[Federal Register Volume 61, Number 40 (Wednesday, February 28, 1996)]
[Notices]
[Pages 7542-7568]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-4342]



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NUCLEAR REGULATORY COMMISSION

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from February 5, 1996, through February 15, 1996. 
The last biweekly notice was published on February 14, 1996 (61 FR 
5809). 

[[Page 7543]]


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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike, 
Rockville, Maryland from 7:30 a.m. to 4:15 p.m. Federal workdays. 
Copies of written comments received may be examined at the NRC Public 
Document Room, the Gelman Building, 2120 L Street, NW., Washington, DC. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By March 29, 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 

[[Page 7544]]
Document Room, the Gelman Building, 2120 L Street, NW., Washington DC, 
by the above date. Where petitions are filed during the last 10 days of 
the notice period, it is requested that the petitioner promptly so 
inform the Commission by a toll-free telephone call to Western Union at 
1-(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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: December 20, 1995.
    Description of amendments request: The proposed amendment would 
change the instrumentation setpoint for the reactor trip and main steam 
isolation signal (MSIS) actuation on low steam generator pressure from 
greater than or equal to 919 psia with an allowable value of greater 
than or equal to 911 psia to greater than or equal to 895 psia with an 
allowable value of greater than or equal to 890 psia.
    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 amendment does not involve any change to the method 
of operation of any plant equipment that is used to mitigate the 
consequences of an accident. The proposed change only affects the 
instrument setpoint for steam generator low pressure reactor trip 
and MSIS actuation. The proposed setpoint meets the requirement of 
ensuring a reactor trip and MSIS actuation prior to steam generator 
pressure reaching the analytical limits even under worst-case 
accident conditions. Thus, the proposed change does not involve a 
significant increase in the probability of an accident previously 
evaluated.
    The proposed amendment does not alter any of the assumptions or 
bounding conditions currently in the UFSAR [updated final safety 
analysis report] and meets the requirement of ensuring a reactor 
trip and MSIS actuation prior to steam generator pressure reaching 
the analytical setpoint under worst-case accident conditions. As a 
result, the proposed amendment does not involve a significant 
increase in the consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve any change to the method of 
operation of any plant equipment that is used to mitigate the 
consequences of an accident. Accordingly, no new failure modes have 
been defined for any plant system or component important to safety 
nor has any new limiting failure been identified as a result of the 
proposed change. The intent of the proposed change is to increase 
the margin between normal operating parameters and trip setpoints. 
This minimizes the possibility of unnecessary challenges to safety 
systems improving the safety of operation. The method of protecting 
the facility for an excess steam demand event remains unchanged and 
therefore, the amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change is the implementation of a setpoint value 
which was derived using methodologies endorsed by Revision 2 of NRC 
Regulatory Guide 1.105, ``Instrument Setpoints.'' The new setpoint 
ensures that sufficient margin exists below the full load operating 
value for steam pressure so as not to interfere with normal plant 
operation, but still high enough to provide the required protection 
(reactor trip and main steam line isolation) in the event of an 
excessive steam demand event. The new setpoint ensures that safety 
margins are maintained within the results of existing calculations. 
The margin of safety between the analyzed trip value and the point 
at which safety analysis results become unacceptable remain 
unchanged since the analytical setpoints are not affected by the 
amendment. The new setpoint resulted from the reduced instrument 
uncertainty and will ensure that the reactor trip and MSIS actuation 
on low steam generator pressure will occur before the analyzed value 
and hence, this change does not involve a significant reduction in 
the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
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.

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: January 5, 1996.
    Description of amendments request: The proposed amendment would 
revise paragraph 2.C.(1) of the operating licenses and Section 1.26 of 
the TS for each of the three PVNGS Units to increase the authorized 100 
percent reactor core power (rated thermal power) from 3800 megawatts 
thermal (Mwt) to 3876 Mwt, an increase of 2 percent. The proposed 
amendment would also revise TS 4.1.1.4, TS 3.1.3.4, and TS 3.2.6 
(Figure 3.2-1) to lower the allowable reactor coolant system cold leg 
temperature limits for each of the three PVNGS Units, and revise TS 
3.4.2.1 and TS 3.4.2.2 to lower the pressurizer safety valve setpoints 
for Units 1 and 3 to support the increased power operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed amendment does not change the method of operation 
or modify the plant configuration other than minor changes in 
equipment setpoints. Thus no increase in the probability of an 
accident is created by this amendment. System and programmatic 
reviews have been performed on the nuclear 

[[Page 7545]]
steam supply system controls, reactor coolant system mechanical, steam 
generator mechanical, balance of plant systems, and fire protection, 
equipment qualification, and probabilistic risk assessment programs. 
The conclusion of these reviews was that operation in accordance 
with the changes proposed in this amendment was acceptable and posed 
no significant risk to the health and safety of the public. The 
analyses supporting this amendment demonstrate that the consequences 
of events using the changes specified in the amendment are within 
the criteria which are the current licensing basis for the PVNGS 
Units. Therefore the amendment, as proposed, 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 amendment does not modify the configuration of the 
units except for minor equipment setpoints. No equipment changes and 
no new methods of plant operation are being proposed, therefore, no 
new failure modes are introduced by the proposed amendment. The 
setpoint changes proposed have been evaluated and shown to be 
acceptable in providing their design function. The increased rated 
thermal power and associated changes have been incorporated into the 
safety analysis performed in support of this amendment request and 
the results have been shown to be similar to those previously 
obtained. No possibility of a new or different kind of accident from 
any accident previously evaluated will be created as a result of the 
proposed amendment.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The changes proposed were evaluated in the safety analysis 
performed to justify the amendment request. Although the 
consequences of some events increased slightly, the results continue 
to meet the criteria which form the PVNGS licensing basis. The 
programmatic and system reviews provide further assurance of the 
capability of the units to continue to operate safely with the 
changes proposed in this amendment. Therefore the amendment does not 
involve a significant reduction in a margin of safety.

    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: January 29, 1996.
    Description of amendment request: The proposed change would revise 
the technical specifications (TS) table 4.1-3, item 4 to change the 
frequency of main steam safety valve (MSSV) testing to that specified 
in NUREG-1431, the improved ``Standard Technical Specifications, 
Westinghouse Plants'' (one third of the MSSVs each refueling outage). 
In addition, the licensee proposed adding the MSSV test acceptance 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Neither the valves' nor the system's configuration or functions 
are being altered. The valves' setpoints and their ``as-left'' 
range, +/-1%, will not be changed. The changes are to the testing 
frequency and the ``as found'' tolerance of the MSSV setpoint.
    The proposed changes in testing frequency and the higher 
tolerance are in the less conservative direction, but are not 
significant for several reasons. First, the new standards are based 
on the American Society of Mechanical Engineers (ASME) Boiler and 
Pressure Vessel Code. The new standards have been accepted by the 
nuclear industry and the NRC, and are referenced in the improved 
Standard Technical Specifications. Based on a discussion with the H. 
B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2 MSSV 
manufacturer (i.e., Crosby), HBRSEP, Unit No. 2 has not experienced 
more problems with the Crosby MSSVs than the nuclear industry in 
general, thus, the new level of safety will be equivalent to that of 
the nuclear industry. Second, if a MSSV does fail the surveillance 
test, the proposed TS will require additional MSSVs to be tested. 
This requirement provides assurance that testing will reveal 
possible generic problems. The impact of the tolerance on the 
Chapter 15 accidents was analyzed and found to be within acceptable 
limits.
    Since no Updated Final Safety Analysis Report (UFSAR) Chapter 15 
accident analysis is significantly impacted by the proposed changes, 
there would be no increase in the consequences of an accident 
previously evaluated. The testing in accordance with the ASME Boiler 
and Pressure Vessel Code will provide an adequate level of assurance 
that the MSSVs will be able to perform their intended function; 
therefore the probability of a previously evaluated accident is not 
increased.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    No new systems or equipment are involved with the proposed 
changes; and the plant's configuration and operational procedures 
are unaffected. Since the proposed changes do not impact the plant's 
operation, it can not create a new or different kind of accident.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The change in testing frequency is in a less conservative 
direction, but it is based on the ASME Code and the improved 
Standard Technical Specifications. Since HBRSEP, Unit No. 2 has not 
experienced a greater number of failures associated with these MSSVs 
than the nuclear industry in general, the decrease in the MSSV 
testing frequency will not significantly impact the margin of 
safety. Also, analyses have been performed that demonstrate that the 
impact of the setpoint tolerance change on the UFSAR Chapter 15 
accident analysis results is not significant. Therefore, the 
proposed changes do not involve a significant reduction in a margin 
of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550.
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

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: January 31, 1996.
    Description of amendment request: The proposed change would revise 
the Technical Specifications section 4.4 to allow the use of 10 CFR 
Part 50, Appendix J, Option B, Performance-Based Containment Leakage 
Rate Testing. A new TS section 6.12 is proposed to describe the 
containment leakage rate testing program, committing to meet 10 CFR 
50.54(o) and 10 CFR Part 50, Appendix J, Option B for type A tests; and 
to meet 10 CFR part 50, Appendix J, Option A, for types B and C tests. 
The bases would be changed to reflect the proposed changes. 

[[Page 7546]]

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed change does not involve a significant hazards 
consideration for the following reasons.
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The HBRSEP [H. B. Robinson Steam Electric Plant], Unit No. 2 
Type A testing history provides substantial justification for the 
proposed test schedule change to one test in a 10 year period. Three 
Structural Integrity Tests (SITs) and seven Integrated Leak Rate 
Tests (ILRTs) have been performed with acceptable results. Previous 
testing has affirmed the acceptable reliability of the containment 
structure to minimize leakage as designed, and provides assurance 
that its performance to continuously function as designed is not 
challenged due to this test schedule extension to once in 10 years.
    Therefore, this proposed change to the TS that revises the Type 
A testing frequency does not involve an increase in the probability 
of an accident previously evaluated.
    This proposed change to revise the test schedule frequency does 
not impact nor alter the design of any system, structure or 
component. The limit on allowable leakage is not increased. Type A 
testing provides periodic verification of the leak tight integrity 
of the containment and the systems and components that penetrate the 
containment structure.
    NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' 
provides the technical basis for the NRC's rulemaking to revise 
containment leakage testing requirements for nuclear power reactors 
in 10 CFR 50, Appendix J. Section 10.1.2 of NUREG-1493, ``Summary of 
Technical Findings, Leakage-Testing Intervals,'' states the 
following.
    1. Reducing the frequency of Type A tests (ILRTs) from the 
current three per 10 years to one per 20 years was found to lead to 
an imperceptible increase in risk. The estimated increase in risk is 
very small because ILRTs identify only a few potential containment 
leakage paths that cannot be identified by Type B and C testing, and 
the leaks found by Type A tests have been only marginally above 
existing requirements.
    2. Given the insensitivity of risk to containment leakage rate 
and the small fraction of leakage paths detected solely by Type A 
testing, increasing the interval between ILRTs is possible with 
minimal impact on public risk.
    Therefore, based on the previous Type A test results, the 
proposed change does not involve a significant increase in the 
consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change only incorporates the performance based 
testing approach authorized in 10 CFR 50, Appendix J, Option B, and 
is justified based on previous plant-specific Type A test results. 
Plant structures, systems, and components will not be operated in a 
different manner as a result of this proposed change and no physical 
modifications to equipment are involved. The interval extensions 
allowed by Option B of 10 CFR 50, Appendix J, do not have the 
potential for creating the possibility of new or different type of 
accidents from those previously evaluated.
    3. The proposed change does not involve a significant reduction 
in the margin of safety.
    The proposed change does not change the allowable leak rate from 
the containment, it only allows an extension of the interval between 
the performance of Type A leak rate testing. NUREG-1493, which 
provides the technical basis for the NRC's rulemaking to revise 
containment leakage testing requirements for nuclear power reactors 
in 10 CFR 50, Appendix J. Section 10.1.2 of NUREG-1493, ``Summary of 
Technical Findings, Leakage-Testing Intervals,'' states the 
following.
    ``1. Reducing the frequency of Type A tests (ILRTs) from the 
current three per 10 years to one per 20 years was found to lead to 
an imperceptible increase in risk. The estimated increase in risk is 
very small because ILRTs identify only a few potential containment 
leakage paths that cannot be identified by Type B and C testing, and 
the leaks found by Type A tests have been only marginally above 
existing requirements.
    2. Given the insensitivity of risk to containment leakage rate 
and the small fraction of leakage paths detected solely by Type A 
testing, increasing the interval between ILRTs is possible with 
minimal impact on public risk.''

