[Federal Register Volume 61, Number 21 (Wednesday, January 31, 1996)]
[Notices]
[Pages 3497-3508]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20131]



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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 January 5, 1996, through January 19, 1996. 
The last biweekly notice was published on January 22, 1996.

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 1, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above. 

[[Page 3498]]

    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC, by the above 
date. Where petitions are filed during the last 10 days of the notice 
period, it is requested that the petitioner promptly so inform the 
Commission by a toll-free telephone call to Western Union at 1-(800) 
248-5100 (in Missouri 1-(800) 342-6700). The Western Union operator 
should be given Datagram Identification Number N1023 and the following 
message addressed to (Project Director): petitioner's name and 
telephone number, date petition was mailed, plant name, and publication 
date and page number of this Federal Register notice. A copy of the 
petition should also be sent to the Office of the General Counsel, U.S. 
Nuclear Regulatory Commission, Washington, DC 20555, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC, and at the local public document room for 
the particular facility involved.

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

    Date of amendment request: January 12, 1996
    Description of amendment request: Compliance with 10 CFR Part 50, 
Appendix J, provides assurance that the primary containment, including 
those systems and components that penetrate the primary containment, do 
not exceed the allowable leakage rate values specified in the Technical 
Specifications and Bases. The allowable leakage rate is determined so 
that the leakage assumed in the safety analyses is not exceeded.
    On February 4, 1992, the NRC published a notice in the Federal 
Register (57 FR 4166) discussing a planned initiative to begin 
eliminating requirements marginal to safety that impose a significant 
regulatory burden. Appendix J to 10 CFR Part 50, ``Primary Containment 
Leakage Testing for Water-Cooled Power Reactors,'' was considered for 
this initiative and the staff undertook a study of possible changes to 
this regulation. The study examined the previous performance history of 
domestic containments and examined the effect on risk of a revision to 
the requirements of Appendix J. The results of this study are reported 
in NUREG-1493, ``Performance-Based Leak-Test Program.''
    Based on the results of this study, the staff developed a 
performance based approach to containment leakage rate testing. On 
September 12, 1995, the NRC approved issuance of this revision to 10 
CFR Part 50, Appendix J, which was subsequently published in the 
Federal Register on September 26, 1995, and became effective on October 
26, 1995. The revision added Option B ``Performance-Based 
Requirements'' to Appendix J to allow licensees to voluntarily replace 
the prescriptive testing requirements of Appendix J with testing 
requirements based on both overall and individual component leakage 
rate performance.
    Regulatory Guide 1.163, ``Performance-Based Containment Leak Test 
Program,'' was developed as a method acceptable to the staff for 
implementing Option B. Accordingly, the licensee has submitted, in its 
application dated January 12, 1996, proposed changes to the TS to 
implement 10 CFR Part 50, Appendix J, Option B, by referring to 
Regulatory Guide 1.163, ``Performance-Based Containment Leakage-Test 
Program.''
    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:
    1The proposed change will not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Containment leak rate testing is not an initiator of any 
accident; the proposed change does not affect reactor operations or 
accident analysis, and has no significant radiological consequences. 
Therefore, this proposed change will not involve an increase in the 
probability or consequences of any previously-evaluated accident.
    2. The proposed change will not create the possibility of any 
new accident not previously evaluated.
    The proposed change does not affect normal plant operations or 
configuration, nor does it affect leak rate test methods. The test 
history at Catawba (no ILRT [integrated leak rate test] failures) 
provides continued assurance of the leak tightness of the 
containment structure.
    3. There is no significant reduction in a margin of safety. 
    
[[Page 3499]]

    The proposed changes are based on NRC-accepted provisions, and 
maintain necessary levels of reliability of containment integrity. 
The performanced-based approach to leakage rate testing recognizes 
that historically good results of containment testing provide 
appropriate assurance of future containment integrity; this supports 
the conclusion that the impact on the health and safety of the 
public as a result of extended test intervals is negligible.
    Based on the above, no significant hazards consideration is 
created by the proposed change.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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 amendment request: December 27, 1995
    Description of amendment request: The proposed amendments would 
modify Tables 3.3-11 and 4.3-7 of Beaver Valley Power Station Unit Nos. 
1 and 2 (BVPS-1 and BVPS-2) Technical Specification (TS) 3.3.3.8 such 
that only one valve position indication system for the power operated 
relief valves and safety valves is required to be operable. The 
licensee stated that the proposed amendments would then be consistent 
with the NRC's Improved Standard Technical Specifications, NUREG-1431, 
Revision 1, and with the guidance of Regulatory Guide 1.97, NUREG-0578, 
and NUREG-0737.
    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?
    The proposed change involves instrumentation which is redundant 
in monitoring the position of valves and, as such, does not 
influence the potential for an initiating event involving the power 
operated relief valves (PORVs) or the safety valves (SVs). 
Implementation of these changes will reduce the potential for 
challenges to the plant due to a potential shutdown which should not 
be necessary due to the restrictive nature of having unnecessary 
redundant position indication in the technical specification. By 
deleting the Unit No. 1 technical specification operability 
requirements for the PORV acoustic detectors, and by deleting, on 
both units, the technical specification operability requirements for 
the SV temperature detector position indicators, the potential for 
unnecessary shutdowns is reduced. When inoperable, the PORV acoustic 
detectors and the SV temperature detectors presently invoke an 
unnecessary action statement as another fully qualified safety-
related position indication system exists to provide indication. The 
proposed change modifies Specification 3.3.3.8 actions and 
surveillance requirements, but does not affect the BASES.
    The remaining instrumentation on these tables [3.3-11 and 4.3-7] 
will be unaffected. The remaining position indication systems for 
the PORVs and SVs are fully qualified and satisfy regulatory 
criteria for post accident monitoring of valve position. These 
changes do not affect the ability to satisfy analysis assumptions 
regarding operation of the PORVs and SVs. They do not affect the 
ability to continue to meet the guidance of Regulatory Guide 1.97, 
the post Three Mile Island criteria contained in NUREG 0578 and 
NUREG 0737, and reflect the guidance provided in NUREG 1431, 
``Improved Standard Technical Specifications'' (ISTS). Therefore, we 
have concluded that these changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated in the Updated Final Safety Analysis Report 
(UFSAR).
    2. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change will reduce the potential to challenge 
safety systems due to eliminating the potential for unnecessary 
plant shutdowns. The proposed changes are limited to PORV and SV 
position indication and do not involve any physical changes to the 
PORVs or SVs or their setpoints. These changes do not delete any 
design basis accident functions previously provided by the PORVs or 
SVs nor has the probability of inadvertent opening been increased. 
Accordingly, no new single failure has been identified as a result 
of these changes. Therefore, these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated in the UFSAR.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed changes have been incorporated to eliminate a 
degree of equipment redundancy and is consistent with the Improved 
Standard Technical Specifications (ISTS). The Unit No. 1 
specification presently requires operability of both redundant PORV 
position indication systems and the primary and backup SV position 
indication systems. The Unit No. 2 specification also requires 
operability of the primary and backup SV position indication 
systems. These changes will potentially eliminate some challenges 
and potential unnecessary shutdowns by eliminating equipment 
determined to be no longer necessary. Only one safety-related 
position indication system is necessary to satisfy regulatory 
criteria; therefore, operation of the plant in accordance with the 
proposed amendment would not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

