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The results of WCAPs 14117 and 13878 demonstrate that slave 
relays are highly reliable. The WCAPs also provide guidance to 
assure that slave relays remain highly reliable. The aging 
assessment concludes that the age/temperature-related degradation of 
all ND relays, and NE relays produced after May 1990, is 
sufficiently slow such that a refueling frequency surveillance 
interval will not significantly increase the probability of slave 
relay failures. Finally, the evaluation of the interposing slave 
relays in the emergency diesel generator start circuitry, control 
room ventilation and auxiliary building ventilation realignments, 
steam generator blowdown isolation and radwaste isolation systems 
has concluded that based on the tests of the interposing relays 
performed during other equipment testing, reasonable assurance is 
provided that failures will be identified if the associated slave 
relays are tested on a refueling frequency.
    The removal of table notation 4 from TS Table 4.3-2 is an 
administrative change that eliminates unnecessary redundancy from 
the TS and does not affect plant operation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not alter the performance of the ESFAS 
mitigation systems assumed in the plant safety analysis. Changing 
the interval for periodically verifying ESFAS slave relays (assuring 
equipment operability) will not create any new accident initiators 
or scenarios.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated for DCPP.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect the total ESFAS response 
assumed in the safety analysis since the reliability of the slave 
relays will not be significantly affected by the increased 
surveillance frequency.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

South Carolina Electric & Gas Company, South Carolina Public 
Service Authority, Docket No. 50-395, Virgil C. Summer Nuclear 
Station, Unit No. 1, Fairfield County, South Carolina

    Date of amendment request: August 18, 1995, as supplemented on 
November 1, 1995
    Description of amendment request: The proposed amendment would 
revise the Operating License and Technical Specifications to allow for 
a power uprate to 2900 MWt. The current maximum power level is 2775 
MWt.
    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 probability or consequences of an accident previously 
evaluated is not significantly increased.
    Implementation of uprate power operation does not contribute to 
any accident evaluated in the FSAR [Final Safety Analysis Report]. 
The NSSS [Nuclear Steam Supply System] Components (RV [reactor 
vessel], RCPs [reactor coolant pumps], CRDMs [control rod drive 
mechanisms], SGs [steam generators], and piping) are compatible with 
the revised operating conditions. These components have been 
reanalyzed and the results show that ASME [American Society of 
Mechanical Engineers] Code requirements remain satisfied and are 
within the current Licensing Basis.
    Interfacing Systems which are important to safety are not 
adversely impacted and will continue to perform their design 
function. Overall secondary plant performance is not significantly 
altered by the proposed changes.
    The revision to the Pressure Temperature Limits will not 
adversely impact the RCS [reactor coolant system] Pressure Boundary. 
The length of time these curves will be applicable, due to increased 
neutron fluence, is being reduced. Before the 13 Effective Full 
Power Years have elapsed, new curves will be generated to reflect 
the analysis of the specimen capsule and will be derived utilizing 
NRC approved methodology.
    Therefore, since the Reactor Coolant pressure boundary integrity 
and system functions are not adversely impacted, the probability of 
occurrence of an accident evaluated in the VCSNS [Virgil C. Summer 
Nuclear Station] FSAR will be no greater than the original design 
basis of the plant.
    An extensive analysis has been performed to evaluate the 
consequences of the following accident types currently evaluated in 
the VCSNS FSAR: 

[[Page 62496]]