    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: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

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: January 29, 1996.
    Description of amendment request: The proposed change would revise 
the technical specifications (TS) to: (1) add TS 4.6.1.5 to provide 
criteria for 24-hour full-load testing of the emergency diesel 
generators (EDGs) to be performed during each refueling outage; (2) 
revise TS 4.6.1.2 to allow testing of the EDG protective bypasses 
listed in TS 3.7.1.d to be done independent of the safety injection or 
loss of offsite power testing; and (3) revise TS 4.6.1.3 to include the 
EDG protective bypass inspection and a requirement to inspect the EDGs 
at least once every refueling outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    The proposed changes do not involve a significant hazards 
consideration for the following reasons.
    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability of an accident previously evaluated. The proposed 
changes require additional testing of the EDGs and will change the 
requirement for when the protective bypasses are tested. The 
function of the EDGs remains unchanged. Since the additional testing 
involves the EDGs, which are required to mitigate an accident and 
are not involved in the initiation of an accident, the proposed 
changes will not increase the probability of an accident.
    The proposed changes do not involve a significant increase in 
the consequences of an accident previously evaluated. The proposed 
changes require additional testing to verify the reliability of the 
EDGs and to show the EDGs can withstand maximum accident loading 
conditions. The proposed changes will also require the testing of 
the EDG protective bypasses to be accomplished during EDG outages 
and not during the SI/LOOP testing during a refueling outage. The 
ability of the EDGs to perform their accident mitigation function 
remains unchanged. Therefore, the proposed changes will not increase 
the consequences of an accident.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not create the possibility of a new kind 
of accident from any previously evaluated. The proposed changes are 
an enhancement to the EDG testing requirements. The most significant 
change will require additional testing of the EDGs to demonstrate 
adequate reliability and to determine if the EDGs can withstand 
maximum accident loading conditions. The remaining changes will 
augment the TS to allow on-line EDG inspections and testing. Since 
the function of the EDGs remains unchanged and they are not the 
initiator of an accident, the proposed changes will not 

[[Page 7547]]
create the possibility of a new kind of accident from any previously 
evaluated.
    The proposed changes do not create the possibility of a 
different kind of accident from any accident previously evaluated. 
The proposed changes require additional testing of the EDGs (i.e., 
the 24 hour full-load test) and revise the requirement for testing 
the EDG protective bypasses during the SI/LOOP testing. The 
additional testing of the EDGs will demonstrate sufficient 
reliability and determine if the EDGs can withstand maximum accident 
loading conditions. The EDG protective bypasses will be statically 
tested during an EDG outage thus preventing possible damage to 
equipment from a transient if the protective bypass fails. The 
function of the EDGs remains unchanged by these proposed changes. 
Since the EDGs are required to mitigate an accident and are not the 
initiators of an accident, the proposed changes will not create a 
different kind of accident from any kind of accident previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes do not reduce the margin of safety as 
defined in the TS. The proposed changes are being submitted as an 
enhancement to the testing requirements outlined in the TS. The 
changes include additional testing, revising the requirement to test 
the engine protective bypasses during the SI/LOOP testing and 
clarification of the periodicity of inspecting the EDGs. The 
additional testing demonstrates increased reliability and determines 
that the EDGs can cope with maximum accident loading. The remaining 
proposed changes provide clarification as to when the EDG 
inspections and testing are required. The ability of the EDGs to 
perform their function will not be reduced. Therefore, the margin of 
safety will not be reduced by the proposed changes.

    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: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602.
    NRC Project Director: David B. Matthews.

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

    Date of amendment request: December 6, 1995.
    Description of amendment request: The proposed amendment would 
change the technical specifications of these plants to incorporate 10 
CFR Part 50, Appendix J, ``Primary Reactor Containment Leakage Testing 
for Water-Cooled Power Reactors'', Option B.
    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:

    ComEd proposes to revise Byron Nuclear Power Station, Units 1 
and 2 (Byron), and Braidwood Nuclear Power Station, Units 1 and 2 
(Braidwood) Technical Specification (TS) Section 3/4.6.1, ``Primary 
Containment,'' and the associated Bases to reflect recent changes to 
Appendix J to 10 CFR 50, ``Primary Reactor Containment Leakage 
Testing for Water-Cooled Power Reactors.'' The proposed revisions 
include:
    1. Adding TS Definitions 1.15.a for the maximum allowable 
primary containment leakage rate (La) and 1.20.a for the 
maximum calculated primary containment pressure (Pa). The 
redundant definitions throughout TS Section 3/4.6.1 are deleted,
    2. Adding numerous statements throughout TS Section 3/4.6.1 that 
leak rate testing is performed in accordance with Regulatory Guide 
(RG) 1.163, Revision 0, ``Performance-Based Containment Leak-Test 
Program,'' and its referenced documents,
    3. Deleting TS requirements that are taken verbatim from 10 CFR 
50, Appendix J. The specific requirements will be placed in the 
containment leakage rate test program in accordance with RG 1.163, 
and its referenced documents, and
    4. Clarifying Technical Specification Surveillance Requirement 
(TSSR) 4.6.1.1.a for consistency with NUREG-1431, Revision 1, 
``Standard Technical Specifications for Westinghouse Plants.''
    A. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    10 CFR 50, Appendix J, has been amended to include provisions 
regarding performance-based leakage testing requirements (Option B). 
Option B allows plants with satisfactory Integrated Leak Rate 
Testing (ILRT) performance history to reduce the Type A testing 
frequency from three tests in ten years to one test in ten years. 
For Type B and Type C tests, Option B allows plants to reduce 
testing frequency based on the leak rate test history of each 
component. In addition, Option B establishes controls to ensure 
continued satisfactory performance of the affected penetrations 
during the extended testing interval. To be consistent with the 
requirements of Option B to 10 CFR 50, Appendix J, ComEd proposes to 
include appropriate changes to the TSs that incorporate the 
necessary revisions.
    Some of the proposed changes represent minor curtailments to 
current TS requirements, but are based on the requirements specified 
by Option B to 10 CFR 50, Appendix J. Any such changes are 
consistent with the current plant safety analyses and have been 
determined to represent sufficient requirements for the assurance of 
the reliability of equipment assumed to operate in the safety 
analyses, or provide continued assurance that specified parameters 
associated with containment integrity remain within their acceptance 
limits. The other proposed changes maintain consistency with those 
requirements specified by Option B to 10 CFR 50, Appendix J and are 
consistent with the current plant safety analyses. Implementation of 
these changes will provide continued assurance that specified 
parameters associated with containment integrity will remain within 
their acceptance limits, and as such, will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    The associated systems affecting the leak rate integrity are not 
assumed in any safety analyses to initiate any accident sequence; 
therefore, the probability of occurrence of any accident previously 
evaluated is not increased. In addition, the proposed changes to the 
limiting conditions for operation and surveillance requirements for 
such systems are consistent with the current 10 CFR 50, Appendix J, 
requirements. The proposed changes maintain an equivalent level of 
reliability and availability for all affected systems.
    Maintaining allowable leakage within the analyzed limit assumed 
for the accident analyses does not adversely affect either the 
onsite or offsite dose consequences. Furthermore, containment 
leakage is not an accident initiator. As such, there is no adverse 
impact on the probability of accident initiators. Thus, there is no 
significant increase in the probability or occurrence of any 
previously analyzed accident, or increase the consequences of any 
previously analyzed accident.
    B. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Option B of 10 CFR 50, Appendix J, specifies, in part, that a 
Type A test may be conducted at a periodic interval based on the 
performance of the overall containment system. Type A tests measure 
both the containment system overall integrated leakage rate at the 
containment pressure boundary and system alignments assumed during a 
large break loss-of-coolant accident (LOCA), and demonstrate the 
capability of the primary containment to withstand an internal 
pressure load. The acceptable leakage rates are specified in the 
TSs. For Type B and C tests, intervals are proposed for 
establishment based on the performance history of each component. 
Acceptance criteria for each component are based upon demonstration 
that the leakage rates at design basis pressure conditions for 
applicable penetrations are within the limits specified in the TSs.
    The proposed changes reflect the requirements specified in the 
amended 10 CFR 50, Appendix J, and are consistent with the current 
plant safety analyses. Some minor curtailments of current TS 
requirements are 

[[Page 7548]]
based on generic guidance or similarly approved provisions for other 
plants. These changes do not involve revisions to the design of the 
plant. Some of the changes may involve revision in the testing of 
components at the plant; however, these are in accordance with the 
current plant safety analyses and provide for appropriate testing or 
surveillance that is consistent with Option B to 10 CFR 50, Appendix 
J. The proposed changes will not introduce new failure mechanisms 
beyond those already considered in the current plant safety 
analyses.
    No new modes of operation are introduced by the proposed 
changes. Surveillance requirements are changed to reflect 
corresponding changes associated with Option B to 10 CFR 50, 
Appendix J. The proposed changes maintain at least the present level 
of operability of any such system that affects plant containment 
integrity. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated. The associated systems that affect plant leak 
rate integrity related to the proposed amendment are not assumed to 
initiate any accident sequence. In addition, the proposed 
surveillance requirements for any such affected systems are 
consistent with the current requirements specified within the TSs 
and are consistent with the requirements of Option B to 10 CFR 50, 
Appendix J. The proposed surveillance requirements maintain an 
equivalent level of reliability and availability of all affected 
systems and, therefore, do not affect the consequences of any 
previously evaluated accident. As such, the probability of systems 
associated with leak rate test integrity failing to perform their 
intended function is unaffected by the proposed limiting conditions 
for operation and surveillance requirements.
    C. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The provisions specified in Option B to 10 CFR 50 Appendix J, 
allows changes to Type A, B, and C test intervals based upon the 
performance of past leak rate tests. The effect of extending 
containment leak rate test intervals is a corresponding increase in 
the likelihood of containment leakage. The degree to which intervals 
can be extended has a direct impact on the potential effect on 
existing plant safety margins and the public health and safety that 
can occur due to an increased likelihood of containment leakage.
    Changing Type A, B, and C test intervals from those currently 
provided in the TS to those provided for in 10 CFR 50, Appendix J, 
Option B, slightly increases the risk associated with Type A, B, and 
C specific accident sequences. Historical data suggest that 
increasing the Type C test interval can slightly increase the 
associated risk; however, this is compensated by the corresponding 
risk reduction benefits associated with reduction in component 
cycling, stress, and wear associated with increased test intervals. 
In addition, when considering the total integrated risk, which 
includes all analyzed accident sequences, the additional risk 
associated with increasing test intervals is negligible.
    The proposed changes are consistent with those provisions 
specified in Option B of 10 CFR 50, Appendix J, and are consistent 
with current plant safety analyses. In addition, these proposed 
changes do not involve revisions to the design of the plant. As 
such, the proposed individual changes will maintain the same level 
of reliability of the equipment associated with containment 
integrity, assumed to operate in the plant safety analysis, or 
provide continued assurance that specified parameters affecting 
plant leak rate integrity, will remain within their acceptance 
limits. Therefore, the proposed changes provide continued assurance 
of the leakage integrity of the containment without adversely 
affecting the public health and safety and, as such, will not 
significantly reduce existing plant safety margins.
    The proposed changes are based on United States Nuclear 
Regulatory Commission (USNRC) accepted provisions and maintain 
necessary levels of system or component reliability affecting plant 
containment integrity. The performance-based approach to leakage 
rate testing concludes that the impact on public health and safety 
due to revised testing intervals is negligible. The proposed changes 
will not reduce the availability of systems associated with 
containment integrity when they are required to mitigate accident 
conditions; therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: For Byron, the Byron Public 
Library District, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; 
for Braidwood, the Wilmington Public Library, 201 S. Kankakee Street, 
Wilmington, Illinois 60481.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603.
    NRC Project Director: Robert A. Capra.

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

    Date of amendment requests: December 4, 19, 19, 20, 20, and 20, 
1995.
    Description of amendment request: Each proposed amendment would 
change the surveillance requirement frequency from the current once per 
18-month interval to once per 24-month which is the proposed length of 
a Haddam Neck refueling cycle. The changes pertain to the following 
equipment:
    December 4, 1995, Reactivity control systems flow paths, rod 
position indication system, and Rod drop time.
    December 19, 1995, Containment Air Recirculation System.
    December 19, 1995, Main steam line (MSL) Code Safety Valves self 
actuation, auxiliary feedwater system, service water system, snubber 
testing, feedwater isolation valve actuation, and primary auxiliary 
building cleanup system.
    December 20, 1995, reactor coolant system (RCS) interlock, 
containment sump, High Pressure Safety Injection Pump and Low Pressure 
Safety Injection autostart and alignment, containment spray, and PH 
control.
    December 20, 1995, Trip actuating devices and channel trips, 
reactor trip system, reactor trip system instrumentation, and accident 
monitoring instrumentation.
    December 20, 1995, RCS flow indicators, Loop stop valve interlock, 
Pressurizer code safety valves, Emergency power supply for the 
pressurizer heaters, Containment main sump and volume control tank 
(VCT) level monitoring system, RCS pressure boundary valves, Low 
temperature overpressure protection (LTOP) system, and RCS vent path.
    Basis for proposed no significant hazards consideration 
determination: The Commission has made a proposed determination that 
the amendment request involves 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. 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:
    1. The changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes to surveillance requirements of the Haddam 
Neck Plant Technical Specifications extend the frequency for checking 
the operability of the affected components/equipment. The proposal 
would extend the frequency from at least once per 18 months to at least 
once each refueling interval (i.e., nominal 24-months).