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

    Date of amendment request: December 22, 1995
    Description of amendment request: The proposed amendment would 
revise the Duane Arnold Energy Center (DAEC) Technical Specifications 
(TS) Sections 3.7.A and 4.7.A, ``Primary Containment,'' by deleting 
information also contained in 10 CFR Part 50, Appendix J, Option A and 
incorporating references to the Primary Containment Leakage Rate 
Testing Program. These changes will allow the use of the performance 
based option of containment leak testing. The request also adds 
Operability and Surveillance Requirements (SRs) for the drywell air 
lock. Minor administrative changes are also made. These changes are 
consistent with comparable specifications in the Improved Standard 
Technical Specifications (ITS), NUREG-1433. In addition to the 
licensee's proposed revision to the DAEC TS, the staff will be 
executing administrative changes and corrections to the TS Bases, as 
submitted in letters(2) dated February 13, 1995. Sections that will be 
changed or corrected are Section 1.2, Bases; Section 2.2, Bases Reactor 
Coolant System Integrity; Section 3.2, Bases; Section 3.7.H/4.7.H, 
Bases Containment Atmosphere Dilution; and Section 3.7.I/4.7.I, Bases 
Oxygen Concentration.
    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 

[[Page 3500]]
consideration, which is presented below:
    . The proposed revision does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Information contained in 10 CFR 50, Appendix J was deleted and 
references to the Primary Containment Leakage Rate Testing Program 
were added. These are administrative changes to allow the use of 
performance-based containment leakage testing methods. The 
containment testing program will conform with the requirements of 
Option B of 10 CFR Part 50, Appendix J and approved exemptions. The 
performance of the leakage tests themselves is not an input or 
consideration in any accident previously evaluated, thus the 
proposed change will not increase the probability of any such 
accident occurring. The same operability requirements remain for the 
primary containment, therefore the consequences of an accident are 
not significantly increased.
    Drywell air lock operability and surveillance requirements were 
added. Actions for one air lock door inoperable have been added 
consistent with the ITS. In addition, notes have been added to allow 
entry and exit to perform repairs of the air lock components and to 
explain that the previous overall leak test is not invalidated by an 
inoperable door. This change represents an additional restriction on 
plant operation, since the previous condition of one air lock door 
inoperable did not require any actions to be taken. A requirement to 
verify proper operation of interlock mechanism was also added. This 
will ensure that one door is always closed which maintains primary 
containment integrity.
    The addition of these new drywell air lock requirements provides 
more stringent provisions than previously existed in the [current 
Technical Specifications]. The more stringent requirements will not 
result in operation that will increase the probability of initiating 
an analyzed event. If anything, the new requirements may decrease 
the probability or consequences of an analyzed event by 
incorporating the more restrictive changes discussed above. These 
changes will not alter assumptions relative to mitigation of an 
accident or transient event. The more restrictive requirements will 
not alter the operation of process variables, structures, systems, 
or components as described in the safety analyses.
    The TS revision includes the relocation of certain requirements 
from the current technical specification (CTS) to licensee 
controlled documents. CTS 4.7.A.1.e contains a requirement to 
replace the T-ring inflatable seals for the 18 inch purge valves 
every four years. This provision is not in the ITS as it is a 
maintenance issue and not a surveillance for operability. CTS 
4.7.A.1.e also contains a requirement to verify (during Type C 
testing) that the mechanical modification which limits the maximum 
opening angle for the 18 inch purge valves is intact. The ITS only 
requires this surveillance if the mechanical modification is not 
permanent. At DAEC, the 18 inch purge valves are permanently blocked 
to restrict opening to 30 deg.. These CTS provisions will be 
relocated to plant procedures. Any changes to these relocated 
requirements will require an evaluation in accordance with 10 CFR 
50.59. CTS 4.7.A.1.a and 4.7.A.1.d contain some procedural details 
that are not contained in Appendix J. These details will also be 
relocated to plant procedures, consistent with the ITS. Since any 
changes to these licensee controlled documents will be evaluated in 
accordance with 10 CFR 50.59, no significant increase in the 
probability or consequences of an accident previously evaluated will 