    - Non-LOCA [loss-of-coolant accident] Events
    - Large Break and Small Break LOCA
    - Steam Generator Tube Rupture
    With the [delta]75 SGs and revised operating conditions, the 
calculated results (i.e., DNBR [departure from nucleate boiling 
ratio], Primary and Secondary System Pressure, Peak Clad 
Temperature, Metal Water Reaction, Challenge to Long Term Cooling, 
Environmental Conditions Inside and Outside containment, etc.) for 
the accidents are similar to those currently reported in the VCSNS 
FSAR and remain within applicable Regulatory Acceptance Criteria. 
Select results (i.e., Containment Pressure during a Steam Line 
Break, Minimum DNBR for Rod Withdrawal from Subcritical, etc.) are 
slightly more limiting than those currently reported in the FSAR due 
to the use of the assumed operating conditions with the [delta]75 
SGs and in some cases, use of an uprated core power of 2900 MWt. 
However, in all cases, the calculated results do not challenge the 
integrity of the primary/secondary/ containment pressure boundary 
and remain within the regulatory acceptance criteria applied to 
VCSNS's current licensing basis.
    Given that calculated radiological consequences are not 
significantly higher than current FSAR results and remain well 
within 10 CFR 100 limits, it is concluded that the consequences of 
an accident previously evaluated in the FSAR are not significantly 
increased.
    2. The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Uprate power operation will not introduce any new accident 
initiator mechanisms. Structural integrity of the RCS is maintained 
during all plant conditions through compliance with the ASME code 
and 10 CFR 50 Appendix G requirements. Design requirements of 
auxiliary systems are met with the RSGs [replacement steam 
generators] and uprate power operation. No new failure modes or 
limiting single failures have been identified. Since the safety and 
design requirements continue to be met and the integrity of the 
reactor coolant system pressure boundary is not challenged, no new 
accident scenarios have been created. Therefore, the types of 
accidents defined in the FSAR continue to represent the credible 
spectrum of events to be analyzed which determine safe plant 
operation.
    3. The proposed license amendment does not involve a significant 
reduction in a margin of safety.
    Although uprate power operation will require changes to the 
VCSNS Technical Specifications, the proposed changes are supported 
by extensive LOCA, NON-LOCA and SGTR [steam generator tube rupture] 
analyses. These analyses show acceptable consequences with margin to 
the applicable regulatory limits. All equipment required to function 
during accident conditions has been shown to remain qualified and 
thus will perform their design function, and all components remain 
in compliance with the codes and standards in effect when VCSNS was 
originally licensed (with the exception of the replacement steam 
generators which use the 1986 ASME Code Section III Edition).
    Low Temperature Overpressure transients which could challenge 
RCS structural integrity are not impacted by the revision to the 
Pressure Temperature Limitations Curves. The curves are not directly 
impacted, the changes do not reduce any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180
    Attorney for licensee: Randolph R. Mahan, South Carolina Electric & 
Gas Company, Post Office Box 764, Columbia, South Carolina 29218
    NRC Project Director: Frederick J. Hebdon

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: August 29, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications for allowable values and trip 
setpoints for selected plant process instrumentation. The new allowable 
values/setpoints are in accordance with the instrument setpoint 
methodology accepted by the NRC staff in a letter dated July 18, 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed revised Trip Setpoints and Allowable Values are 
more conservative than those currently approved in the Technical 
Specifications. Therefore, any proposed system or component 
actuations will occur earlier, resulting in a more conservative 
plant response. Thus, the proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change to the Technical Specifications does not 
introduce any new components nor does it modify the design of any 
existing components. Other than making Trip Setpoints and Allowable 
Values of existing instrumentation more conservative, the change 
does not affect the design or function of any plant system, 
structure, or component, nor does it change the way plant systems 
are operated. Thus, the possibility of a new or different kind of 
accident previously evaluated is not created.
    3. The proposed change does not result in a significant 
reduction in the margin of safety.
    Since the proposed revised Trip Setpoints and Allowable Values 
are more conservative than the existing values, the margin of safety 
would be increased by issuance of the changes. Thus, the proposed 
change does 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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: November 2, 1995
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow 120 volt AC buses EV-1-A 
and EV-1-B to be energized from either their normal inverter power 
supply or from their alternate power supply.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated:
    These buses are not used as the initiator of any analyzed 
accidents. Therefore, the probability of any previously evaluated 
accident has not increased. If an accident were to occur while the 
buses are supplied from the alternate power supply, there would 