[[Page 7549]]

    Changing the frequency of surveillance requirements from at least 
once per 18 months to at least once each refueling interval does not 
change the basis for the frequency. The frequency was chosen because of 
the need to perform this verification under the conditions that apply 
during a plant outage, and to avoid the potential of an unplanned 
transient if the surveillance were conducted with the plant at power.
    The proposed changes do not alter the intent or method by which the 
surveillance are conducted, do not involve any physical changes to the 
plant, do not alter the way any structure, system, or component 
functions, and do not modify the manner in which the plant is operated. 
As such, the proposed changes in the frequency of surveillance 
requirements will not degrade the ability of the equipment/components 
to perform its safety function.
    Additional assurance of the operability of the components/equipment 
is provided by additional surveillance requirements (e.g., monthly or 
quarterly surveillance).
    Equipment performance over the last four operating cycles was 
evaluated to determine the impact of extending the frequency of 
surveillance requirements. This evaluation included a review of 
surveillance results, preventive maintenance records, and the frequency 
and type of corrective maintenance. It concluded that there is no 
indication that the proposed extension could cause deterioration in the 
condition or performance of any of the subject components.
    In addition to the substantive changes, there are format changes 
which are merely editorial and because format changes produce no 
physical change they do not influence the probability or consequences 
of accidents.
    Since the proposed changes only affect the surveillance frequency 
for safety systems that are used to mitigate accidents, the changes 
cannot affect the probability of any previously analyzed accident. 
While the proposed changes can lengthen the intervals between 
surveillance, the increases in intervals has been evaluated and it is 
concluded that there is no significant impact on the reliability or 
availability of the safety system and consequently, there is no impact 
on the consequences on any analyzed accident.
    2. The changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The proposed changes to surveillance requirements of the Haddam 
Neck Plant Technical Specifications extend the frequency for verifying 
the operability of the affected components/equipment. The proposal 
would extend the frequency from at least once per 18 months to at least 
once each refueling interval (nominal 24 months).
    Changing the frequency of surveillance requirements from at least 
once per 18 months to at least once each refueling interval does not 
change the basis for the frequency. The frequency was chosen because of 
the need to perform this verification under the conditions that apply 
during a plant outage, and to avoid the potential of an unplanned 
transient if the surveillance were conducted with the plant at power.
    In addition to the substantive changes, there are format changes 
which are merely editorial and because format changes produce no 
physical change they do not influence the probability of new or 
different types of accidents.
    The proposed changes do not alter the intent or method by which the 
surveillance are conducted, do not involve any physical changes to the 
plant, do not alter the way any structure, system, or component 
functions, and do not modify the manner in which the plant is operated. 
As such, the proposed changes cannot create the possibility of a new or 
different kind of accident from any previously evaluated.
    3. The changes do not involve a significant reduction in a margin 
of safety.
    The proposed changes to surveillance requirements of the Haddam 
Neck Plant Technical Specifications extend the frequency for verifying 
the operability of the components/equipment. The proposal would extend 
the frequency from at least once per 18-months to at least once each 
refueling interval (24-months).
    In addition to the substantive changes, there are format changes 
which are merely editorial and because format changes produce no 
physical change they do not influence the margin of safety.
    The proposed changes to surveillance frequency are still consistent 
with the basis for the frequency, and the intent or method of 
performing the surveillance is unchanged. Further, the current 
inservice testing requirements and the previous history of reliability 
of the system provides assurance that the changes will not affect the 
reliability of the auxiliary feedwater system. Thus, it is concluded 
that there is no impact on the margin of safety.
    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 and Northeast Nuclear Energy Company, et al., 
Docket Nos. 50-245, 50-336, and 50-423, Millstone Nuclear Power 
Station, Units 1, 2, and 3, New London County, Connecticut

    Date of amendment request: June 6, 1995 (published August 2, 1995, 
60 FR 39434), as supplemented November 22, 1995.
    Description of amendment request: The proposed amendments will 
modify the size of the Plant Operations Review Committee (PORC) which 
will collectively have the experience and expertise in various areas of 
plant operation, and will clarify the composition of the Site 
Operations Review Committee (SORC).
    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:

    . . . These proposed changes do not involve an SHC because the 
changes do not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The plant operations review committee (PORC) is an oversight 
group and helps to ensure that the units are operated in a safe 
manner. To accomplish this the PORCs provide their recommendations 
on the safety related activities to the Vice President--Haddam Neck 
Plant for Haddam Neck and to the respective Nuclear Unit Directors 
for Millstone. Each Millstone Unit has its own PORC. It is proposed 
that the members of the Millstone PORCs be selected by the 
respective Nuclear Unit Director based on their knowledge and 
expertise in specific key plant functions. The Millstone Station has 
one site operations review committee (SORC). The SORC is also an 
oversight group whose charter is to advise the Senior Vice 
President--Millstone Station on all matters related to nuclear 
safety at the Millstone site. The Haddam Neck Plant, being a single 
unit site, has one PORC, which advises the Vice President--Haddam 
Neck Plant. The members of the Haddam Neck Plant PORC will be 
selected by the Vice President--Haddam Neck Plant based on their 
knowledge and expertise in specific key 

[[Page 7550]]
plant functions. The PORC and SORC add to the defense-in-depth concept 
provided by the design, operation, maintenance, and quality 
oversight by promoting excellence through the conduct of their 
affairs and by maintaining a diligent watch over their 
responsibilities.
    These administrative changes will revise the composition section 
of the technical specifications for the PORC members. Millstone Unit 
individuals will be appointed by the Nuclear Unit Directors if the 
individual meets one or more of the following areas of expertise: 
Plant Operations, Engineering, Reactor Engineering, Maintenance, 
Instrumentation and Controls, Health Physics, Chemistry, Work 
Planning and Control, and Quality Services. The Haddam Neck Plant, 
due to its broader scope of review also include an individual 
experienced in Security and specific expertise in Electrical 
Maintenance and Mechanical Maintenance. The individuals who will 
serve on PORC shall continue to meet the criteria of ANSI N18.1-1971 
along with the qualification requirements contained in the technical 
specifications. This approach is consistent with the standard 
technical specifications and NUREG 0800, Section 13.4. For SORC at 
the Millstone Station, the method of identifying who shall serve as 
Vice Chairperson has been modified for clarity. Finally, the 
individual who shall represent Quality and Assessment Services shall 
be modified to allow a qualified member of Quality and Assessment 
Services to serve on SORC.
    The remaining portions of the technical specifications related 
to PORC and SORC are not being revised.
    These modifications broaden the unit committee participation and 
reflect current organizational positions and will not increase the 
probability of occurrence or the consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed administrative enhancements to the composition of 
the PORC and Millstone Station SORC will not affect the way in which 
the units are physically operated. These administrative changes to 
PORC and SORC continue to meet the guidelines of ANSI N18.7-1976. 
The modifications to PORC and SORC continue to allow these groups to 
provide a thorough review of activities at the units.
    The proposed modification does not impact any initiating events, 
and therefore, cannot create the possibility of any new or different 
kind of accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    These proposed administrative changes will not impact the margin 
of safety provided by PORC and SORC. The PORC and SORC will continue 
to be staffed by qualified individuals experienced in the operation 
of the plants. These administrative changes will modify how the 
composition of the PORC and SORC members are presented in the 
technical specifications, but will not adversely impact their 
ability to review and comment on operations at the units.

    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, Connecticut 06457, for the Haddam Neck Plant, and 
the Learning Resources Center, Three Rivers Community-Technical 
College, 574 New London Turnpike, Norwich, CT 06360, for Millstone 1, 
2, and 3.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee.

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

    Date of amendment request: November 22, 1995 (NRC-95-0124).
    Description of amendment request: The proposed amendment would 
modify the allowed out-of-service time for one onsite alternating 
current (ac) electrical power division from 72 hours to 7 days. The 
proposed amendment would also eliminate accelerated testing and special 
reports as a result of diesel generator surveillance failures in 
accordance with Generic Letter 94-01, ``Removal of Accelerated Testing 
and Special Reporting Requirements for Emergency Diesel Generators,'' 
dated May 31, 1994.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident. Changing the out-of-
service time, surveillance frequency and reporting requirements for 
emergency diesel generators (EDGs) will not affect the initiation of 
an accident, since EDGs are not associated with any accident 
initiation mechanism. The proposed changes will not impact the plant 
design or method of EDG operation. The increased out-of-service time 
has been evaluated to have only a small impact on plant risk. 
Performing the EDG inspections during plant operations will decrease 
plant risk during plant outages. Deleting the accelerated testing 
provisions will not affect the consequences of an accident since the 
implementation of a maintenance and monitoring program for EDGs 
consistent with the provisions of the maintenance rule will assure 
EDG performance as discussed in Generic Letter 94-01. Deleting 
reporting requirements has no impact on consequences of an accident 
since reporting has no accident effect. Based on the amount of 
electrical system redundancy, the small increase in plant risk 
during operations and the decrease in plant risk during outages, 
this change will not result in a significant increase in the 
probability or consequences of an accident.
    2. The proposed changes do not create the possibility of a new 
or different accident from any previously evaluated. The proposed 
changes do not modify the plant design or method of diesel 
operation. Therefore, no new accident initiator is introduced, nor 
is a new type of failure created. For these reasons, no new or 
different type of accident is created by these changes.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety. Since implementation of a maintenance program 
for the EDGs consistent with the Maintenance Rule will ensure that 
high EDG performance standards are maintained, the accelerated 
testing schedule is not needed to maintain the margin of safety. 
Deleting reporting requirements has no impact on safety or margin of 
safety. Increasing the allowed out-of-service time for one division 
of onsite AC power will slightly increase EDG unavailability during 
plant operation. However, this change does not impact the redundancy 
of offsite power supplies, the allowed out-of-service time if both 
divisions are inoperable, or the ability to cope with a station 
blackout event. This request also does not change the Action 
statement for AC electrical power systems required when the plant is 
shutdown. The increase in core damage frequency was assessed to be 
small by an evaluation using the plant PSA [probabilistic safety 
assessment] for the operating condition. Enabling the diesel 
generator inspections to be performed on-line will improve safety 
while shutdown by reducing EDG out-of-service time during outages. 
For these reasons, the proposed changes do not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: John N. Hannon. 
    
[[Page 7551]]


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

    Date of amendment request: December 21, 1995 (NRC-95-0133).
    Description of amendment request: The proposed amendment would 
implement Option B of the recently revised 10 CFR Part 50 Appendix J in 
a manner consistent with Regulatory Guide 1.163, ``Performance-Based 
Containment Leak Test Program,'' and industry guidance contained in NEI 
94-01, Revision 0, ``Industry Guideline for Implementing Performance-
Based Option of 10 CFR 50, Appendix J,'' with the exception of 
previously approved exemptions which the licensee wishes to remain in 
effect. The previously approved exemptions are for reduced pressure for 
testing MSIVs [main steam isolation valves] and testing of LPCI [low 
pressure coolant injection] isolation valves in accordance with 
Technical Specification (TS) 4.4.3.2.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. This request does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change implements the new Option B of 10 CFR Part 
50 Appendix J on performance-based containment leakage testing. The 
proposed change does not involve a change to the plant design or 
operation. As a result, the proposed change does not affect any 
parameters or conditions that contribute to the initiation of any 
accidents previously evaluated. Thus, the proposed change cannot 
increase the probability of any accident previously evaluated.
    The proposed change potentially affects the leak-tight integrity 
of the containment structure designed to mitigate the consequences 
of a loss-of-coolant accident (LOCA). The function of the 
containment is to maintain functional integrity during and following 
the peak transient pressures and temperatures which result from any 
loss-of-coolant accident (LOCA). The containment is designed to 
limit fission product leakage following the design basis LOCA. 
Because the proposed change does not alter the plant design, only 
the frequency of measuring Type A, B, and C leakage, the proposed 
change does not directly result in an increase in containment 
leakage. However, decreasing the test frequency can increase the 
probability that an increase in containment leakage could go 
undetected for an extended period of time. Test intervals will be 
established based on the performance history of components being 
tested. The risk resulting from the proposed changes is 
characterized as follows, based primarily on the results contained 
in NUREG-1493 [''Performance-Based Containment Leakage Test 
Program''], the principal Technical Support Document used by the NRC 
as the basis for the Appendix J final rule (Reference 9 [of 
application]) and the NRC's Final Regulatory Impact Analysis as 
contained in SECY-95-181 [Final Regulatory Impact Analysis, 
Performance-Based Containment Leakage-Test Program (Attachment 2 to 
NRC Rulemaking Issue Affirmation, SECY-95-181 dated July 17, 1995, 
Final Amendment to 10 CFR 50, Appendix J, ``Containment Leakage 
Testing,'' to Adopt Performance-Oriented and Risk-Based Approaches)] 
(Reference 10 [of application]):

Type A Testing

    NUREG-1493 found that the effect of containment leakage on 
overall accident risk is minimal since risk is dominated by accident 
sequences that result in failure or bypass of the containment.
    Industry wide, ILRTs [integrated leak rate tests] have only 
found a small fraction of the leaks that exceed current acceptance 
criteria. Only three percent of all leaks are detectable only by 
ILRTs, and therefore, by extending the Type A testing intervals, 
only three percent of all leaks have a potential for remaining 
undetected for longer periods of time. In addition, when leakage has 
been detected by ILRTs, the leakage rate has been only marginally 
above existing requirements. The Fermi Type A testing confirms the 
industry-wide experience that a majority of the leakage experienced 
during Type A testing is through components tested by Type B and C 
tests.
    NUREG-1493 found that these observations, together with the 
insensitivity of reactor accident risk to the containment leakage 
rate, show that increasing the Type A leakage test intervals would 
have a minimal impact on public risk.