be allowed.
    The proposed revision does not involve any change to the 
configuration or method of operation of any plant equipment that is 
used to mitigate the consequences of an accident, nor does it affect 
any assumptions or conditions in the accident analysis. The proposed 
revision does not degrade any existing plant programs, nor modify 
any functions of safety related systems or accident mitigation 
functions previously credited at the DAEC. The proposed changes do 
not impact initiators of analyzed events. They also do not impact 
the assumed mitigation of accidents or transient events. These TS 
changes will not alter assumptions made in the safety analysis and 
licensing basis.
    Therefore, the proposed revision does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed revision does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Deleting information from the TS which is contained in 10 CFR 
50, Appendix J and adding references to the Primary Containment 
Leakage Rate Testing Program are purely administrative changes to 
allow the use of performance-based containment leakage testing 
methods. The containment testing program will conform with the 
requirements of Option B of 10 CFR Part 50, Appendix J and approved 
exemptions. The use of Option B will maintain the containment safety 
functions as a barrier to the release of radioactivity to the 
environment.
    The proposed revision does not make any physical or operational 
changes to existing plant systems or components, nor does it alter 
any plant parameters, revise any safety limit setpoint, or provide 
any new release pathways. The proposed revision does not change any 
transient responses assumed in the Design Bases of the plant.
    The proposed changes which relocate requirements to licensee 
controlled documents will not alter the plant configuration (no new 
or different type of equipment will be installed) or change the 
methods governing normal plant operation. These changes will not 
alter assumptions made in the safety analysis or licensing basis.
    The proposed changes which add more restrictive requirements to 
the CTS will not alter the plant configuration (no new or different 
type of equipment will be installed) or change the methods governing 
normal plant operation. These changes do impose different 
requirements. However, they are consistent with assumptions made in 
the safety analyses.
    Therefore, the revision does not create the possibility of a new 
or different kind of accident previously evaluated.
    3. The proposed revision will not significantly reduce any margin 
of safety.
    Deleting information from the TS which is contained in 10 CFR 
50, Appendix J and adding references to the Primary Containment 
Leakage Rate Testing Program do not involve a significant reduction 
in the margin of safety. These changes are administrative in nature 
and either eliminate a redundant requirement or clarify the 
applicability and acceptability of an alternative, NRC approved, 
leak rate testing provision within the TS. The containment testing 
program will conform to the requirements of Option B of 10 CFR Part 
50, Appendix J and approved exemptions. The use of Option B will 
maintain the containment safety functions as a barrier to the 
release of radioactivity to the environment.
    The proposed revision does not require any modifications to 
existing plant systems or equipment, safety limit settings, or 
parameters utilized in the licensing bases for the safety analysis. 
The proposed revision does not change any safety analysis or any 
accident mitigation action for which DAEC has previously taken 
credit. The proposed changes do not involve any technical changes; 
they have no impact on any safety analysis assumptions. The addition 
of new requirements either increases or does not affect the margin 
of safety.
    The proposed changes that relocate requirements from the CTS to 
licensee controlled documents will not reduce a margin of safety 
since they have no impact on any safety analysis assumptions. In 
addition, the requirements to be relocated from the CTS to the 
licensee controlled document are unchanged. Since any future changes 
to this licensee controlled document will be evaluated in accordance 
with the requirements of 10 CFR 50.59, no significant reduction in a 
margin of safety will be allowed.
    The proposed changes are consistent with NUREG-1433, which was 
approved by the NRC Staff. The changes are also consistent with NRC 
guidance provided for the implementation of Option B. The change 
controls for proposed relocated details and requirements are 
acceptable. Therefore, revising the TS to reflect the NRC accepted 
level of detail and requirements ensures that there is no reduction 
in a margin of safety.
    Therefore, the proposed revision will not significantly reduce 
any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S.E., Cedar Rapids, Iowa 52401 

[[Page 3501]]