[[Page 62497]]
be no change in the analyzed accident scenario since even in the event 
of a loss of offsite power event, the safety functions would be 
completed. Thus, the consequences of any previously evaluated 
accident have not increased.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated:
    The proposed change introduces no new mode of plant operation 
and it does not involve physical modification to the plant. 
Therefore, it does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety:
    This change does not involve a significant reduction in a margin 
of safety since the proposed change maintains a safety related, 
diesel-backed power supply to these buses whether the power is 
supplied from the inverters or from the alternate power supply. If a 
loss of offsite power event were to occur while the buses were 
supplied from the alternate power source, the safety functions being 
performed by components supplied from these buses would occur. Thus, 
there has been no reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of amendment request: November 2, 1995
    Description of amendment request: The proposed amendment to the 
Perry Nuclear Power Plant Technical Specifications revises those 
specifications associated with handling irradiated fuel in Primary 
Containment and the Fuel Handling Building, and selected specifications 
associated with CORE ALTERATIONS. Specifically, analysis identifies 
that only recently irradiated fuel contains sufficient fission 
products to require OPERABILITY of accident mitigation features to meet 
the accident analysis assumptions. Analyses also show that accident 
mitigation features such as building INTEGRITY and engineered safety 
feature (ESF) ventilation systems are not required for CORE ALTERATION 
events.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
     The proposed requirements are imposed during specific 
activities which can be postulated to result in significant 
radioactive releases. The proposed APPLICABILITY requirements are 
consistent with either the original design basis analyses or with 
revised analyses performed to support this proposed amendment. 
Because the equipment controlled by the revised Specifications is 
not considered an initiator to any previously analyzed accident, 
inoperability of the equipment cannot increase the probability of 
any previously evaluated accident.
    Consistent with the original design basis analysis, the reanalysis 
concludes that radiological consequences of the fuel handling accident 
are well within the 10 CFR 100.11 limits, as defined by acceptance 
criteria in Standard Review Plan Section 15.7.4. The reanalysis has 
previously been submitted to the Nuclear Regulatory Commission for 
review, and NRC confirmatory calculations reached consistent results 
(reference NRC Safety Evaluation for License Amendment No. 35). The 
results of the CORE ALTERATION events other than the fuel handling 
accident remain unchanged from the original design basis, which showed 
that these events do not result in fuel cladding integrity damage or 
radioactive releases. Therefore, the proposed changes do not 
significantly increase the consequences of any previously evaluated 
accident.
    Based on the above, the proposed changes do not significantly 
increase the probability or consequences of any accident previously 
evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed requirements are imposed when specific activities 
represent situations where significant radioactive releases can be 
postulated. The proposed APPLICABILITY requirements are consistent 
with design basis analyses. The proposed changes do not introduce 
any new modes of plant operation and do not involve physical 
modifications to the plant. Therefore, the proposed changes do not 
create the possibility of a new or different kind of accidident from 
any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
     The proposed change imposes controls to ensure that during 
performance of activities which represent situations where 
radioactive releases are postulated, the radiological consequences 
are at or below the established licensing limit. Safety margins and 
analytical conservatisms have been evaluated and are well 
understood. Substantial conservatism is retained to ensure that the 
analysis adequately bounds all postulated event scenarios. The 
current margin of safety is retained.
     Specifically, the margin of safety for the fuel handling 
accident is the difference between the 10 CFR 100 limits and the 
licensing limit defined by the Standard Review Plan (NUREG 0800), 
Section 15.7.4. The licensing limit is defined by the Standard 
Review Plan as being well within the 10 CFR 100 limits, with 
``well within'' defined as 25% of the 10 CFR 100 limits for the fuel 
handling accident. Excess margin is the difference between the 
postulated doses and the corresponding licensing limit. In the NRCs 
initial licensing review of the Perry Nuclear Power Plant (NUREG-
0887, Section 15.3.3), the NRC accepted the design and analyses 
based on the results of the analyses being well within the guideline 
values of 10 CFR 100.
    The proposed APPLICABILITY requirements continue to ensure that 
the whole-body and thyroid doses at the exclusion area and low 
population zone boundaries as well as control room doses are at or 
below the corresponding licensing limit. The margin of safety is 
unchanged; therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The margin of safety for the CORE ALTERATION events other than 
the fuel handling accident discussed above also remains the same as 
in the original design basis analyses, since the proposed changes do 
not impact on the Technical Specification requirements for systems 
needed to prevent or mitigate such CORE ALTERATION events.
    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: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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 

[[Page 62498]]
requirements of the Atomic Energy Act of 1954, as amended (the Act), 
and the Commission's rules and regulations. The Commission has made 
appropriate findings as required by the Act and the Commission's rules 
and regulations in 10 CFR Chapter I, which are set forth in the license 
amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for A Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document rooms for 
the particular facilities involved.