Type B and C Testing

    NUREG-1493 found that while Type B and C tests can identify the 
vast majority (greater than 95 percent) of all potential leakage 
paths, performance-based alternatives to current local leakage-
testing requirements are feasible without significant risk impacts. 
The risk model used in NUREG-1493 suggests that the number of 
components tested would be reduced by about 60 percent with less 
than a three-fold increase in the incremental risk due to 
containment leakage. Since, under existing requirements, leakage 
contributes less than 0.1 percent of overall accident risk, the 
overall impact is very small. In addition, the NRC's Final 
Regulatory Impact Analysis concluded that while the extended testing 
intervals for Type B and C tests led to minor increases in potential 
offsite dose consequences, the beneficial expected decrease in 
onsite (LLRT [local leak rate testing] & ILRT worker) dose exceeds 
(by at least an order of magnitude) the potential off-site dose 
consequences.
    The editorial change to the bases has no impact on the 
probability or consequence of an accident since it is strictly a 
correction to achieve consistency between the bases and the 
specifications.
    Based on the above, DECO [the licensee] has concluded that the 
proposed change will not result in a significant increase in the 
probability or consequences of any accident previously evaluated.
    2. The request does not create the possibility of occurrence of 
a new or different kind of accident from any accident previously 
evaluated.
    The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. This change involves the reduction of 
Type A, B, and C test frequency. Except for the method of defining 
the test frequency, the methods for performing the actual tests are 
not changed. No new accident modes are created by extending the 
testing intervals. No safety-related equipment or safety functions 
are altered as a result of this change. Extending the test frequency 
has no influence on, nor does it contribute to, the possibility of a 
new or different kind of accident or malfunction from those 
previously analyzed.
    The editorial change to the bases has no effect on any kind of 
accident since it is strictly a correction to achieve consistency 
between the bases and the specifications.
    Based on the above, DECO has concluded that the proposed change 
will not create the possibility [of] a new or different kind of 
accident previously evaluated.
    3. The request does not involve a significant reduction in a 
margin to safety.
    The proposed change only affects the frequency of Type A, B, and 
C testing. Except for the method of defining the test frequency, the 
methods for performing the actual tests are not changed. However, 
the proposed change can increase the probability that an increase in 
leakage could go undetected for an extended period of time. NUREG-
1493 has determined that, under several different accident 
scenarios, the increased risk of radioactivity release from 
containment is negligible with the implementation of these proposed 
changes.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to containment 
leakage rate. The containment isolation system is designed to limit 
leakage to La, which is defined by the Fermi 2 Technical 
Specifications to be 0.5 percent by weight of the containment air 
per 24 hours at 56.5 psig (Pa). The limitation on containment 
leakage rate is designed to ensure that total leakage volume will 
not exceed the value assumed in the accident analyses at the peak 
accident pressure (Pa). The margin to safety for the offsite 
dose consequences of postulated accidents directly related to the 
containment leakage rate is maintained by meeting the 1.0 La 
acceptance criteria. The La value is not being modified by this 
proposed Technical Specification change.
    Except for the method of defining the test frequency, no change 
in the method of testing is being proposed. The Type B and C tests 
will continue to be done at full pressure (Pa) or greater with 
the exception of the Main Steam Isolation Valves, which have an 
approved exemption. Other programs are in 

[[Page 7552]]
place to ensure that proper maintenance and repairs are performed 
during the service life of the primary containment and systems and 
components penetrating the primary containment.
    The editorial change to the bases has no effect on the margin of 
safety since it is strictly an editorial change to achieve 
consistency between the bases and the specifications.
    As a result, DECO has concluded that the proposed change will 
not result in a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.
    Attorney for licensee: John Flynn, Esq., Detroit Edison Company, 
2000 Second Avenue, Detroit, Michigan 48226.
    NRC Project Director: John N. Hannon.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, Docket 
No. 50-498, South Texas Project, Unit 1, Matagorda County, Texas

    Date of amendment request: January 22, 1996.
    Description of amendment request: The proposed amendment would 
modify the steam generator tube plugging criteria in Technical 
Specification 3/4.4.5, Steam Generators, and the allowable leakage in 
Technical Specification 3/4.4.6.2, Operational Leakage, and the 
associated Bases. The amendment would allow the implementation of 
alternate steam generator tube plugging criteria for the tube support 
plate (TSP)/tube intersections for Unit 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?

Structural Considerations

    Industry testing of model boiler and operating plant tube 
specimens for free span tubing at room temperature conditions show 
typical burst pressures in excess of 5000 psi for indications of 
outer diameter stress corrosion cracking with voltage measurements 
at or below the structural limit of 4.0 volts. One model boiler 
specimen with a voltage amplitude of 19 volts also exhibited a burst 
pressure greater than 5000 psi. Burst testing performed on one 
intersection pulled from STP Unit 1 in 1993 with a 0.51 volt 
indication yielded a measured burst pressure of 8900 psi at room 
temperature. Burst testing performed on another intersection pulled 
from STP Unit 1 in 1995 with a 0.48 volt indication yielded a 
measured burst pressure of 9950 psi at room temperature.
    The projected end-of-cycle (EOC) voltage compares favorably with 
the 4.7 volt structural limit considering the EPRI [Electric Power 
Research Institute] voltage growth rate for indications at STP. 
Using the methodology of the NRC Generic Letter 95-05, the 
structural limit is reduced by allowances for uncertainty and growth 
to develop a beginning-of-cycle (BOC) repair limit which should 
preclude EOC indications from growing in excess of the structural 
limit. The non-destructive examination (NDE) uncertainty to be 
applied per EPRI is approximately 20 percent. The EPRI recommended 
growth allowance of 30 percent/EFPY [effective full power year] is 
also to be applied. This growth value is conservative for STP Unit 1 
based on previous inspection history. By adding NDE uncertainty 
allowances and a crack growth allowance to the repair limit, the 
structural limit can be validated. Therefore, the maximum allowable 
BOC repair limit (RL) based on the structural limit of 4.7 volts can 
be represented as:

RL + (0.20 x RL) + (0.45* x RL) = 4.7 volts, which yields RL of 2.85 
volts.

    * The 30% growth rate for 1 EFPY was scaled up to the cycle 
length used at South Texas.

    This repair limit (2.85 volts) reasonably could be applied for 
APC [alternate plugging criteria] implementation to repair bobbin 
indications greater than the 1.0 volt criterion specified by NRC 
Generic Letter 95-05 and is independent of RPC [rotating pancake 
coil-probe] confirmation of the indications. STP has chosen to use a 
steam generator tube upper repair limit of 2.85 volts to assess tube 
integrity for those bobbin indications which are above 1.0 volt but 
do not have confirming RPC calls. This 2.85 volt upper limit for 
non-confirmed RPC calls is consistent with the NRC Generic Letter 
95-05. Since the upper bound for repair of non-confirmed RPC is 
limited to a value far less than the structural limit associated 
with a full alternate criteria, the establishment of the repair 
limits are determined to be reasonable and conservative with respect 
to the industry pulled tube data base used.

Leakage Considerations

    As part of the implementation of APC, the distribution of EOC 
cracking indications at the TSP intersections has been used to 
calculate the primary-to-secondary leakage which is bounded by the 
maximum leakage required to remain within applicable dose limits. 
This limit was calculated using the Technical Specification RCS 
[reactor coolant system] Iodine-131 transient spiking values 
consistent with NUREG-0800. Application of the APC criteria requires 
the projection of postulated MSLB [main steam line break] leakage 
based on the projected EOC voltage distribution for the beginning of 
cycle. Projected EOC voltage distribution is developed using the 
most recent EOC eddy current results and a voltage measurement 
uncertainty. Draft NUREG-1477 requires that all indications to which 
APC is applied must be included in the leakage projection.
    The projected MSLB leakage rate calculation methodology 
prescribed in EPRI TR-100407 will be used to calculate the EOC 
leakage. A Monte Carlo approach will be used to determine the EOC 
leakage, accounting for all of the ECT [eddy current testing] 
uncertainties, voltage growth, and an assumed probability of 
detection (POD) of 0.6 for a 1.0 volt repair limit. The fitted 
logarithmic function probability of leakage correlation will be used 
to establish the STP MSLB leak rate used for comparison with a 
bounding allowable leak rate in the faulted loop which would result 
in radiological consequences which are within applicable dose 
limits. Due to the relatively low voltage levels of indications at 
STP and low voltage growth rates, it is expected that the actual 
calculated leakage values will be far less than this limit.
    Therefore, implementation of APC does not adversely affect steam 
generator tube integrity and implementation will be shown to result 
in acceptable dose consequences. The proposed amendment does not 
result in any increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    Implementation of the proposed steam generator tube alternate 
plugging criteria for ODSCC [outer diameter stress corrosion 
cracking] at the TSP intersections does not introduce any 
significant changes to the plant design basis. Use of the criteria 
does not provide a mechanism which could result in an accident 
outside of the region of the TSP elevations since no ODSCC has been 
identified outside the thickness of the TSPs. It is therefore 
expected that for all plant conditions, neither a single or multiple 
tube rupture event would occur in a steam generator where APC has 
been applied.
    Specifically, STP will implement, for Unit 1, a maximum leakage 
rate of 150 gpd [gallons per day] per steam generator (SG) to help 
preclude the potential for excessive leakage during all plant 
conditions. The current technical specification limits on primary-
to-secondary leakage at operating conditions are 1 gpm [gallon per 
minute] for all steam generators or 500 gpd for any one SG. The RG 
[Regulatory Guide] 1.121 criterion for establishing operational 
leakage rate limits governing plant shutdown is based upon leak-
before-break (LBB) considerations to detect a free span crack before 
potential tube rupture as a result of faulted plant conditions. The 
150 gpd limit is intended to provide for leakage detection and plant 
shutdown in the event of an unexpected crack propagation resulting 
in excessive leakage. RG 1.121 acceptance criteria for establishing 
operating leakage limits are 

[[Page 7553]]
based on LBB considerations such that plant shutdown is initiated if 
the permissible crack is exceeded.
    The predicted EOC leakage for STP is based on the calculated 
growth rate and does not take credit for the TSP proximity during 
normal operation. Thus, the 150 gpd limit provides for plant 
shutdown prior to reaching critical crack lengths. Additionally, 
this leak-before-break evaluation assumes that the entire crevice 
area is uncovered during the secondary side blowdown of a MSLB. 
Typically, it is expected for the vast majority of intersections 
that only partial uncovery will occur. Thus, the proximity of the 
TSP will enhance the burst capacity of the tube.
    Steam generator tube integrity is continually maintained through 
inservice inspection and primary-to-secondary leakage monitoring. 
Any tubes falling outside the APC repair limits are removed from 
service. Therefore, the possibility of a new or different kind of 
accident from any accident previously developed is not created.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The use of the voltage based bobbin probe for dispositioning 
ODSCC degraded tubes within TSP intersections by APC is demonstrated 
to maintain steam generator tube integrity in accordance with the 
requirements of RG 1.121. RG 1.121 describes a method acceptable to 
the NRC staff for meeting GDCs [General Design Criterion] 14, 15, 
31, and 32 by reducing the probability or the consequences of steam 
generator tube rupture. This is accomplished by determining the 
limiting conditions of degradation of steam generator tubing, as 
established by inservice inspection, for which tubes with 
unacceptable cracking are removed from service. Upon implementation 
of the criteria, even under the worst case conditions, the 
occurrence of ODSCC at the TSP elevation is not expected to lead to 
a steam generator tube rupture event during normal or faulted plant 
conditions. The EOC distribution of crack indications at the TSP 
elevations will be confirmed to result in acceptable primary-to-
secondary leakage during all plant conditions and that radiological 
consequences are not adversely impacted.
    In addressing the combined effects of loss of coolant accident 
(LOCA) and safe shutdown earthquake (SSE) on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case at STP as the TSP may become deformed as a result of 
lateral loads at the wedge supports at the periphery of the plate 
due to the combined effects of the LOCA rarefaction wave and SSE 
loadings. The resulting secondary-to-primary pressure differential 
on the deformed tubes may cause some of the tube to collapse.
    There are two concerns associated with steam generator tube 
collapse. First, the collapse of steam generator tubing reduces the 
RCS flow area through the tubes. The reduction in flow area 
increases the resistance to flow of steam from the core during a 
LOCA which, in turn, may potentially increase peak clad temperature 
(PCT). Second, there is a potential that through wall cracks in 
tubes could sufficiently enlarge during tube deformation or 
collapse, causing sufficient in-leakage of secondary water back to 
the core which dilutes the poisoning effect of boron injection from 
the emergency cooling system. Again, an increase in core PCT may 
result.
    Consequently, since the LBB methodology is applicable to the STP 
reactor coolant loop piping, the probability of breaks in the 
primary loop piping is sufficiently low that they need not be 
considered in the structural design of the plant. The analysis 
identified tubes located adjacent to wedge regions that are subject 
to potential collapse during combined LOCA and SSE. These tubes will 
be excluded from application of APC. Thus, existing tube integrity 
requirements apply to these tubes and the margin of safety is not 
reduced.
    Implementation practices using the bobbin probe voltage based 
tube plugging criteria bounds RG 1.83 considerations by:
    (1) Using enhanced eddy current inspection guidelines consistent 
with those used by EPRI in developing the correlations. This 
provides consistency in voltage normalization,
    (2) Performing a 100 percent bobbin coil inspection for all hot 
leg tube support plate intersections and all cold leg intersections 
down to the lowest cold leg tube support plate with outer diameter 
stress corrosion cracking (ODSCC) indications. The determination of 
the tube support plate intersections having ODSCC indications shall 
be based on the performance of at least a 20% random sampling of 
tubes inspected over their full length, and
    (3) Incorporating RPC inspection for all tubes with larger 
indications left in service. This further establishes the principal 
degradation morphology as ODSCC.
    Implementation of APC at TSP intersections will decrease the 
number of tubes which must be repaired. Since the installation of 
tube plugs (to remove ODSCC degraded tubes from service) reduces the 
RCS flow margin, APC implementation will help preserve the margin of 
flow that would otherwise be reduced.
    For each cycle the projected EOC primary-to-secondary leak rate 
allowed is bounded by a leak rate which limits the radiological 
consequences of a EOC MSLB to within applicable dose limits. 
Therefore, this change does not involve a significant reduction in 
the margin to safety.
    It is therefore concluded that the proposed license amendment 
request does not result in a significant reduction in the margin of 
safety as defined in the plant Final Safety Analysis Report or 
Technical Specifications.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869.
    NRC Project Director: William D. Beckner.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, Docket 
No. 50-498, South Texas Project, Unit 1, Matagorda County, Texas

    Date of amendment request: January 22, 1996.
    Description of amendment request: The proposed amendment would 
modify the steam generator tube plugging criteria in Technical 
Specification 3/4.4.5, Steam Generators, and the associated Bases, to 
allow the implementation of alternate steam generator tube plugging 
criteria for the tube-to-tubesheet joints (known in the industry as F*) 
for Unit 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes to the Steam Generator section of Technical 
Specifications do not affect any accident initiators or precursors 
and do not alter the design assumptions for the systems or 
components used to mitigate the consequences of an accident. The 
requirements approved by the NRC will not be reduced by this 
request. Since F* utilizes the ``as rolled'' tube configuration that 
exists as part of the original steam generator design, all of the 
design and operating characteristics of the steam generator and 
connected systems are preserved. The F* joint has been analyzed and 
tested for design, operating and faulted condition loadings in 
accordance with Regulatory Guide 1.121 safety factors. At worst 
case, a tube leak would occur with the result being a primary to 
secondary leak.
    Should a tube leak occur, the impact is bounded by the ruptured 
tube evaluation submitted by STP for the Unit 1 operating license. 
No new or unreviewed accident conditions are created by the use of 
F* criteria. The potential for a tube rupture is not increased from 
the original submittal, thus there is no impact on accidents 
evaluated as the design basis. Therefore use of the F* criteria will 
not increase the probability of occurrence 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. 