    Attorney for licensee: Jack Newman, Kathleen H. Shea, Morgan, 
Lewis, & Bockius, 1800 M Street, NW., Washington, DC 20036-5869
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: December 14, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specifications 3.3.1.1, ``Reactor Protection System 
(RPS) Instrumentation,'' and 3.3.6.1, ``Primary Containment and Drywell 
Isolation Instrumentation,'' to eliminate periodic response time 
testing of selected analog trip modules (ATMs). This request is 
supported by analyses prepared by the Boiling Water Reactor Owners' 
Group topical report NEDO-32291, ``System Analyses for Elimination of 
Selected Response Time Testing Requirements,'' which demonstrate that 
other periodic tests required by technical specifications, such as 
channel calibrations, channel functional tests and logic system 
functional tests, are adequate to ensure ATM response times remain 
within acceptable limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    (1) The purpose of the proposed Technical Specification (TS) 
change is to eliminate response time testing requirements for 
selected analog trip modules (ATMs) in the Reactor Protection System 
(RPS) and the main steam isolation valve (MSIV) isolation actuation 
instrumentation. The Boiling Water Reactor Owners' Group (BWROG) has 
completed an evaluation which demonstrates that response time 
testing is redundant to the other TS-required testing. These other 
tests, in conjunction with actions taken in response to NRC Bulletin 
90-01, ``Loss of Fill-Oil in Transmitters Manufactured by 
Rosemount,'' and Supplement 1, are sufficient to identify failure 
modes or degradations in instrument response time and ensure 
operation of the associated systems within acceptable limits. There 
are no known failure modes that can be detected by response time 
testing that cannot also be detected by other TS-required testing. 
This evaluation was documented in NEDO-32291, ``System Analyses for 
Elimination of Selected Response Time Testing Requirements,'' 
January 1994. Illinois Power (IP) has confirmed the applicability of 
this evaluation to Clinton Power Station (CPS). In addition, IP has 
completed the actions identified in the NRC staff's safety 
evaluation of NEDO-32291.
    Because of the continued application of other existing TS-
required tests such as channel calibrations, channel checks, channel 
functional tests, and logic system functional tests, the response 
time of these systems will be maintained within the acceptance 
limits assumed in plant safety analyses and required for successful 
mitigation of an initiating event. The proposed changes do not 
affect the capability of the associated systems to perform their 
intended function within their required response time, nor do the 
proposed changes themselves affect the operation of any equipment. 
As a result, IP has concluded that the proposed changes do not 
involve a significant increase in the probability or the 
consequences of an accident previously evaluated.
    (2) The proposed changes only apply to the testing requirements 
for ATMs in the systems identified above and do not result in any 
physical change to these or other components or their operation. As 
a result, no new failure modes are introduced. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    (3) The current TS-required response times are based on the 
maximum values assumed in the plant safety analyses. These analyses 
conservatively establish the margin of safety. As described above, 
the proposed changes do not affect the capability of the associated 
systems to perform their intended function within the allowed 
response time used as the basis for the plant safety analyses. The 
potential failure modes for the components within the scope of this 
request were evaluated for impact on instrument response time. This 
evaluation confirmed that the remaining TS-required testing is 
sufficient to identify failure modes or degradations in instrument 
response times and to ensure that operation of the instrumentation 
within the scope of this request is within acceptable limits. As a 
result, it has been concluded that plant and system response to an 
initiating event will remain in compliance with the assumptions of 
the safety analysis.
    Further, although not explicitly evaluated, the proposed changes 
will provide an improvement to plant safety and operation by 
reducing the time safety systems are unavailable, reducing the 
potential for safety system actuations, reducing plant shutdown 
risk, limiting radiation exposure to plant personnel, and 
eliminating the diversion of key personnel resources to conduct 
unnecessary testing. Therefore, IP has concluded that this request 
will result in an overall increase in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Sheldon Zabel, Esq., Schiff, Hardin and 
Waite, 7200 Sears Tower, 233 Wacker Drive, Chicago, Illinois 60606
    NRC Project Director: Gail H. Marcus

Northern States Power Company, Docket No. 50-263, Monticello 
Nuclear Generating Plant, Wright County, Minnesota

    Date of amendment request: December 11, 1995
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) Section 4.7, Surveillance 
Requirements for Primary Containment Automatic Isolation Valves. 
Specifically, the proposed amendment would delete TS Surveillance 
Requirement 4.7.D.4, which requires replacement of the seat seals for 
the drywell and suppression chamber purge and vent valves every 5 
years.
    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 amendment will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    An evaluation of the operational performance of the 18-inch 
purge and vent valves has concluded that deletion of the Monticello 
Technical Specification surveillance requirement 4.7.D.4 will have 
no adverse impact on the seat leakage performance of these primary 
containment isolation valves, no adverse impact on the testing 
performed in accordance with 10 CFR 50, Appendix J, and thus no 
adverse impact on the containment isolation function of these 
primary containment isolation valves. The material of which the T-
shaped elastomer seat is comprised of has been found to withstand 
normal and accident thermal exposures for the design life of the 
plant based on a thermal aging analysis. Radiation effects will not 
have an adverse impact on the elastomer seat material. Therefore, 
this amendment will not cause a significant increase in the 
probability or consequences of an accident previously evaluated for 
the Monticello plant.
    The proposed amendment will not create the possibility of a new 
or different kind of accident from any accident previously analyzed.
    The proposed change to the Technical Specifications for the 
Primary Containment Purge and Vent valves does not alter the 
function of these components or their interrelationships with other 
systems. Therefore, this amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    The proposed amendment will not involve a significant reduction 
in the margin of safety.
    The operating experience of these valves has demonstrated that 
the testing performed 