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

    Date of application for amendments: October 6, 1995, and 
supplemented November 20, 1995
    Brief description of amendments: The amendments revise the 
Technical Specifications by incorporating a new acceptance criterion 
for steam generator tubes with degradation in the tubesheet roll 
expansion region.
    Date of issuance: November 21, 1995
    Effective date: November 21, 1995
    Amendment Nos.: 172 and 159
    Facility Operating License Nos. DPR-39 and DPR-48: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 16, 1995 (60 FR 
53648) The supplemental letter provided clarifying information that did 
not affect the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 21, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085

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

    Date of application for amendment: August 10, 1995
    Brief description of amendment: The amendment revises the Haddam 
Neck Technical Specification Section 3/4.4.3, ``Pressurizer,'' to add a 
footnote to allow the pressurizer level to be controlled, outside of 
the programmed level, between 25 to 50 percent, plus or minus 5 percent 
in Mode 3 when the reactor coolant system is borated to the required 
Mode 5 concentrations.
    Date of Issuance: November 14, 1995
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 186
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52928) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated November 14, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457

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

    Date of application for amendments: September 13, 1995, as 
supplemented October 16,1995
    Brief description of amendments: These amendments revise the 
Administrative Controls section of the BVPS-1 and BVPS-2 TSs to make 
them consistent with the requirements of the Offsite Dose Calculation 
Manual (ODCM). The ODCM was recently updated to reflect the radioactive 
liquid and gaseous effluent release limits and the liquid holdup tank 
activity limit of BVPS-1 License Amendment No. 188 and BVPS-2 License 
Amendment No. 70 which were issued June 12, 1995.
    Date of issuance: November 21, 1995
    Effective date: As of the date of issuance, to be implemented 
within 10 days.
    Amendment Nos.: 194 and 77
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 22, 1995 (60 
FR 49292) The October 16, 1995, letter did not change the initial 
proposed no significant hazards consideration determination or expand 
the amendment request beyond the scope of the September 22, 1995, 
Federal Register notice. The Commission's related evaluation of the 
amendments is contained in a Safety Evaluation dated November 21, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001

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

    Date of amendment request: February 14, 1994, as supplemented by 
letters dated July 25, August 15, and August 29, 1995
    Brief description of amendment: The amendment changes the Appendix 
A Technical Specifications (TSs) to make them consistent with the 
revised 10 CFR Part 20, Standards for Protection Against Radiation.
    Date of issuance: November 17, 1995
    Effective date: November 17, 1995
    Amendment No.: 116
    Facility Operating License No. NPF-38. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14888) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 17, 1995. The July 25, 
August 15, and August 29, 1995 letters provided clarifying information 
that did not change the initial propose no significance hazards 
consideration determination.
    No significant hazards consideration comments received: No
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, LA 70122

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: May 12, 1995, as supplemented 
by letters dated July 6 and October 2, 1995. 

[[Page 62499]]

    Brief description of amendments: The amendments revise Technical 
Specification Surveillance Requirement 4.6.1.2 to add the provision 
that 10 CFR Part 50, Appendix J, applies, except as modified by NRC-
approved exemptions.
    Date of issuance: November 17, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 91 and 69
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35078) The July 6 and October 2, 1995, letters provided clarifying 
information that did not change the scope of the May 12, 1995, 
application and initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 17, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