[[Page 7554]]

    The use of the proposed F* alternate plugging criteria will not 
introduce significant or adverse changes to the plant design basis. 
The failure of a tube which remained unplugged in accordance with 
the F* criteria would result in a tube leak, which is a previously 
analyzed condition. Since this leak would occur below the secondary 
face of the tubesheet, its leak rate would be limited by the tube-
to-tubesheet interface. Qualification testing and previous 
experience indicates that normal and faulted leakage would be well 
below the technical specification limits creating no threat 
associated with tube rupture type leakages. This conclusion is 
consistent with previous F* programs approved and used at other 
operating plants.
    However, in the unlikely event the failed tube severed 
completely at a point below the F* region, the remaining F* joint 
would retain engagement in the tubesheet due to its length of 
expanded contact within the tubesheet bore, preventing any 
interaction with neighboring tubes. If the tube severs at a point 
above the F* region, then it is covered by the tube rupture event as 
a part of the UFSAR [Updated Final Safety Analysis Report]. Thus, 
the possibility of a new or different type of accident from any 
accident previously evaluated is not created.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Based on previous responses (above), the protective boundaries 
of the steam generator are preserved. A tube with degradation can be 
kept in service through F* criteria which provided an un-degraded 
expanded interface with the tubesheet and which satisfies all of the 
necessary structural and leakage requirements in accordance with 
Regulatory Guide 1.121 and the Technical Specifications. Since the 
joint is constrained within the tubesheet bore there is no 
additional risk associated with tube rupture. Since the UFSAR 
analyzed accident scenarios remain bounding, the use of an F* 
criteria does not reduce the margin of safety.
    Thus, these changes do not involve a significant reduction in 
the margin of safety. Therefore, based on the above evaluation, STP 
has concluded that these changes do not involve any significant 
hazards considerations.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488.
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner.

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

    Date of amendment requests: January 12, 1996 (AEP:NRC:1233).
    Description of amendment requests: The proposed amendments would 
modify technical specification section 4.4.11 to eliminate the 
surveillance requirement (SR) demonstrating operability of the 
emergency power supply for the pressurizer power-operated relief valves 
(PORVs) and block valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    Per 10 CFR 50.92, a proposed change does not involve significant 
hazards consideration if the change does not:
    1. involve a significant increase in the probability or 
consequence of an accident previously evaluated,
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated, or
    3. involve a significant reduction in a margin of safety.

Criterion 1

    The proposed change is consistent with NUREG-1431 [Standard 
Technical Specifications Westinghouse Plants]. Due to the high 
reliability and continued testing of the Class 1E power supply, we 
conclude that the elimination of the SR will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.

Criterion 2

    The proposed change does not involve the addition of any new 
plant operation or procedures, and the elimination of the SR is 
consistent with NUREG-1431. For these reasons, we believe that the 
proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

Criterion 3

    The proposed change is consistent with NUREG-1431, and it does 
not affect the acceptance criteria of any of the other PORV and 
block valve tests currently performed. For these reasons, we believe 
that the proposed amendment will not involve a significant reduction 
in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and the 
applicable Bases of the Standard Technical Specifications Westinghouse 
Plants. The Bases for the applicable surveillance, 3.4.11.4, states 
``This Surveillance is not required for plants with permanent 1E power 
supplies to the valves.'' Based on this review, it appears that the 
three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC 
staff proposes to determine that the amendment requests involve no 
significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037.
    NRC Project Director: John N. Hannon.

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

    Date of amendment request: January 22, 1996.
    Description of amendment request: The proposed change relocates the 
containment isolation valve (CIV) list, Table 3.6-2, from the Technical 
Specifications to the Technical Requirements Manual (TRM). This change 
affects Technical Specifications Sections 1.8.1a, 4.6.1.1a, 3.6.3.1, 
4.6.3.1.1 and 4.6.3.1.2, and the Basis Section 3/4.6.3. A note at the 
bottom of Table 3.6-2 regarding the CIVs that are subject to 
administrative control is retained in the Technical Specifications by 
relocating it to Sections 1.8.1a and 4.6.1.1a. This change is being 
performed in accordance with Generic Letter 91-08, which provides 
guidance for removal of component lists from the Technical 
Specifications.
    Additionally, a change to provide relief in the surveillance 
requirement in Section 4.6.1.1a is included. The change allows valves, 
blind flanges, and deactivated automatic valves located inside the 
containment and are locked, sealed, or otherwise secured in the closed 
position to be verified closed during each cold shutdown but not more 
often than once per 92 days. The current requirements check the valve 
position once per 31 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration (SHC), which is presented below:

    Pursuant to 10CFR50.92, Northeast Nuclear Energy Company (NNECO) 
has reviewed the proposed changes. NNECO concludes that these 
changes do not involve a significant hazards consideration (SHC) 
since the proposed changes satisfy the criteria in 10CFR50.92(c). 
That is, the proposed changes do not: 

[[Page 7555]]

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change to remove the Containment Isolation Valve 
(CIV) list from the Technical Specifications will not result in any 
hardware or operating changes. The proposed change is based upon NRC 
Generic Letter (GL) 91-08 and merely removes the CIV table and all 
references to the table from the technical specifications without 
affecting the operability requirements of any of the listed valves. 
The technical specifications will continue to require the CIVs to be 
operable. Limiting Condition for Operation and surveillance 
requirements for the valves will also remain in the technical 
specifications. The CIV table will be relocated to the Millstone 
Unit No. 2 Technical Requirements Manual (TRM) which is controlled 
in accordance with 10CFR50.59.
    This change is administrative in nature and does not involve an 
increase in the probability or consequence of an accident previously 
evaluated. Furthermore, the proposed change does not alter the 
design, function, or operation of the valves involved, and therefore 
does not affect the probability or consequences of any previously 
evaluated accident.
    The change to Section 4.6.1.1a that reduces the surveillance 
requirement for valves, blind flanges, and deactivated automatic 
valves located inside the containment provides consistency with 
NUREG-1432, ``Standard Technical Specifications for Combustion 
Engineering Plants'' as well as the Technical Specifications of 
Millstone Unit No. 3, Haddam Neck Plant, and Seabrook. The 
probability or consequences of any previously evaluated accidents 
are not affected.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The change to relocate the CIV list from the technical 
specifications to the TRM will not impose any different operational 
or surveillance requirements, nor will the change remove any such 
requirements. Adequate control of information will be maintained. 
Furthermore, as stated above, the proposed change does not alter the 
design, function, or operation of the valves involved, and therefore 
no new accident scenarios are created.
    The change to Section 4.6.1.1a that reduces the surveillance 
requirement for valves, blind flanges, and deactivated automatic 
valves located inside the containment does not alter the design, 
function, or operation of the valves involved, and therefore no new 
accident scenarios are created.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes will not reduce the margin of safety since 
it has no impact on any safety analysis assumption. The proposed 
changes do not decrease the scope of equipment currently required to 
be operable or subject to surveillance testing, nor does the 
proposed change affect any instrument setpoints or equipment safety 
functions.
    The relocation of the valve list is consistent with the guidance 
provided in GL 91-08. The change to the surveillance interval is 
consistent with NUREG-1432, ``Standard Technical Specifications for 
Combustion Engineering Plants'' as well as the Technical 
Specifications of Millstone Unit No. 3, Haddam Neck Plant, and 
Seabrook. The intent of the technical specification will be met 
since the change will not alter function or operability requirements 
for any CIV.

    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: January 17, 1996.
    Description of amendment request: The amendment request would 
delete a license requirement to submit responses to and to implement 
requirements of Generic Letter 83-28, because the requirement has been 
completed. Generic Letter 83-28 pertains to the Salem anticipated 
transient without scram (ATWS) event.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    . . . The proposed change does not involve an SHC because the 
change would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    NNECO's proposal to delete License Condition 2.C(4) is an 
administrative change. The NRC Staff has accepted Millstone Unit No. 
3's responses regarding the actions required by GL 83-28, thus, the 
license condition has been met and is no longer necessary. The 
proposed change does not affect the configuration, operation, or 
performance of any system, structure, or component. Additionally, 
the limiting conditions for operation, limiting safety system 
settings, and safety limits specified in the Millstone Unit No. 3 
Technical Specifications are unchanged. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The NRC Staff has accepted Millstone Unit No. 3's responses 
regarding the actions required by GL 83-28, thus, the license 
condition has been met and is no longer necessary. The proposed 
change to delete License Condition 2.C(4) does not affect the 
configuration, operation, or performance of any system, structure, 
or component. Additionally, the limiting conditions for operation, 
limiting safety system settings, and safety limits specified in the 
Millstone Unit No. 3 Technical Specifications are unchanged. 
Therefore, this proposed change cannot create the possibility of a 
new or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The NRC Staff has accepted Millstone Unit No. 3's responses 
regarding the actions required by GL 83-28, thus, the license 
condition has been met and is no longer necessary. The proposed 
change to delete License Condition 2.C(4) does not affect the 
configuration, operation, or performance of any system, structure, 
or component. Additionally, the limiting conditions for operation, 
limiting safety system settings, and safety limits specified in the 
Millstone Unit No. 3 Technical Specifications are unchanged. 
Therefore, this 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: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360.

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

    Date of amendment request: December 22, 1995.
    Description of amendment request: The proposed changes will revise 
Limerick Generating Station, Units 1 and 2, Technical Specification 
3.6.1.8 ``Drywell and Suppression Chamber Purge System,'' increasing 
the Drywell and Suppression Chamber Purge System operating time limit 
from 90 hours each 365 days to 180 hours each 365 days.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the 

[[Page 7556]]
issue of no significant hazards consideration, which is presented 
below:

    1. The proposed Technical Specification [TS] changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    These TS changes do not increase the probability of occurrence 
of an accident previously evaluated in the SAR [Safety Analysis 
Report]. This activity involves changing the allowable operating 
limit for the Drywell and Suppression Chamber Purge System from 90 
hours each 365 days to 180 hours each 365 days. This change 
increases the probability that this system will be in service should 
a LOCA [loss of coolant accident] occur, but does not increase the 
probability that a LOCA will occur.
    Increasing the operating limit for the Drywell and Suppression 
Chamber Purge System from 90 hours to 180 hours each 365 days does 
not increase the consequences of a LOCA as previously evaluated in 
the SAR. These proposed TS changes increase the probability of a 
LOCA occurring during the time the Drywell and Suppression Chamber 
Purge System is in operation, and therefore, increase the 
probability of the failure of the operating SGTS [Standby Gas 
Treatment System] filter bank. However, the risk to containment 
integrity was previously evaluated and found to be acceptable (UFSAR 
[Updated Final Safety Analysis Report] Section 9.4.5.1.2.2 and 
WASH--1400 ``Reactor Safety Study'').
    Increasing the duration that the vent/purge line isolation 
valves may be open does not increase the probability that these 
valves will not perform as designed (i.e., close upon receipt of an 
isolation signal) in response to a LOCA. However, the changes will 
increase the likelihood that the vent and purge valves will be 
called on to close. As discussed in UFSAR Section 6.2.4.2, the 
containment purge valves have undergone extensive testing and 
analyses to demonstrate the operability of these valves following a 
LOCA.
    In addition to the existing Safety Analysis Report (SAR) 
evaluations, a Level 2 PSA [Probabilistic Safety Assessment] 
Analysis (containment failure) was performed to determine the 
additional risk associated with changing the operating limit from 90 
to 180 hours each 365 days. The PSA evaluation conservatively 
assumed a 200 hour vent/purge duration per a 365 day period. The 
figure of merit evaluated is the large early release frequency 
(LERF) which represents the likelihood of containment failure 
following core damage that could significantly affect the public 
(e.g., release of a large amount of radioactive material early 
enough in the accident that evacuation of the public has not 
occurred). The 200 hour vent/purge duration increased the LERF 
approximately 3% from the base value of 2.57E-8 for all PSA 
initiators. This analysis concluded that the increase in risk of 
containment failure is well within the bounds of the EPRI 
[Electrical Power Research Institute] PSA Applications Guideline for 
permanent changes. The same relative increase applies to the large 
Design Basis Accident LOCA LERF.
    These changes do not directly or indirectly degrade the 
performance of any other safety systems (assumed to function in the 
accident analysis) below their design basis. The potential for other 
equipment failures in the reactor enclosure due to duct-work impact, 
impingement, and the resulting environmental conditions was 
evaluated. It was concluded that the environmental qualifications 
for the LGS equipment are sufficient to ensure operability under the 
predicted environmental conditions, and there is no impact or 
impingement-related damage to essential equipment. Although the 
probability of occurrence of a malfunction of equipment important to 
safety is increased, the existing SAR analysis and Level 2 PSA 
Analysis demonstrate the increased risk and radiological 
consequences are not significant.
    2. The proposed TS changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    This activity does not change the function of the Drywell and 
Suppression Chamber Purge System, the containment isolation system, 
or SGTS as previously evaluated in the SAR. Changing the duration of 
operation of the vent and purge system does not create an accident 
initiator not considered in the SAR. Therefore, the possibility of 
an accident of a different type is not created.
    This activity does not create a failure mode not considered in 
the SAR. All possible equipment failures that could occur as a 
result of a LOCA during high volume purging have previously been 
identified and evaluated in the SAR. Therefore, this activity does 
not create the possibility of a different type of malfunction of 
equipment important to safety.
    Therefore, the proposed TS changes will not create the 
possibility of a new or different type of accident from any accident 
previously evaluated.
    3. The proposed TS changes do not involve a significant 
reduction in a margin of safety.
    The Bases of Technical Specification 3.6.1.8 states that the 
intent of the 90 hour per 365 day operating limit for the Drywell 
and Suppression Chamber Purge System is to protect the integrity of 
the SGTS filters. As discussed above, the requirements specified in 
ODCM paragraph 3.3.6 assure the availability of the backup SGTS 
filter train during operation of the vent and purge system. 
Furthermore, as discussed above, revising the operating limit from 
90 hours to 180 hours each 365 days does not involve a significant 
increase in risk. The margin of safety as defined in the Bases of 
Technical Specification 3.6.1.8 is maintained.
    Therefore, the implementation of the proposed TS changes will 
not involve a significant reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V.P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101.
    NRC Project Director: John F. Stolz.