[[Page 3502]]
in accordance with 10 CFR 50, Appendix J, provides a high level of 
confidence in the ability of these valves to perform their safety 
function with respect to valve leak tightness. The proposed 
amendment will 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: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment requests: December 27, 1995
    Description of amendment requests: The amendments would revise the 
combined Technical Specifications (TS) 3/4.6.1.1, ``Containment 
Integrity;'' 3/4.6.1.2, ``Containment Leakage;'' 3/4.6.1.3, 
``Containment Air Locks;'' 3/4.6.1.6, ``Containment Structural 
Integrity;'' 3/4.6.3, ``Containment Isolation Valves;'' and their 
associated Bases; and would add TS 6.8.4.j, ``Containment Leakage Rate 
Testing Program,'' to implement the performance-based leakage rate 
testing program, as permitted by 10 CFR Part 50, Appendix J. These 
changes will support the implementation of the performance-based 
testing of Option B to Appendix J for Types A, B, and C containment 
leakage rate testing and the appropriate rescheduling of testing. The 
amendment changes the TS to implement 10 CFR Part 50, Appendix J, 
Option B, by referring to Regulatory Guide 1.163, ``Performance-Based 
Containment Leakage Test Program.''
    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 Technical Specification (TS) 3/4.6.1.1, 
3/4.6.1.2, 3/4.6.1.3, 3/4.6.1.6, 3/4.6.3, and the addition of 6.8.4 
j., to implement the performance-based Containment Leakage Rate 
Testing Program have no effect on plant operation. The proposed 
changes only provide mechanisms within the TS for implementing a 
performance-based methodology for determining the frequency of leak 
rate testing that has been approved by the Commission. The test type 
and test method used for testing would not be changed. The test 
acceptance criteria would not be changed, and containment leakage 
will continue to be maintained within the required limits.
    Directly referencing the Containment Leakage Rate Testing 
Program for containment ILRT [integrated leak rate testing] and LLRT 
[local leak rate test] requirements does not involve any 
modification to plant equipment or affect the operation or design 
basis of the containment. Leakage rate testing is not a precursor to 
or an initiating event for any accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes only allow for the implementation of 10 CFR 
50, Appendix J, Option B, testing frequencies and do not involve any 
modifications to any plant equipment or affect the operation or 
design basis of the containment. The proposed changes do not affect 
the response of the containment during a design basis accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect or change a Safety Limit or 
affect plant operations. The changes only implement the allowed 10 
CFR 50, Appendix J, Option B testing frequencies that have been 
determined by the Commission not to involve a safety concern. The 
testing method, acceptance criteria, and basis for testing are not 
changed and still provide assurance that the containment will 
provide its intended function.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

Saxton Nuclear Experimental Corporation (SNEC), Docket No. 50-146, 
Saxton Nuclear Experimental Facility (SNEF), Bedford County, 
Pennsylvania

    Date of amendment request: November 21, 1995.
    Description of amendment request: The proposed amendment would 
change the license and technical specifications to add GPU Nuclear 
Corporation (GPUN) as a licensee for the SNEF along with SNEC and would 
transfer from SNEC to GPUN all management-related responsibilities for 
the SNEF. Responsibility for safely maintaining the containment vessel 
and performing characterization activities would change from SNEC to 
GPUN. Technical specification organizational positions would be changed 
from SNEC titles to GPUN titles. GPUN would take responsibility from 
SNEC for administration of all SNEF functions, for radiation safety 
activities, and for providing on-site management and continuing 
oversight of production activities. The appointment of members to the 
Saxton Radiation Safety Committee and the reporting of the Committee 
would change from the SNEC President to the GPUN Vice President of the 
Nuclear Services Division. The GPUN President would have the authority 
to request audits and would receive audit reports instead of the SNEC 
President. Procedure control methodology and the administrative 
procedure for procedures would be changed from SNEC procedures to GPUN 
procedures. The responsibility for records retention and reporting 
would change from SNEC to GPUN. The organization chart for the facility 
would be changed to reflect the addition of GPUN as a licensee.
    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 
considerations because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Because the proposed changes are administrative in nature they 
would have no effect on the likelihood or impact on the potential 
accidents of fire, flood or radiological hazard. 

[[Page 3503]]

    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    Because the proposed changes are administrative in nature they 
would not create the possibility of a new or different kind of 
accident from any accident previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    Because the proposed changes are administrative in nature they 
would not involve any reduction in a margin of safety.
    The NRC staff has reviewed the analysis of the licensee 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: Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678 Attorney for the Licensee: 
Ernest L. Blake, Jr., Esquire, Shaw, Pittman, Potts, and Trowbridge, 
2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Seymour H. Weiss