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

    Date of application for amendment: July 28, 1995, as supplemented 
September 12, October 18, and October 31, 1995.
    Brief description of amendment: In order to support a full-core 
offload as a normal end-of-cycle event, the amendment adds License 
Condition 2.C(6) and will require that: (1) the reactor be subcritical 
for at least 100 hours prior to the start of reactor refueling 
operations, (2) the spent fuel pool bulk temperature be maintained less 
than or equal to 140F, and (3) two trains of shutdown cooling be 
operable during reactor refueling operations.
    Date of issuance: November 9, 1995
    Effective date: As of the date of issuance.
    Amendment No.: 89
    Facility Operating License No. DPR-21. Amendment revised the 
license.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45180) The September 12, October 18, and October 31, 1995, submittals 
provided additional information that did not change the initial 
proposed no significant hazards consideration determination. The 
Commission's related evaluation of the amendment and Final No 
Significant Hazards Consideration Determination are contained in a 
Safety Evaluation dated November 9, 1995.
    No significant hazards consideration comments received: No public 
comments received. A request for a hearing was received from We the 
People, the Seacoast Anti-Pollution League, the New England Coalition 
on Nuclear Pollution, and Donald Del Core of Uncasville, Connecticut.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360

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

    Date of application for amendment: October 6, 1995, supplemented 
October 23, November 2, and November 15, 1995.
    Brief description of amendment: The amendment adds footnotes to 
Action Statement (AS) 3.8.1.1.a of the Technical Specification (TS) and 
its bases to allow a one-time extension of the allowed outage time 
(AOT) for an inoperable offsite power source from the current 72 hours 
to 7 days.
    Date of issuance: November 22, 1995
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.:  192
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 17, 1995 (60 FR 
53812). The October 23, November 2, and November 15, 1995, letters 
provided clarifying information and slight modifications to the 
original request that were not outside the scope of the original notice 
and 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 November 22, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location:  Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360

Northern States Power Company, Docket No. 50-282, Prairie Island 
Nuclear Generating Plant, Unit No. 1, Goodhue County, Minnesota

    Date of application for amendment: January 10, 1995, as 
supplemented August 9 and September 20, 1995.
    Brief description of amendment: The amendments revise the Prairie 
Island event monitoring instrumentation Technical Specifications and 
associated Bases to conform to Standard Technical Specifications for 
post-accident monitoring.
    Date of issuance: November 9, 1995
    Effective date: November 9, 1995, with full implementation within 
30 days.
    Amendment Nos.: 121/114
    Facility Operating License No. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8753) The August 9 and September 20, 1995, letters provided updated 
Technical Specification pages and clarifying information in response to 
discussions with the staff during various teleconferences conducted 
during the review process. This information was within the scope of the 
original application and did not change the staff's initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated November 9, 1995.
    No Significant hazards consideration comments received: No
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401

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: December 2, 1994, as 
supplemented May 12, 1995.
    Brief description of amendments: These amendments relocate the fire 
protection requirements from the Technical Specifications to the 
Updated Final Safety Analysis Report in accordance with the guidance in 
Generic Letter (GL) 86-10, ``Implementation of Fire Protection 
Requirements,'' and GL 88-12, ``Removal of Fire Protection Requirements 
from Technical Specifications.''
    Date of issuance: November 20, 1995 Effective date: As of date of 
issuance, both units, to be implemented within 30 days.
    Amendment Nos.: 104 and 68
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications and the License.
    Date of initial notice in Federal Register: April 26, 1995 (60 FR 
20524) The supplemental letter provided clarifying information and did 
not 

[[Page 62500]]
change the initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 20, 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: September 14, 1995 and 
supplemented by letter dated October 27, 1995
    Brief description of amendments: These amendments revise the 
technical specifications by deleting Reactor Enclosure and Refueling 
Area Secondary Containment Isolation Valve Tables 3.6.5.2.1-1 and 
3.6.5.2.2-1, and references to them, in accordance with Generic Letter 
91-08, ``Removal of Component lists from Technical Specifications.'' 
The TS have been modified to state requirements in general terms that 
include the components listed in the tables removed from the TS.
    Date of issuance: November 20, 1995
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: November 20, 1995
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52934) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 20, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464

Tennessee Valley Authority, Docket No. 50-296, Browns Ferry Nuclear 
Plant, Unit 3, Limestone County, Alabama