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

    Date of amendment request: February 6, 1996.
    Description of amendment request: The amendments would change the 
Technical Specifications to lower the 125 Volt Battery Charger 
surveillance amperage from at least 200 amps to at least 170 amps.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment will permit replacement of aging battery 
chargers while ensuring these replacement battery chargers will 
restore the battery from the design minimum charge to its fully 
charged state while supplying normal steady-state loads. This meets 
the design basis for the 125V DC system and is consistent with Salem 
Unit 1 and 2 commitment to IEEE 308-1971 in UFSAR Section 3A.
    The 125V DC battery chargers are not addressed as a contributor 
to any accident analyzed in the UFSAR, therefore, changes to the 
battery charger output current will not increase the probability of 
an accident occurring.
    The limiting analyzed accident considered in this proposed TS 
amendment is the Loss of Offsite Power coincident with a Loss of 
Coolant Accident. This is currently the limiting design duty cycle 
for the batteries. The 125V batteries are sized to maintain all 
emergency loads for a period of 2 hours without battery chargers. 
This is demonstrated by performing the surveillance specified in TS 
4.8.2.3.2.f, which is not being changed. Since the chargers are not 
required to be available during this 2 hour period, and since the 
proposed charging rate will supply the necessary loads following 
restoration of AC power, the proposed amendment will have no effect 
on the consequences of this accident.
    The current limiter is calculated to extend the recharging time 
from 20 hours to 30 hours, but this is not considered significant 
since two, sequential battery discharge events are not considered 
plausible.
    PSE&G calculation substantiates the capability of the chargers 
to restore the battery from the design minimum charge to its fully 
charged state while supplying 

[[Page 7557]]
normal steady-state loads following a Station Blackout (SBO) Event 
which exceeds the current design duty cycle.
    In addition, a review of 125V DC Battery System load profiles 
indicated that the battery chargers are capable of supplying 
expected loads when restoring the battery from a design minimum 
charge state to a fully charged state irrespective of the status of 
the plant.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Will not create the possibility of a new or different kind of 
accident from any previously evaluated.
    The proposed amendment does not result in any design or physical 
configuration changes to the 125V DC system. This change supports 
the installation of the replacement chargers and ensures the 
chargers are surveilled within the bounds of limiting input 
amperage. No changes are being made to the function, design basis, 
or operation of the 125V DC system by this proposed change. 
Therefore, the proposed amendment will not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety.
    The proposed amendment to TS 4.8.2.3.2.e ensures that the 
replacement battery chargers have sufficient capacity to restore 
each 125V battery from the design minimum charge to its fully 
charged state while supplying normal steady-state loads. A margin of 
safety is maintained on both the AC input and DC output of the 
chargers since the specified current is above that required to 
support the 125V DC system and will result in AC current below the 
ampacity rating of the battery charger input cables.
    Testing to a charger output current of at least 170 amps will 
maintain a margin of safety to the current required during actual 
worst case normal loading on the 125V DC buses.
    Therefore, the proposed amendment will not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502.
    NRC Project Director: John F. Stolz.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. Ginna 
Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 9, 1996.
    Description of amendment request: The proposed amendment would 
allow an installed overhead door assembly, to be used in lieu of the 
equipment hatch closure, to isolate the hatch opening to the 
containment building during fuel movement and core alterations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. Containment 
closure is used with respect to the mitigation of fuel handling 
accidents, and as such, any change to these requirements will not 
affect the probability of an accident. The proposed changes will 
also not result in a significant increase in the consequences of an 
accident previously analyzed since the technical specification 
requirements remain bounded by the fuel handling accident assumption 
of no containment closure.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. The proposed changes 
will not impose any new or different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. Containment closure is not assumed in the accident analyses 
for Ginna Station. Also, the proposed change remains acceptable with 
respect to SRP [NUREG-800, ``Standard Review Plan for the Review of 
Safety Analysis Reports for Nuclear Power Plants, July 1981''] 
15.7.4 and GDC [General Design Criterion] 19 requirements. 
Therefore, no question of safety is involved, and 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: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Project Director: Ledyard B. Marsh.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 9, 1996.
    Description of amendment request: The proposed amendment would 
incorporate the methodology for determining the Low Temperature 
Overpressure Protection (LTOP) limits into the Administrative Controls 
Section 5.6.6 of the Ginna Technical Specifications (TS). The proposed 
amendment will allow the licensee to perform future LTOP evaluations, 
using NRC-approved methodology, without requiring changes to 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:

    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
changes only require that future LTOP limits be developed using NRC 
approved methodology as specified within the Administrative Controls 
section and do not involve any technical changes. As such, these 
changes are administrative in nature and do not impact initiators or 
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 analyzed.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
changes do not involve a physical alteration of the plant (i.e., no 
new or different type of equipment will be installed) or changes in 
the methods governing normal plant operation. The proposed changes 
will not impose any new or different requirements. Thus, this change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The proposed changes will not reduce a margin of plant 
safety because 

[[Page 7558]]
the changes do not impact any safety analysis assumptions other than 
requiring future evaluations of LTOP limits to be performed in 
accordance with NRC approved methodology. These changes are 
administrative in nature. As such, no question of safety is 
involved, and 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: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Project Director: Ledyard B. Marsh.

Rochester Gas and Electric Corporation, Docket No. 50-244, R.E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of amendment request: February 9, 1996.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications setpoints for steam generator (SG) 
water level-high feedwater isolation function. It would take advantage 
of a greater allowable operating band for SG water level afforded by 
replacement SGs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant increase in the probability 
or consequences of an accident previously evaluated. The proposed 
setpoint change does not degrade the performance of any plant 
equipment. Therefore, the probability of an accident is not 
increased. Since the revised trip setpoint and allowable value 
remain bounded by the accident analysis value of 100% steam 
generator narrow range level, the consequences of any accident are 
not adversely affected.
    2. Operation of Ginna Station in accordance with the proposed 
changes does not create the possibility of a new or different kind 
of accident from any accident previously evaluated. The proposed 
change does not involve a physical alteration to the plant (i.e., no 
new or different types of equipment will be installed) or changes in 
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.
    3. Operation of Ginna Station in accordance with the proposed 
changes does not involve a significant reduction in a margin of 
safety. The revised setpoint and allowable value remain bounded by 
the accident analysis assumptions. The existing values are based on 
design considerations and not accident analysis parameters. The 
replacement steam generators are not restricted by the same design 
considerations with respect to the ESFAS [engineered safety features 
actuation system] Steam Generator Water Level--High function. 
Therefore, this change does not involve a reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.
    Attorney for licensee: Nicholas S. Reynolds, Winston & Strawn, 1400 
L Street, NW., Washington, DC 20005.
    NRC Project Director: Ledyard B. Marsh.

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

    Date of amendment request: February 9, 1996.
    Description of amendment request: The proposed amendment would 
change Technical Specification 5.3.1 to allow the use of Zirlo fuel 
cladding material.
    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 methodologies used in the accident analysis remain 
unchanged. The proposed changes do not change or alter the design 
assumptions for the systems or components used to mitigate the 
consequences of an accident. Use of ZIRLO fuel cladding does not 
adversely affect fuel performance or impact nuclear design 
methodology. Therefore accident analyses are not impacted.
    The operating limits will not be changed and the analysis 
methods to demonstrate operation within the limits will remain in 
accordance with NRC approved methodologies. Other than the changes 
to the fuel assemblies, there are no physical changes to the plant 
associated with this technical specification change. A safety 
analysis will continue to be performed for each cycle to demonstrate 
compliance with all fuel safety design bases.
    VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods meet the 
same fuel assembly and fuel rod design bases as other VANTAGE 5 fuel 
assemblies. In addition, the 10 CFR 50.46 criteria are applied to 
the ZIRLO clad rods. The use of these fuel assemblies will not 
result in a change to the reload design and safety analysis limits. 
Since the original design criteria are met, the ZIRLO clad fuel rods 
will not be an initiator for any new accident. The clad material is 
similar in chemical composition and has similar physical and 
mechanical properties as Zircaloy-4. Thus, the cladding integrity is 
maintained and the structural integrity of the fuel assembly is not 
affected. ZIRLO cladding improves corrosion performance and 
dimensional stability. No concerns have been identified with respect 
to the use of an assembly containing a combination of Zircaloy-4 and 
ZIRLO clad fuel rods. Since the dose predictions in the safety 
analyses are not sensitive to fuel rod cladding material, the 
radiological consequences of accidents previously evaluated in the 
safety analysis remain valid.
    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.
    VANTAGE 5 fuel assemblies with ZIRLO clad fuel rods satisfy the 
same design bases as those used for other VANTAGE 5 fuel assemblies. 
All design and performance criteria continue to be met and no new 
failure mechanisms have been identified. The ZIRLO cladding material 
offers improved corrosion resistance and structural integrity.
    The proposed changes do not affect the design or operation of 
any system or component in the plant. The safety functions of the 
related structures, systems or components are not changed in any 
manner, nor is the reliability of any structure, system or component 
reduced. The changes do not affect the manner by which the facility 
is operated and do not change any facility design feature, structure 
or system. No new or different type of equipment will be installed. 
Since there is no change to the facility or operating procedures, 
and the safety functions and reliability of structures, systems or 
components are not affected, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    Use of ZIRLO cladding material does not change the VANTAGE 5 
reload design and safety limits. The use of these fuel assemblies 
will take into consideration the normal core operating conditions 
allowed in the Technical Specifications. For each cycle reload core, 
the fuel assemblies will be 

[[Page 7559]]
evaluated using NRC-approved reload design methods, including 
consideration of the core physics analysis peaking factors and core 
average linear heat rate effects.
    The use of Zircaloy-4, ZIRLO or stainless steel filler rods in 
fuel assemblies will not involve a significant reduction in the 
margin of safety because analyses using NRC-approved methodologies 
will be performed for each configuration to demonstrate continued 
operation within the limits that assure acceptable plant response to 
accidents and transients. These analyses will be performed using 
NRC-approved methods that have been approved for application to the 
fuel configuration.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street NW., Washington, D.C. 20037.
    NRC Project Director: William H. Bateman.

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

    Date of amendment request: January 30, 1996.
    Description of amendment request: The proposed amendments would 
modify the Technical Specifications to increase the minimum allowable 
reactor coolant system total flow rate from 284,000 gpm (for Unit 1) 
and 275,300 gpm (for Unit 2) to 295,000 gpm for both units. Through the 
1980's and into the 1990's the North Anna Unit 1 and 2 steam generators 
experienced increasing levels of steam generator tube plugging. There 
was a corresponding decrease in the reactor coolant flow rate. As a 
result, the Commission issued several amendments in the 1989 to 1992 
time frame to reduce the minimum reactor coolant flow rate. 
Subsequently, the licensee replaced the steam generators in both units, 
with steam generators having an increased number of tubes compared to 
the replaced steam generators. With the increased number of tubes and 
less flow resistance, a greater reactor coolant flow rate is 
attainable. When the amendments were issued decreasing the minimum 
required reactor coolant flow rate, the transmittal letters stated the 
revision was temporary and would be increased when the steam generators 
were replaced.
    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 of occurrence or the consequences of an 
accident or malfunction of equipment important to safety previously 
evaluated in the safety analysis report would not increase. The 
proposed Technical Specifications change only increases the minimum 
allowable RCS total flow rate in the applicable Limiting Condition 
of Operation. No other changes are being made to allowable operating 
conditions defined by Technical Specifications, procedures, or to 
any plant design feature by the implementation of this change. There 
is no impact on the actual plant performance. Changes in the assumed 
initial conditions for the accident have no bearing on the 
probability of occurrence of the assumed accident or malfunction. 
The RCS flow rate is an assumption in applicable safety analyses. 
Existing analyses of record have assumed RCS flow rates which are 
bounding with respect to expected actual plant behavior. Therefore, 
the implementation of the proposed Technical Specifications change 
does not affect the probability nor increase the consequences of an 
accident previously evaluated.
    2. The possibility for an accident or malfunction of a different 
type than any evaluated previously in the safety analysis report 
would not be created. The proposed change to North Anna Units 1 and 
2 Technical Specifications Table 3.2-1 does not involve any 
alterations to the physical plant which would introduce any new or 
unique operational modes or accident precursors. Only the allowable 
value for measured Reactor Coolant System Total Flow Rate will be 
changed.
    3. The margin of safety as defined in the basis for any 
technical specifications is not reduced. The proposed Technical 
Specifications change only increases the minimum allowable RCS total 
flow rate in the applicable Limiting Condition of Operation. The RCS 
flow rate is an assumption in applicable safety analyses. Existing 
analyses of record have assumed RCS flow rates which are bounding 
with respect to expected actual plant behavior. Therefore, the 
margin of safety is not reduced by the proposed increase in the 
allowable RCS Total Flow Rate.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews.