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

    Date of amendment request: January 4, 1996 (TS 95-22)
    Description of amendment request: The proposed change would extend 
the functional testing interval for the following isolation radiation 
monitor instruments from monthly to quarterly: (1) Engineered Safety 
Feature Actuation System Instrumentation Surveillance Requirements 
Table 4.3-2, Item 3.c.3, Containment Purge Air Exhaust Monitor 
Radioactivity-High; (2) Radiation Monitoring Instrumentation 
Surveillance Requirements Table 4.3-3, Item 1.a, Fuel Storage Pool Area 
Radiation Monitor; (3) Table 4.3-3, Item 2.a, Containment Purge Air 
Exhaust; (4) Table 4.3-3, Item 2.b.i, Containment Gaseous Activity RCS 
Leakage Detection; (5) Table 4.3-3, Item 2.b.ii, Containment 
Particulate Activity RCS Leakage Detection; and (6) Table 4.3-3, Item 
2.c, Control Room Isolation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    TVA has evaluated the proposed technical specification (TS) 
change and has determined that it does not represent a significant 
hazards consideration based on criteria established in 10 CFR 
50.92(c). Operation of Sequoyah Nuclear Plant (SQN) in accordance 
with the proposed amendment will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Review of the past history for the affected and similar 
radiation monitors revealed that extending the functional testing 
interval for these monitors will not adversely affect system 
operability and will effectively increase system availability. These 
radiation monitors are not accident initiating equipment, thus 
increasing the surveillance interval on these monitors will not 
affect the probability of any accident previously evaluated. Based 
on the above statements, it is concluded that the probability or 
consequences of an accident previously evaluated is not increased.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    No new type of accident or malfunction will be created since the 
radiation monitors are not accident initiating equipment. The 
proposed change merely increases the functional testing interval for 
the affected radiation monitors, and does not change the method and 
manner of plant operation. The safety design bases in the Updated 
Final Safety Analysis Report have not been altered.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes do not change the plant configuration in a 
way that introduces a new potential hazard to the plant and do not 
involve a significant reduction in the margin of safety. The 
proposed changes do not affect applicable safety analysis acceptance 
criteria and will not affect system operating conditions. 
Additionally, plant operating experience with similar monitors has 
shown that there has not been additional failures due to the 
quarterly testing frequency. Thus, it is concluded that the margin 
of safety is not reduced.
    The NRC 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: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET 11H, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: November 22, 1995
    Description of amendment request: The proposed amendment replaces 
the requirements associated with the boron dilution mitigation system 
(BDMS) in the Wolf Creek Generating Station Technical Specifications 
with alarms, indicators, procedures, and controls to assure proper 
resolution of potential inadvertent boron dilution events.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    . The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The only event potentially impacted by the proposed change is 
the inadvertent boron dilution event. The discussion of the 
probability and consequences of an inadvertent boron dilution event 
at WCGS is provided in USAR [Updated Safety Analysis Report] Section 
15.4.6. Primarily, the proposed changes revise the method of 
detecting and mitigating the event. The only aspect of the changes 
that impact[s] the potential causes of an inadvertent boron dilution 
event is the increased requirement to isolate potential dilution 
sources in Modes 3, 4 and 5. As a result, the overall probability of 
the event is slightly decreased.
    The alternate methods to detect and mitigate this event achieve 
the same basic goal as the current BDMS; to prevent a return to 
critical during an inadvertent dilution event. The proposed changes 
to the BDMS will result in an improved system that will provide an 
improved response to the inadvertent boron dilution event, and that 
will prevent a return to critical. Thus, it can be concluded that 
the proposed change will not significantly increase the consequences 
of a postulated inadvertent boron dilution event.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The revisions to plant procedural requirements to either operate 
a reactor coolant pump or to isolate/control potential dilution 
sources does [sic] not create the potential for a new or different 
kind of accident because these new requirements are configurations 
which have always been allowed. Similarly, the new normal position 
for the letdown divert valve does not create a new or different 
accident because the new normal position has always been an allowed 
position. The other procedural changes only increase the plant 
operators' awareness of potential boron dilution problems or provide 
the steps needed to respond to available indications and alarms to 
mitigate the potential event. As a result, these procedural changes 
do not create the possibility of a new or different kind of 
accident.
    The proposed changes also include addition of new redundant VCT 
high level alarms and a new alarm indicating that the 

[[Page 3504]]
letdown divert valve is not in the ``VCT'' position. Because the alarms 
are passive, they do not create the possibility of a new or 
different kind of accident.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The design criterion and margin of safety for the current BDMS 
is that the dilution event is terminated prior to the loss of all 
shutdown margin. The same criterion will be met following the 
implementation of the proposed changes. Therefore, there is no 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: December 20, 1995
    Description of amendment request: This amendment request proposes 
to revise Technical Specification 3/4.6.1.1, ``Containment Integrity,'' 
and 3/4.6.1.3, ``Containment Air Locks,'' and to add Technical 
Specification 6.8.4i, ``Containment Leakage Rate Testing Program,'' to 
implement the new performance-based leakage rate testing program as 
permitted by 10 CFR 50, Appendix J. Also, Technical Specification 1.7e, 
``Containment Integrity,'' would be revised to reference Technical 
Specification 4.6.1.1.c. These proposed changes will implement the 
performance-based testing of Option B to Appendix J, for Type A, B, and 
C containment leak testing by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leakage-Test Program.''
    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 Technical Specifications 3/4.6.1.1 and 
3/4.6.1.3, and the addition of Technical Specification 6.8.4i to 
implement the new performance based Containment Leakage Rate Testing 
Program, have no effect on plant operation. The proposed changes 
only provide mechanisms within the technical specifications for 
implementing a performance-based methodology, for determining the 
frequency of leak rate testing, which has been approved by the NRC. 
The test type and test method used for testing would not be changed. 
The test acceptance criteria would not be changed, and containment 
leakage will continue to be maintained within the required limits.
    Directly referencing the Containment Leakage Rate Testing 
Program for containment integrated leak rate test and local leak 
rate test requirements does not involve any modification to plant 
equipment or affect the operation or design basis of the 
containment. Leakage rate testing is not a precursor to or an 
initiating event for any accident.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes only allow for implementation of 10 CFR 50, 
Appendix J, Option B, testing frequencies and do not involve any 
modifications to any plant equipment or affect the operation or 
design basis of the containment. The proposed changes do not affect 
the response of the containment during a design basis accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not affect or change a Safety Limit, any 
limiting condition for operation or affect plant operations. The 
changes only implement the allowed Option B testing frequencies that 
have been determined by the NRC not to involve a safety concern. The 
testing method, acceptance criteria, and bases are not changed and 
still provide assurance that the containment will provide its 
intended function.
    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 locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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. 

[[Page 3505]]


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

    Date of application for amendments: November 14, 1995, as 
supplemented January 4, 1996.
    Brief description of amendments: The amendments revise the 
Technical Specifications to incorporate 10 CFR Part 50, Appendix J, 
``Primary Reactor Containment Leakage Testing for Water-Cooled Power 
Reactors,'' Option B. Technical Specification changes for the LaSalle 
facility will be addressed under separate correspondence.
    Date of issuance: January 11, 1996
    Effective date: January 11, 1996
    Amendment Nos.: 148, 142, 169, and 165
    Facility Operating License Nos. DPR-19, DPR-25, DPR-29 and DPR-30. 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 7, 1995 (60 FR 
62896). The January 4, 1996, supplement provided a specific 
implementation date for the requested amenement. This information was 
within the scope of the original application and did not change the 
staff's initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 1996. No significant 
hazards consideration comments received: No
    Local Public Document Room location: for Dresden, Morris Area 
Public Library District, 604 Liberty Street, Morris, Illinois 60450; 
for Quad Cities, Dixon Public Library, 221 Hennepin Avenue, Dixon, 
Illinois 61021.