    Date of application for amendments: October 4, 1995 (TS 368)
    Brief description of amendment: The amendment delete requirements 
for daily checks for certain instruments that do not have indications, 
and provides editorial changes.
    Date of issuance: November 13, 1995
    Effective Date: November 13, 1995
    Amendment No.: 202
    Facility Operating License No. DPR-68: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52935) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 13, 1995.
    No significant hazards consideration comments received: None
    Local Public Document Room location:  Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendments: August 7, 1995 (TS 95-03)
    Brief description of amendments: The amendments address operation 
with a rod urgent failure condition, including limited operation with 
one control or shutdown bank inserted up to 18 steps below its 
insertion point. In addition, the surveillance interval for rod 
movement verifications has been increased from 31 to 92 days.
    Date of issuance: November 21, 1995
    Effective date: November 21, 1995
    Amendment Nos.: 215 and 205
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: August 30, 1995 (60 FR 
45186) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 21, 1995.
    No significant hazards consideration comments received: None
    Local Public Document Room location:  Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

The Cleveland Electric Illuminating Company, Centerior Service 
Company, Duquesne Light Company, Ohio Edison Company, Pennsylvania 
Power Company, Toledo Edison Company, Docket No. 50-440, Perry 
Nuclear Power Plant, Unit No. 1, Lake County, Ohio

    Date of application for amendment: April 28, 1995
    Brief description of amendment: The amendment removes the license 
conditions for the Transamerica Delaval, Inc. emergency diesel 
generators specified by paragraph 2.C.(9) and defined in Attachment 2 
to the Operating License.
    Date of issuance: November 16, 1995
    Effective date: November 16, 1995
    Amendment No.: 74
    Facility Operating License No. NPF-58: This amendment revises the 
license.
    Date of initial notice in Federal Register: June 6, 1995 (60 FR 
29889) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 16, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081

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

    Date of application for amendment: June 23, 1995, and facsimile 
transmission dated October 31, 1995
    Brief description of amendment: This amendment relocates TS 3/
4.3.3.3, ``Seismic Instrumentation;'' TS 3/4.3.3.4, ``Meteorological 
Instrumentation;'' and TS 3/4.4.11, ``Reactor Coolant System Vents;'' 
and the Bases for each of the three sections from the TS to the Updated 
Safety Analysis Report, and eliminates the special reporting 
requirements for inoperable seismic and meteorological monitoring 
instrumentation from TS 6.9.2.
    Date of issuance: November 14, 1995 Effective date: November 14, 
1995, and shall be implemented not later than 90 days after issuance.
    Amendment No.: 201
    Facility Operating License No. NPF-3: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39455) The October 31, 1995, facsimile transmission was clarifying in 
nature and did not affect the initial no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated November 14, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606

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

    Date of application for amendment: June 7, 1995
    Brief description of amendment: This amendment revises Technical 
Specification 3/4.9.4, Refueling Operations - Containment Penetrations; 


[[Page 62501]]
Bases 3/4.9.4, Containment Penetrations; and Limiting Condition for 
Operation (LCO) 3.9.4.b to allow both doors of the containment 
personnel airlock to be open during core alterations or movement of 
irradiated fuel within the containment, provided that certain specified 
conditions are meet. Additional changes revise or clarify TS LCO 
3.9.4.c, TS Action 3.9.4.a, and TS Surveillance Requirement 4.9.4, and 
modify the associated Bases.
    Date of issuance: November 17, 1995
    Effective date: November 17, 1995
    Amendment No.: 202
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39454) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 17, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location:  University of Toledo, William 
Carlson Library, Government Documents Collection, 2801 West Bancroft 
Avenue, Toledo, Ohio 43606

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

    Date of amendment request: December 6, 1994
    Brief description of amendments: These changes revise Technical 
Specifications to allow appropriate remedial action for high 
particulate levels in the diesel generator fuel oil inventory and other 
out-of-limit properties in new diesel generator fuel oil that has been 
added to the existing diesel generator fuel oil storage inventory.
    Date of issuance: November 17, 1995
    Effective date: November 17, 1995
    Amendment Nos.: Unit 1 - Amendment No. 43; Unit 2 - Amendment No. 
29
    Facility Operating License Nos. NPF-87 and NPF-89. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 1, 1995 (60 FR 
6311) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 17, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, Texas 76019.