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

    Date of amendment request: January 31, 1996.
    Description of amendment request: The amendments would revise the 
Technical Specifications to reduce the minimum volume of fuel that must 
be maintained in the diesel generator day tanks from 750 to 450 
gallons. The amendments would also revise the surveillance requirements 
for the diesel generators to permit some surveillances to be performed 
while the reactor units are at power where the licensee considers it 
safe to do so without compromising the availability of the diesel 
generators to perform their intended function.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve an increase in the probability of occurrence of an 
accident previously evaluated.
    The proposed changes do not result in any physical modifications 
to any plant systems or components nor change the operation of any 
plant equipment. The EDG [emergency diesel generator] fuel oil 
supply system will continue to provide adequate fuel supply to the 
EDGs in a manner consistent with applicable accident analyses. 
Performing surveillance tests or portions of surveillance tests at 
power that do not jeopardize stable plant operations does not 
increase the probability of occurrence of previously analyzed 
accidents.
    Therefore, there is no increase in the probability of occurrence 
of any accident.
    2. Increase the consequences of an accident previously 
evaluated.
    The proposed changes do not result in any physical modifications 
to any plant systems or components nor change the operation of any 
plant equipment. The EDG fuel oil system remains capable of 
supplying the EDGs with sufficient quantities of fuel oil to provide 
power for long term loss of offsite power. The EDG surveillances 
will continue to be performed in a manner that will ensure that the 
EDGs will be capable of performing their intended safety functions. 
The proposed changes to the electrical distribution system 
surveillances will continue to ensure that the electrical 
distribution system remains 

[[Page 7560]]
operable to power the required safety systems.
    Therefore, these proposed changes will not result in an increase 
in the consequences of any evaluated accidents.
    3. Create the possibility for an accident of a different type 
than was previously evaluated.
    The proposed changes do not result in any physical modifications 
to any plant systems or components nor change the operation of any 
plant equipment. Only those surveillance tests or portions of 
surveillance tests that do not jeopardize stable plant operation 
will be performed at power. Overlap testing to fully test the 
electrical distribution system protection functions does not 
introduce any unique accident precursors. The EDG fuel oil system 
remains capable of supplying the EDGs with sufficient quantities of 
fuel oil to provide power for long term loss of offsite power. The 
EDG surveillances will continue to be performed in a manner that 
will ensure that the EDGs will be capable of performing their 
intended safety functions.
    Therefore, there are no new precursors generated that would 
result in the possibility of a different type of an accident than 
was previously evaluated in the SAR [Safety Analysis Report].
    4. Decrease the margin of safety as described in the bases 
section of Technical Specifications.
    The EDG fuel oil system will continue to provide adequate fuel 
supply in a manner consistent with applicable accident analyses. The 
EDG surveillances will continue to be performed in a manner that 
will ensure that the EDGs are capable of performing their intended 
safety functions. The proposed changes to the electrical 
distribution system surveillances will continue to ensure that the 
electrical distribution system remains operable to power the 
required safety systems.
    Therefore, the margin of safety as described in the Technical 
Specifications is not reduced.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: David B. Matthews

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

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

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

    Date of application for amendment: July 21, 1995, August 8, 1995, 
and December 15, 1995.
    Brief description of amendment request: The proposed amendment 
would modify the requirements for testing an emergency diesel generator 
(EDG) when the other is inoperable. The amendment would correct an 
editorial error in the Duane Arnold Energy Center Operating License and 
would correct an erroneous reference in the Technical Specification.
    Date of publication of individual notice in Federal Register: 
February 2, 1996 (61 FR 3953).
    Expiration date of individual notice: March 4, 1996.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401.

Pacific Gas and Electric Company, Docket No. 50-275, Diablo Canyon 
Nuclear Power Plant, Unit No. 1, San Luis Obispo County, California

    Date of amendment request: January 18, 1995.
    Description of amendment request: The proposed amendment would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2, to allow operation of Unit 1 in 
Mode 3 (Hot Standby) during replacement of nonvital auxiliary 
transformer 1-1. Specifically, TS 3/4.8.1.1, ``Electrical Power 
Systems--A.C. Sources--Operating,'' Action Statement (a), would be 
revised to permit a one-time extension of the allowed outage time (AOT) 
from 72 hours to 120 hours.
    Date of individual notice in Federal Register: February 1, 1996 (61 
FR 3737).
    Expiration of individual notice: March 4, 1996.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of amendment request: February 5, 1996, as supplemented by 
letter dated February 14, 1996.
    Brief description of amendment request: The amendment changes 
Technical Specifications 4.6.2.3.b, ``Suppression Pool Cooling'', and 
TS 4.6.2.2.b, ``Suppression Pool Spray'', to include flow through the 
RHR heat exchanger bypass line (in addition to the RHR heat exchanger) 
in the Suppression Pool Cooling and Suppression Pool Spray flow path 
used during RHR pump testing.
    Date of publication of individual notice in Federal Register: 
February 9, 1996 (61 FR 5040).
    Expiration date of individual notice: March 11, 1996.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

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: January 16, 1996.
    Brief description of amendment request: The proposed amendment 
would change the Technical Specification surveillance frequency for the 
drywell bypass leakage rate test from 18 months to 120 months (10 
years) with a more frequent testing requirement if performance 
degrades. Additionally, specific leakage limits would be deleted for 
the air lock seal and barrel tests. Also, surveillance frequencies for 
the air lock interlock test and seal pneumatic system leak test would 
be changed from 18 months to 24 months. Finally, the surveillance 
frequencies for the air lock barrel test would be changed from ``each 
COLD SHUTDOWN if not performed within 

[[Page 7561]]
the previous 6 months'' to ``at least once per 24 months'' and from 18 
months to 24 months. The licensee requested that this amendment be 
approved for use during the current refueling outage which began on 
January 27, 1996.
    Date of publication of individual notice in Federal Register: 
February 2, 1996 (61 FR 3951).
    Expiration date of individual notice: March 4, 1996.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

Notice of Issuance of Amendments to Facility Operating Licenses

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

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit No. 1, Calvert County, 
Maryland

    Date of application for amendments: December 7, 1995.
    Brief description of amendments: The amendments add the convolution 
analytical technique for the analysis of the pre-trip main steam line 
break event to the list of approved core operating limits analytical 
methods listed in Technical Specification 6.9.1.9, ``Core Operating 
Limits Report.'' The convolution analytical technique was previously 
reviewed and approved by the NRC staff and the supporting safety 
evaluation was provided to Baltimore Gas and Electric Company by letter 
dated May 11, 1995.
    Date of issuance: February 5, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: 210 and 188.
    Facility Operating License No. DPR-53 and DPR-69: Amendment revised 
the Technical Specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
177)
    The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated February 5, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendments: November 2, 1994, as 
supplemented by letters dated November 16 and December 14, 1995.
    Brief description of amendments: The amendments delete the content 
of the Appendix B, ``Environmental Protection Plan'' (Non-radiological) 
Technical Specifications and modify License Condition 2.C.(2) so as to 
delete that portion which refers to the Environmental Protection Plan.
    Date of issuance: February 5, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-164--Unit 2-146.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications and License Conditions.
    Date of initial notice in Federal Register: March 1, 1995 (60 FR 
11131). The November 16 and December 14, 1995, letters provided 
clarifying information that did not change the scope of the November 2, 
1994, application and the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 5, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

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

    Date of application for amendments: January 13, 1995, as 
supplemented by letter dated August 30, 1995.
    Brief description of amendments: The amendments revise the 
Technical Specifications to increase the surveillance test intervals 
and allowed outage times for the Reactor Trip System and Engineered 
Safety Features Actuation System. The NRC staff has reviewed the 
proposed changes and finds that, with one exception as noted in the 
enclosed Safety Evaluation, the amendments conform to WCAP-10271, 
``Evaluation of Surveillance Frequencies and Out of Service Times for 
the Reactor Protection Instrumentation Systems,'' with its revisions 
and supplements, provides appropriate limiting conditions for operation 
and action statements, and is, therefore acceptable.
    Date of issuance: February 16, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-165--Unit 2-147.
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 15, 1995 (60 FR 
14019).
    The August 30, 1995, letter provided clarifying information that 
did not change the scope of the January 13, 1995, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 16, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223.

[[Page 7562]]


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: July 10, 1995.
    Brief description of amendments: These amendments modify the 
Technical Specifications to minimize the potential for boron dilution 
of the reactor coolant system (RCS) during startup of an isolated RCS 
loop. The changes permit RCS loop isolation only during Modes 5 and 6 
and require the RCS loop isolation valves be open with power removed 
from their valve operators during Modes 1, 2, 3, and 4. The changes 
also require isolation of primary grade water from the RCS during Modes 
4, 5, and 6, except during planned boron dilution or makeup activities.
    Date of issuance: February 12, 1996.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos.: 195 and 78.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42602).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 12, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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: July 20, 1995, as supplemented 
December 4, 1995.
    Brief description of amendments: These amendments revise Technical 
Specification 3/4.8.1.1, ``A.C. Sources-Operating,'' to incorporate 
guidance provided in NRC Generic Letter (GL) 84-15, ``Proposed Staff 
Actions to Improve and Maintain Diesel Generator Reliability,'' and GL 
93-05, ``Line-Item Technical Specification Improvements to Reduce 
Surveillance Requirements for Testing During Power Operation,'' which 
includes (1) revised requirements for testing the operable emergency 
diesel generators (EDGs) for various combinations of inoperable offsite 
circuits and EDGs and (2) revised surveillance requirements for the 
EDGs. The revised surveillance requirements include specifying 
generator voltage, frequency limits, and diesel starting time. The 
amendments also make several editorial changes to TS 3/4.8.1.1 to make 
TS 3/4 8.1.1 consistent with the guidance provided in the NRC's 
Improved Standard Technical Specifications (NUREG-1431).
    Date of issuance: February 12, 1996.
    Effective date: As of date of issuance, to be implemented within 60 
days.
    Amendment Nos.: 196 and 79.
    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 16, 1995 (60 FR 
42603).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 12, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: B.F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

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

    Date of application for amendments: November 22, 1995.
    Brief description of amendments: The amendments consist of changes 
relating to removal of the TS Bases from the TS index.
    Date of issuance: February 13, 1996.
    Effective date: February 13, 1996.
    Amendment Nos.: 182 and 176.
    Facility Operating Licenses Nos. DPR-31 and DPR-41: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65678).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 13, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Florida Power Corporation, et al., Docket No. 50-302, Crystal River 
Unit No. 3 Nuclear Generating Plant, Citrus County, Florida

    Date of application for amendment: May 31, 1995, as supplemented 
November 28, 1995, and December 21, 1995. The supplementary submittals 
did not affect the staff's proposed finding of no significant hazards 
consideration.
    Brief description of amendment: This amendment increases the 
surveillance interval on various instruments from 18 to 24 months.
    Date of issuance: February 13, 1996.
    Effective date: February 13, 1996.
    Amendment No.: 152.
    Facility Operating License No. DPR-72. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35070).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629.

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: October 16, 1995, as 
supplemented by letter dated December 22, 1995.
    Brief description of amendments: The amendments add a footnote to 
Technical Specification 4.6.1.2.d stating the Type B and C tests 
scheduled for Unit 1's refueling outage, cycle 6 (1R6) will be 
conducted in accordance with Option B of 10 CFR Part 50, Appendix J 
(hereafter referred to as Option B) using the guidance of Regulatory 
Guide 1.163, September 1995. This change only applies to Unit 1's 
refueling outage 1R6 because implementation of Option B for Type A, B, 
and C testing for both units is being incorporated into the Improved TS 
that are scheduled to become effective after refueling outage 1R6.
    Date of issuance: February 2, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-93--Unit 2-71.
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62490).
    The December 22, 1995, letter provided clarifying information that 
did not change the scope of the October 16, 1995, application and the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 2, 1996. 

[[Page 7563]]

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

    Date of application for amendment: June 9, 1995, as supplemented 
November 9, 1995.
    Brief description of amendment: The amendment relocates 
Surveillance Requirement 4.6.6.1.d.3 to TS 3.6.6.2 and revises the 
Action Statement of Section 3.6.6.1 to decouple it from Section 
3.6.6.2. In addition, Definition 1.12, ``Secondary Containment 
Boundary'' is deleted and included in the Bases Section 3/4.6.6, 
Secondary Containment. Bases Section 3/4.6.6.2, Secondary Containment 
is expanded using the guidance of the improved standard technical 
specifications (STS) for Westinghouse plants (NUREG-1431).
    Date of issuance: February 5, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 126.
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39445).
    The November 9, 1995, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 1996.
    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 Nos. 50-282 and 50-306, 
Prairie Island Nuclear Generating Plant, Unit Nos. 1 and 2, Goodhue 
County, Minnesota.

    Date of application for amendments: July 17, 1995, as supplemented 
October 16, 1995, and November 28, 1995.
    Brief description of amendments: The amendments revise the Prairie 
Island Radiological Effluent Technical Specifications and other 
sections relating to radiological controls to conform to NUREG-1431, 
``Standard Technical Specifications, Westinghouse Plants,'' Revision 1, 
and Generic Letter 89-01, ``Implementation of Programmatic Controls for 
Radiological Effluent Technical Specifications in the Administrative 
Controls Section of the Technical Specifications and the Relocation of 
Procedural Details of RETS to the Offsite Dose Calculation Manual or to 
the Process Control Program.''
    Date of issuance: January 24, 1996.
    Effective date: January 24, 1996, with full implementation within 
120 days.
    Amendment Nos.: Unit 1-122; Unit 2-115.
    Facility Operating License Nos. DPR-42 and DPR-60. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52933).
    By letters of October 16, 1995, and November 28, 1995, NSP 
forwarded a copy of its revised ODCM to the NRC for use as a reference 
and provided additional clarifying information. This information did 
not change the licensee's amendment request, the scope of the original 
Federal Register notice or the staff's initial proposed no significant 
hazards considerations determination. Therefore, renoticing was not 
warranted. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 24, 1996.
    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: July 28, 1995.
    Brief description of amendments: The amendment eliminates the 
Technical Specifications requirements to perform 10 CFR Part 50, 
Appendix J, Type C hydrostatic tests on certain valves that are assured 
a water seal following a Design Basis Accident.
    Date of issuance: February 8, 1996.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: 110 and 73.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49941).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 8, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: June 19, 1995, as supplemented 
December 21, 1995.
    Brief description of amendment: The amendment revises Technical 
Specification Section 2.2, ``Safety Limits,'' to change the minimum 
critical Power ratio safety Limit due to use of General Electric 13 
fuel product line.
    Date of issuance: February 8, 1996.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No. 111.
    Facility Operating License No. NPF-39. This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52934).
    The December 21, 1995, letter provided clarifying information that 
did not change the initial proposed no significant hazards 
consideration determination nor the Federal Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 8, 1996.
    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: July 28, 1995.
    Brief description of amendments: The amendments delete the 
operability and surveillance requirements involving secondary 
containment differential pressure instrumentation.
    Date of issuance: As of date of issuance, to be implmented within 
30 days.
    Effective date: February 14, 1996.
    Amendment Nos.: 112 and 74.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications. 