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

    Date of application for amendment: November 14, 1995
    Brief description of amendment: The amendment revises the Haddam 
Neck Technical Specifications (TS) to provide an one-time exception to 
TS 3.9.12, '' Fuel Building Storage Air Cleanup System,'' to allow the 
fuel storage building air cleanup system to be inoperable for a limited 
duration during intervals in which new fuel rack modules will be moved 
into and old fuel rack modules will be moved out of the fuel storage 
building.
    Date of Issuance: January 17, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 187
    Facility Operating License No. DPR-61. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 28, 1995 (60 
FR 58688) The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated January 17, 1996 No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.

Consumers Power Company, Docket No. 50-155, Big Rock Point Plant, 
Charlevoix County, Michigan Date of application for amendment: 
November 8, 1995, as supplemented November 17, 1995

    Brief description of amendment: The amendment removes the 
prescriptive Type A containment leakage test rate frequency of 40 plus 
or minus 10 months and adds a reference to perform containment leakage 
rate tests in accordance with the criteria specified in Appendix J of 
10 CFR Part 50 as modified by approved exemptions. In addition, the 
amendment revises the test pressure for Type B and C testing to correct 
a typographical error.
    Date of issuance: January 16, 1996
    Effective date: January 16, 1996
    Amendment No.: 117
    Facility Operating License No. DPR-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62489) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 16, 1996. No significant 
hazards consideration comments received: No.
    Local Public Document Room location:  North Central Michigan 
College, 1515 Howard Street, Petoskey, Michigan 49770.

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: March 29, 1995, as supplemented 
by letters dated September 18 and November 16, 1995
    Brief description of amendments: The amendments revise Technical 
Specification requirements for the Low Temperature Overpressure 
Protection system and update the heatup and cooldown curves for both 
units.
    Date of issuance: January 11, 1996
    Effective date: As of the date of issuance to be implemented within 
60 days
    Amendment Nos.:  Unit 1 - 162; Unit 2 - 144
    Facility Operating License Nos. NPF-9 and NPF-17: Amendments 
revised the Technical Specifications. Date of initial notice in Federal 
Register: September 27, 1995 (60 FR 49933) The September 18 and 
November 16, 1995, letters provided clarifying information that did not 
change the scope of the March 29, 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 January 11, 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

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

    Date of application for amendments: December 2, 1994
    Brief description of amendments: The amendments replace Appendix B, 
``Environmental Technical Specifications,'' with an Environmental 
Protection Plan (Nonradiological) and revise the Operating Licenses to 
reflect these changes.
    Date of issuance: December 19, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: Unit 1 - 199 - Unit 2 - 140
    Facility Operating License Nos. DPR-57 and NPF-5. Amendments 
revised the Technical Specifications and Operating Licenses.
    Date of initial notice in Federal Register: January 4, 1995 (60 FR 
502) The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 19, 1995. No significant hazards 
consideration comments received: No
    Local Public Document Room location: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513 

[[Page 3506]]


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: March 17, 1995, as supplemented 
by letter dated July 6, 1995
    Brief description of amendments: The amendments revise Technical 
Specification 3/4.9.4, Containment Building Penetrations, to allow the 
personnel airlock to be open during core alterations or movement of 
irradiated fuel within the containment.
    Date of issuance: November 30, 1995
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 92 and 70
    Facility Operating License Nos. NPF-68 and NPF-81: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: July 5, 1995 (60 FR 
35077) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 30, 1995. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Burke County Library, 412 
Fourth Street, Waynesboro, Georgia 30830

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

    Date of amendment request: August 17, 1995, as supplemented by 
letters dated November 22, and December 18, 20, and 27, 1995
    Brief description of amendment: The amendment revised the primary 
containment air lock technical specifications to allow the air locks to 
be open in Mode 5 (refueling) during core alterations except for 
movement of recently irradiated fuel. All other provisions of the 
August 17, 1995, requests are defered.
    Date of issuance: January 11, 1996
    Effective date: January 11, 1996
    Amendment No.: 85
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 13, 1995 (60 
FR 47619) The additional information contained in the supplemental 
letters dated November 22, and December 18, 20, and 27, 1995, was 
clarifying in nature and thus, within the scope of the initial notice 
and did not affect the staff's proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated Janaury 11, 1996. 
No significant hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

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

    Date of amendment request: November 20, 1995
    Brief description of amendment: The proposed amendment revised the 
technical specifications to eliminate the response time testing 
requirements for selected Reactor Protection System Instrumentation.
    Date of issuance: January 11, 1996
    Effective date: January 11, 1996
    Amendment No.: 86
    Facility Operating License No. NPF-47. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62492) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated Janaury 11, 1996. No significant 
hazards consideration comments received. No.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, LA 70803

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

    Date of application for amendments: May 25, 1995 (AEP:NRC:1071T)
    Brief description of amendments: The amendments incorporate a 
cycle- and burnup-dependent peaking factor penalty in the Core 
Operating Limits Report and add an appropriate reference to the COLR 
and update the topical report reference in the Technical 
Specifications.
    Date of issuance: January 4, 1996
    Effective date: January 4, 1996, with full implementation within 45 
days
    Amendment Nos.: Unit 1, 206, Unit 2, 190
    Facility Operating License Nos. DPR-58 and DPR-74. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register:  The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated January 4, 1996. No significant hazards consideration 
comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: October 3, 1995
    Brief description of amendment: The amendment removes the Limiting 
Condition for Operation (LCO) and Surveillance Requirements for the 
loss-of-normal power (LNP) trip function from Tables 3.2.2 and 4.2.1 
and inserts new LCO 3.2.F and Surveillance Requirement 4.2.F. In 
addition, the amendment adds a new table to specify the required LNP 
instrumentation for each bus, updates the Table of Contents, makes some 
editorial changes, and revises the associated Bases section.
    Date of issuance: January 17, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 92
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 4, 1995 (60 FR 
62111) The Commission's related evaluation of the amendment is 
contained in a Safety evaluation dated January 17, 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.