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

    Date of amendment request: January 13, 1995
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) 3.3.1 and 3.3.2 to relocate Tables 3.3-2 and 3.3-5, 
which provide the response time limits for the reactor trip system and 
the engineered safety features actuation system instruments, from the 
TS to the updated Final Safety Analysis Report (FSAR). The amendment 
also relocates the Bases discussion for TS 3.3.1 and TS 3.3.2 to 
Section 16.3 of the updated FSAR.
    Date of issuance: November 22, 1995
    Effective date: November 22, 1995, to be implemented within 30 days 
of issuance.
    Amendment No.: 104
    Facility Operating License No. NPF-30. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 15, 1995 (60 
FR 8741) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 22, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room locations: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251

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

    Date of application for amendment: June 6, 1995
    Brief description of amendment: The amendment modifies the Index of 
the WNP-2 Technical Specifications by deleting reference to the Bases 
pages.
    Date of issuance: November 24, 1995
    Effective date: November 24, 1995
    Amendment No.: 143
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 19, 1995 (60 FR 
37102) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 24, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352

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

    Date of application for amendments: September 13, 1995, and October 
19, 1995, as supplemented by letter dated October 25, 1995
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.1, ``Definitions,'' TS Section 15.3.1.G, 
``Operational Limitations'' (and basis), and TS Figure 15.2.1-2, 
``Reactor Core Safety Limits, Point Beach Unit 2.'' The changes reduce 
the reactor coolant system raw measured total flow rate limit and 
reflect new reactor core safety limits for Unit 2.
    Date of issuance: November 17, 1995
    Effective date: November 17, 1995
    Amendment Nos.: 165 and 169
    Facility Operating License Nos. DPR-24 and DPR-27: Amendments 
revised the Technical Specifications. Public comments requested as to 
proposed no significant hazards consideration: Yes (60 FR 54527 dated 
October 24, 1995). That notice provided an opportunity to submit 
comments on the Commission's proposed no significant hazards 
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by November 24, 
1995, but indicated that if the Commission makes a final no significant 
hazards consideration determination any such hearing would take place 
after issuance of the amendment. The Commission's related evaluation of 
the amendment, finding of exigent circumstances, and final 
determination of no significant hazards consideration is contained in a 
Safety Evaluation dated November 17, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

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

    Date of amendment request: September 14, 1995
    Brief description of amendment: The amendment revised Technical 
Specification 3/4.5.5 to increase the allowed outage time for 
adjustment of boron concentration for the refueling water storage tank 
from 1 hour to 8 hours.
    Date of issuance: November 13, 1995 
    
[[Page 62502]]

    Effective date: November 13, 1995, to be implemented within 30 days 
of issuance.
    Amendment No.: 91
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52936) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 13, 1995.
    No significant hazards consideration comments received: No
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (exigent public announcement or emergency 
circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By January 5, 1996, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above. 

[[Page 62503]]

    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).

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

    Date of application for amendments: November 9, 1995, as 
supplemented by letters dated November 13, 1995, and November 16, 1995
    Brief description of amendments: These amendments revise Technical 
Specification (TS) Section 15.4.2, ``In-Service Inspection of Safety 
Class Components,'' to incorporate a new steam generator tube 
acceptance criterion for the Unit 2 steam generators. This criterion 
allows tubes that are degraded or defective in a location (within the 
tubesheet) that does not affect the structural integrity of the tube to 
remain in service. The applicable basis is also changed.
    Date of issuance: November 22, 1995
    Effective date: November 22, 1995
    Amendment Nos.: 166 and 170
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications. Public comments requested as to 
proposed no significant hazards consideration: No The Commission's 
related evaluation of the amendments, finding of emergency 
circumstances, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated November 22, 
1995.
    No significant hazards consideration comments received: No
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.
    Attorney for licensee: Ernest L. Blake, Jr., Shaw, Pittman, Potts & 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: Gail H. Marcus
    Dated at Rockville, Maryland, this 29th day of November 1995.
    For the Nuclear Regulatory Commission
Elinor G. Adensam,
Deputy Director, Division of Reactor Projects - III/IV, Office of 
Nuclear Reactor Regulation
[Doc. 95-29540 Filed 12-5-95; 8:45 am]
BILLING CODE 7590-01-F