[[Page 7564]]

    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49942).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 14, 1996.
    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: July 28, 1995.
    Brief description of amendments: These amendmends revise Technical 
Specifications Table 4.3.1.1-1, ``Reactor Protection System 
Instrumentation Surveillance Requirements,'' to reflect changes the 
surveillance test frequency requirements for various Reactor Protection 
System instrumentation.
    Date of issuance: February 14, 1996.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment Nos.: 113 and 75.
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49944).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 14, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: January 20, 1995, as 
supplemented by letter dated December 18, 1995.
    Brief Description of amendment: The Technical Specification (TS) 
revision represents changes to TS Section 3/4.11.2.6, ``Explosive Gas 
Mixture,'' TS Table 3.3.7.11-1, ``Radioactive Gaseous Effluent 
Monitoring Instrumentation,'' and TS Table 4.3.7.11-1, ``Radioactive 
Gaseous Effluent Monitoring Instrumentation Surveillance 
Requirements.'' The revision removes these TS from the Technical 
Specifications and relocates the Bases to the Hope Creek Updated Final 
Safety Analysis Report and the Surveillance Requirements to the 
applicable surveillance procedures. The Limiting Conditions for 
Operation are eliminated.
    Date of issuance: February 6, 1996.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 91.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 2, 1995 (60 FR 
39452)
    The December 18, 1995 supplement did not effect the proposed no 
significant hazards determination, contained in the January 20, 1995 
application or the Federal Register notice (60 FR 39452).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 1996
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

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

    Date of application for amendment: October 7, 1995 as supplemented 
by letter dated October 27, 1995.
    Brief description of amendment: This amendment changes Technical 
Specification (TS) 4.8.1.1.2, ``A.C. Sources--Operating,'' by replacing 
the reference to an upper voltage and frequency band for the 10-second, 
Emergency Diesel Generator (EDG), starting time test with a minimum 
required voltage and frequency that must be attained within 10 seconds. 
The change to TS 4.8.1.1.2 also includes several related changes to TS 
4.8.1.1.2 as follows: (1) the requirement for an EDG to achieve 514 
rpm, within 10 seconds following a start signal during testing is 
eliminated, (2) the term ``standby'' replaces the term ``ambient'' in 
describing the EDG test restart condition, and (3) the term ``must'' is 
replaced with the term ``may'' in describing the use of manufacturers 
recommendations for EDG loading.
    Date of issuance: February 6, 1996.
    Effective date: As of date of issuance, to be implemented within 30 
days.
    Amendment No.: 92.
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 27, 1995 (60 
FR 58405)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: May 26, 1995, as supplemented 
May 5, 1995, and January 26, 1996.
    Brief description of amendment: The proposed change was to allow 
the storage of fuel with an enrichment not to exceed a nominal 5.0 
weight percent (w/o) Uranium-235 (U-235) in the new (fresh) and spent 
fuel storage racks and change the license to reflect changes related to 
the nuclear fuel cycle.
    Date of issuance: February 6, 1996.
    Effective date: February 6, 1996.
    Amendment No.: 60.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: September 26, 1995 (60 
FR 49636)
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 6, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

Rochester Gas and Electric Corporation, Docket No. 50-244, R. E. 
Ginna Nuclear Power Plant, Wayne County, New York

    Date of application for amendment: May 26, 1995, as supplemented by 
letters dated July 17, August 14, August 31, September 18, October 6, 
October 18, November 1, November 16, two letters of November 20, 
November 21, November 22, two letters of November 27, November 30, 
December 8, and December 28, 1995; and November 27, 1995; and May 23, 
1994, as supplemented by letters dated June 15, 1994, July 11, July 15, 
November 1, and November 16, 1995; and September 15, 1992, as 
supplemented April 20, 1993, April 26, 1995, and July 27, 1995.
    Brief description of amendment: (1) a full conversion from the 
licensee's current Technical Specifications (TSs) to a set of TSs based 
on NUREG-1431, 

[[Page 7565]]
``Standard Technical Specifications, Westinghouse Plants,'' Revision 0, 
dated September 1992 (including approved travellers used in the 
issuance of Revision 1, dated April 1995), in response to the 
licensee's application dated May 26, 1995, as supplemented by letters 
dated July 17, August 14, August 31, September 18, October 6, October 
18, November 1, November 16, two letters of November 20, November 21, 
November 22, two letters of November 27, November 30, December 8, and 
December 28, 1995. (2) a revision to the TSs to implement the amended 
regulation 10 CFR Part 50, Appendix J, Option B (new rule), to provide 
a performance based option for leakage-rate testing of containment, in 
response to the licensee's application dated November 27, 1995. (3) a 
revision to the TSs regarding allowable primary coolant levels of 
specific activity, in response to the licensee's application dated May 
23, 1994, as supplemented by letters dated June 15, 1994, July 11, July 
15, November 1, and November 16, 1995. (4) a revision to the TSs adding 
new requirements that enhance the reliability of power-operated relief 
valves and block valves (PORV/BV) along with TS changes that provide 
additional low-temperature overpressure protection, in response to the 
licensee's application dated September 15, 1992, as supplemented April 
20, 1993, and April 26, 1995. By letter dated July 27, 1995, the 
licensee withdrew this amendment request; however, the licensee 
rescinded this withdrawal request by letter dated December 28, 1995. 
Therefore, the proposed changes to the PORV/BV, as requested in the 
licensee's letter dated May 26, 1995, as supplemented December 28, 
1995, are incorporated into this amendment.
    Date of issuance: February 13, 1996.
    Effective date: February 13, 1996.
    Amendment No.: 61.
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: December 8, 1995 (60 FR 
63071); September 26, 1995 (60 FR 49636); August 30, 1995 (60 FR 
45184); July 6, 1994 (59 FR 34669).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 13, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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

    Date of application for amendments: December 30, 1993, as 
supplemented by letters dated June 3, 1994, August 25, 1994, January 3, 
1995, and January 19, 1995.
    Brief description of amendments: The amendments replace, in their 
entirety, the current technical specifications (TS) with a set of TS 
based on NUREG-1432, ``Standard Technical Specifications--Combustion 
Engineering Reactors,'' September 1992.
    Date of issuance: February 9, 1996.
    Effective date: February 9, 1996, to be implemented by August 9, 
1996.
    Amendment Nos.: Unit 1--Amendment No. 127; Unit 2--Amendment No. 
116.
    Facility Operating License Nos. NPF-10 and NPF-15: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49434) The January 3, 1995, and January 19, 1995, supplemental 
letters provided additional clarifying information and did not change 
the initial no significant hazards consideration determination. The 
Commission's related evaluation of the amendments is contained in a 
Safety Evaluation dated February 9, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Main Library, University of 
California, P. O. Box 19557, Irvine, California 92713.

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: December 8, 1995 (TS 95-24).
    Brief description of amendments: The amendments implement the 
change to 10 CFR Part 50, Appendix J to incorporate Option B, a 
voluntary performance-based option, for determining the frequency for 
performing Type A, B, and C Containment Leak Rate Testing.
    Date of issuance: February 5, 1996.
    Effective date: February 5, 1996.
    Amendment Nos.: 217 and 207.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
182).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 5, 1996.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402.

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

    Date of application for amendments: December 8, 1995 (TS 95-20).
    Brief description of amendments: The amendments decrease the 
frequency for conducting air or smoke tests of the containment spray 
system headers and Residual Heat Removal System headers from every 5 
years to every 10 years to verify each spray nozzle is unobstructed.
    Date of issuance: February 7, 1996.
    Effective date: February 7, 1996.
    Amendment Nos.: 218 and 208.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: January 3, 1996 (61 FR 
182).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 7, 1996.
    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: November 22, 1993 supplemented 
May 5 and December 20, 1995.
    Brief description of amendment: The amendment revised the Technical 
Specifications to reflect the replacement of analog temperature 
instrumentation associated with leak detection with digital equipment.
    Date of issuance: January 29, 1996.
    Effective date: January 29, 1996, and implemented not later than 
120 days following startup from the fifth refueling outage.
    Amendment No.: 79.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: May 12, 1994 (59 FR 
24752).
    The Commission's related evaluation of the amendment is contained 
in a 

[[Page 7566]]
Safety Evaluation dated January 29, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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: November 2, 1995, supplemented 
January 26, 1996.
    Brief description of amendment: The amendment only revised the 
containment personnel air lock Technical Specifications and added a 
license condition to allow the air locks to be open in Modes 4 and 5 
during core alterations except for movement of recently irradiated 
fuel. All other provisions of the request are being deferred for 
further review.
    Date of issuance: February 2, 1996.
    Effective date: To be implemented not later than 90 days after 
issuance.
    Amendment No. 80.
    Facility Operating License No. NPF-58: This amendment revised the 
Technical Specifications and added a license condition.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62497) The supplemental letter provided clarification of administrative 
controls that will be in place, did not change the initial no 
significant hazards consideration determination, and was within the 
scope of the notice issued December 6, 1995.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated February 2, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio 44081.

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 30, 1995, as supplemented by 
letters dated July 28, (TXX-95187), September 14, (TXX-95235), and 
November 29, 1995 (TXX-95299), and January 2, 1996 (TXX-96-003).
    Brief description of amendments: These changes authorized usage of 
the high density fuel storage racks, to increase the spent fuel storage 
capacity, and to adopt the wording, content, and format of the Improved 
Standard Technical Specifications.
    Date of issuance: February 9, 1996.
    Effective date: February 9, 1996.
    Amendment Nos.: Unit 1--Amendment No. 46; Unit 2--Amendment No. 32.
    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 
6313).
    The additional information contained in the supplemental letters 
dated July 28, (TXX-95187), September 14, (TXX-95235), and November 29, 
1995 (TXX-95299), and January 2, 1996 (TXX-96-003), was clarifying in 
nature and thus, within the scope of the initial notice and did not 
affect the staff's proposed no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendments is contained 
in an Environmental Assessment dated February 9, 1996, and a Safety 
Evaluation dated February 9, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 702 College, P.O. Box 
19497, Arlington, TX 76019.

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

    Date of application for amendments: September 19, 1995.
    Brief description of amendments: The amendments increase the 
surveillance test interval for the turbine reheat stop and intercept 
valves from at least once per 31 days to at least once per 18 months, 
extend the visual and surface disassembly inspection interval of the 
turbine reheat stop and intercept valves to 60 months and revise the 
inspection criteria for the throttle, governor, reheat stop, and reheat 
intercept valve disassembly inspections.
    Date of issuance: February 8, 1996.
    Effective date: February 8, 1996.
    Amendment Nos.: 195 and 176.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54725).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 8, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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

    Date of application for amendments: November 20, 1995, as 
supplemented January 23, 1996.
    Brief description of amendments: The amendments revise the North 
Anna Units 1 and 2 Technical Specifications to permit the use of 10 CFR 
Part 50, Appendix J, Option B, Performance-Based Containment Leakage 
Rate Testing.
    Date of issuance: February 9, 1996.
    Effective date: February 9, 1996.
    Amendment Nos.: 196 and 177.
    Facility Operating License Nos. NPF-4 and NPF-7. Amendments revised 
the Technical Specifications.
    Date of initial notice in Federal Register: December 20, 1995 (60 
FR 65685). The January 23, 1996 supplement provided clarifying 
information that was within the scope of the December 20, 1995 notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated February 9, 1996.
    No significant hazards consideration comments received: No.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498.

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 

[[Page 7567]]
the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) The 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By March 29, 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. 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 

[[Page 7568]]
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).

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

    Date of application of amendments: February 6, 1996.
    Brief description of amendments: The amendments revised Technical 
Specification Section 3.16, ``Containment Hydrogen Control Systems.'' 
The change adds a footnote to TS 3.16.3.b. to allow a one-time outage 
duration extension in regard to the Containment Hydrogen Control System 
flow path. This extension is necessary to install and test plant 
modifications, which will allow the Containment Hydrogen Control System 
to perform as designed, without the potential for inoperability due to 
water accumulation in the flow path.
    Date of Issuance: February 7, 1996.
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.: Unit 1-214-Unit 2-214-Unit 3-211.
    Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: The 
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 
February 7, 1996.
    Local Public Document Room location: Oconee County Library, 501 
West South Broad Street, Walhalla, South Carolina 29691.
    Attorney for licensee: J. Michael McGarry, III, Winston and Strawn, 
1200 17th Street NW., Washington, DC 20036.
    NRC Project Director: Herbert N. Berkow.

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 application for amendment: February 10, 1996.
    Brief description of amendment: The amendment revises Technical 
Specifications (TS) Surveillance Requirements 4.7.6.c.2, 4.7.6.d, 
4.9.11.b.2 and 4.9.11.c regarding the testing methodology utilized by 
Virgil C. Summer Nuclear Station, which determines the operability of 
the charcoal filters in the engineering safety features air handling 
units.
    Date of issuance: February 10, 1996.
    Effective date: February 10, 1996.
    Amendment No.: 131.
    Facility Operating License No. NPF-12: Amendment revises the TS.
    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 
February 10, 1996.
    Public comments requested as to proposed no significant hazards 
consideration: No.
    No significant hazards consideration comments received: None.
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180.

    Dated at Rockville, Maryland, this 21st day of February 1996.

    For the Nuclear Regulatory Commission.
Steven A. Varga,
Director, Division of Reactor Projects--I/II, Office of Nuclear Reactor 
Regulation.
[FR Doc. 96-4342 Filed 2-27-96; 8:45 am]
BILLING CODE 7590-01-P