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

    Date of application for amendment: September 11, 1995, as 
supplemented November 15, 1995.
    Brief description of amendment: The amendment changes Technical 
Specification Sections 3.4.8 and 3.9.9, Tables 2.2-1, 3.3-3, 3.3-5 and 
3.3-8, and Bases Sections 3/4.2.1, 3/4.4.8 and 3/4.11.2.1. These 
changes combine several different administrative changes which will 
correct typographical errors, provide clarifications, or make editorial 
changes.
    Date of issuance: January 17, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 194 
    
[[Page 3507]]

    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 11, 1995 (60 FR 
52933) The November 15, 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 January 17, 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.

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

    Date of application for amendment: September 29, 1995, as 
supplemented November 9, 1995.
    Brief description of amendment: The amendment provides three 
changes to the Technical Specifications (TS) relating to the 
pressurizer safety valves (PSV) and the main steam safety valves 
(MSSV).
    The first change is to TS 3.4.2.1 and 3.4.2.2 and involves relaxing 
the as-found setpoint tolerance for the pressurizer safety valves 
(PSVs) and the main steam safety valves (MSSVs) from the current value 
of plus or minus 1% to plus or minus 3%. Table 4.7-1 is also modified 
to correct the as-found tolerance for the MSSV from plus or minus 1% to 
plus or minus 3%. Notes are added to TS 3.4.2.2 and Table 4.7-1 which 
specify that the lift setting should be determined at nominal operating 
conditions and should be set at plus or minus 1% of the lift setting.
    For the second change, Surveillance Requirement 4.7.1.1 and Table 
4.7-1 are modified to eliminate the need to verify the orifice size of 
each MSSV.
    The third change modifies the statement for TS 3.7.1.1 so that if a 
MSSV is inoperable and compensating action cannot be taken, the plant 
must be brought to hot shutdown (Mode 4) within 12 hours instead of 
cold shutdown (Mode 5) in 30 hours.
    Date of issuance: January 18, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 195
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 25, 1995 (60 FR 
54723) 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 January 18, 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.

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

    Date of application for amendments: November 30, 1995
    Brief description of amendments: The amendments change the 
technical specification requirements for control rod drive scram 
accumulator and charging water header minimum pressure.
    Date of issuance: January 11, 1996
    Effective date: Unit 2, as of date of issuance, to be implemented 
concurrently with Amendment 210, issued August 30, 1995; Unit 3, as of 
date of issuance, to be implemented concurrently with Amendment 214, 
issued August 30, 1995.
    Amendments Nos.: 211 and 216
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 8, 1995 (60 FR 
63073) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 1996 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendments: November 21, 1995
    Brief description of amendments: The amendments change the test 
pressure requirements for the high pressure coolant injection system 
and the reactor core isolation cooling system surveillance tests. The 
amendments also change Section 5.5.7 of the technical specifications to 
eliminate reference to a section which was previously eliminated.
    Date of issuance: January 11, 1996
    Effective date: As of date of issuance.
    Amendments Nos.: 212 and 217
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 5, 1995 (60 FR 
62271) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 11, 1996 No significant 
hazards consideration comments received: No
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

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

    Date of application for amendments: December 19, 1995
    Brief description of amendments: These amendments change the 
ventilation filter test program bypass and penetration leakage test 
acceptance criteria from less than 0.05 percent to less than 1.0 
percent. The change corrects an administrative error that occurred 
during the development of the Peach Bottom Improved Technical 
Specifications which were issued as Amendments 210 and 214 to the Peach 
Bottom licenses on August 30, 1995.
    Date of issuance: January 16, 1996
    Effective date: Unit 2, effective as of date of issuance, to be 
implemented concurrently with Amendment 210, issued August 30, 1995; 
Unit 3, effective as of date of issuance, to be implemented 
concurrently with Amendment 214, issued August 30, 1995.
    Amendments Nos.: 213 and 218
    Facility Operating License Nos. DPR-44 and DPR-56: The amendments 
revised the Technical Specifications. Public comments requested as to 
proposed no significant hazards consideration: Yes (60 FR 66997, 
December 27, 1995). That notice provided an opportunity to submit 
comments on the Commission's proposed no significant hazards 

[[Page 3508]]
consideration determination. No comments have been received. The notice 
also provided for an opportunity to request a hearing by January 26, 
1996, but indicated that if the Commission makes a final no significant 
hazards consideration determination any such hearing would take place 
after issuance of the amendment. The Commission's related evaluation of 
the amendments, finding of exigent circumstances, and final 
determination of no significant hazards consideration are contained in 
a Safety Evaluation dated January 16, 1996
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.

    Dated at Rockville, Maryland, this 23rd day of January 1996.

    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation.
[Doc. 96-1683 Filed 1-30-96; 8:45 am]
BILLING CODE 7590-01-F