[Federal Register Volume 59, Number 244 (Wednesday, December 21, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-31196]


[[Page Unknown]]

[Federal Register: December 21, 1994]


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NUCLEAR REGULATORY COMMISSION
 

Biweekly Notice

Applications and Amendments to Facility Operating LicensesInvolving 
No Significant Hazards Considerations

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from November 28, 1994, through December 9, 1994. 
The last biweekly notice was published on December 7, 1994.

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 January 13, 1995, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    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.

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

    Date of amendment requests: November 2, 1994
    Description of amendment requests: The proposed amendment would 
delete the Condenser Vacuum Exhaust release point reference on Figure 
5.1-3 and combine it with the Plant Vent Exhaust release point on the 
revised Figure 5.1-3. In addition to the figure change, Bases Section 
3/4.3.3.6 is amended to note the deletion of radiation monitor RU-142 
and the relocation of RU-144 and RU-146 from Table 3.3-13 (previously 
deleted) to the Offsite Dose Calculation Manual (ODCM).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensees have 
provided their analysis about the issue of no significant hazards 
consideration, which is presented below:
    Standard 1--Does the proposed change involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Combining the condenser vacuum and the plant vent exhausts has 
no affect
    [sic] on the operation of the radiation monitoring system or its 
intended functions. Routing of the condenser vacuum exhaust to the 
plant vent exhaust is in the same area as the old system and does 
not affect accident initiation or consequences. The change has no 
affect
    [sic] on the operation of the plant. The radiation monitors 
affected by this change do not provide engineered safety features or 
protection system actuation signals. Therefore, the change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Standard 2 -- Does the proposed change create the possibility of 
a new or different kind of accident from any accident previously 
evaluated?
    The overall system is designed to assist the operators in 
evaluating and controlling the radiological consequences of normal 
plant operations, anticipated operational occurrences, and 
postulated accidents. The change does not affect the way the system 
is operated. Therefore, the change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Standard 3 -- Does the proposed change involve a significant 
reduction in a margin of safety?
    Combining of the condenser vacuum and plant vent exhaust into a 
single release path does not involve a significant reduction in a 
margin of safety. The change involves the removal of one high range 
monitor in the condenser vent, however, its function is provided by 
the high range monitor in the plant vent. The ranges of the monitors 
are the same. The existing plant effluent radiation monitors will 
serve to monitor both the plant and condenser air removal system 
effluent. The normal range monitors have the ability to adequately 
detect radiation over five decades and these monitors will stay in 
place and they have the ability to perform the anticipated radiation 
release detection. Therefore, the proposed change does not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
that 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: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    Basis for proposed no significant hazards consideration 
determination: I11Attorney for licensees: 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: Theodore R. Quay

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

    Date of amendments request: November 16, 1994 Description of 
amendments request: The proposed Technical Specification (TS) change 
would (1) revise TS 4.6.1.2 by removing the schedular requirements for 
Type A overall integrated leakage rate tests to be performed at 40 plus 
or minus 10 month intervals and replacing the acceptance criteria for 
these Type A integrated leakage rate tests with a reference to the 
containment integrated leakage testing requirements of Appendix J to 10 
CFR Part 50, (2) delete TS 4.6.1.2.a through TS 4.6.1.2.c because they 
are no longer needed, (3) revise TS 4.6.1.2.h to remove the prohibition 
against applying TS 4.0.2 to the 40 plus or minus 10 month integrated 
leakage rate test frequency, (4) delete Unit 1 one-time footnote * 
located on TS page 3/4 6-3A and on Table 4.6.1.2-1 listed on TS page 3/
4 6-3B since the exception provision has expired, (5) delete Unit 1 
one-time footnote ** located on TS page 3/4 6-3A since the exception 
has expired, (6) delete Unit 2 footnote * located on TS page 3/4 6-3 
because the exception constitutes an approved exemption.
    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 amendments do not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed amendments remove the detailed technical and 
schedular information pertaining to primary containment integrated 
leakage rate testing from the Technical Specifications and 
references the corresponding requirements that are located in the 
Appendix J to 10 CFR Part 50. As such, the proposed amendments are 
an administrative change since the actual requirements for the 
performance of primary containment integrated leakage rate testing 
are not being changed. No safety-related equipment, safety function, 
or plant operations will be altered as a result of the proposed 
amendments. The change does not affect the design, materials, or 
construction standards of the primary containment nor the test 
methods, test acceptance criteria, or testing frequencies applicable 
to primary containment integrated leakage rate testing. Based on the 
above, the proposed license amendments do not create a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed amendments would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. As stated above, no safety-related equipment, safety 
function, or plant operations will be altered as a result of the 
proposed change. The proposed amendments do not change the primary 
containment design or the test methods, test acceptance criteria, or 
testing frequencies for primary containment integrated leakage rate 
testing. As such, the proposed license amendments cannot create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendments do not involve a significant 
reduction in the margin of safety. The proposed amendments do not 
involve any changes to the test methods, acceptance criteria, or 
testing frequency for primary containment integrated leakage rate 
testing. Thus, the proposed amendments will not affect the ability 
of the primary containment to perform its intended safety function 
and no margins of safety, as defined by the plant's accident 
analyses, are impacted. Primary containment integrated leakage rate 
testing will continue to be performed in accordance with the 
regulatory requirements of Appendix J to 10 CFR Part 50. Based on 
the above reasoning, the proposed license amendments 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: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: R. E. Jones, General Counsel, Carolina Power 
& Light Company, Post Office Box 1551, Raleigh, North Carolina 27602
    NRC Project Director: William H. Bateman

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

    Date of amendment request: September 28, 1994
    Description of amendment request: Amendments will update the ``Loss 
of Power'' functional unit of the Engineered Safety Features Actuation 
System (ESFAS) Instrumentation tables within the Technical 
Specifications for McGuire Nuclear Station.
    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:
    As required by 10 CFR 50.91, this analysis is provided 
concerning whether the requested amendments involve significant 
hazards considerations, as defined by 10 CFR 50.92. Standards for 
determination that an amendment request involves no significant 
hazards considerations are if operation of the facility in 
accordance with the requested 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 requested amendments update the existing one-level 
undervoltage protection to be exclusively for loss of voltage, and 
add a second level of undervoltage protection to be exclusively for 
degraded voltage.
    In 48 FR 14870, the Commission has set forth examples of 
amendments that are considered not likely to involve significant 
hazards considerations. Example vi describes a change which either 
may result in some increase to the probability or consequences of a 
previously-analyzed accident or may reduce in some way a safety 
margin, but where the results of the change are clearly within all 
acceptable criteria with respect to the system or component 
specified in the Standard Review Plan. The requested amendments are 
similar to example vi in that they result in some increase to the 
probability of a previously-analyzed accident, the Loss of Offsite 
Power accident, but where the changes are clearly based on the 
recommendations of Branch Technical Position PSB-1.
    Criterion 1
    The requested amendments will not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The requested amendments will involve some increase in 
the probability of an accident previously evaluated. Automatic 
separation from offsite power (a LOOP accident) will be more 
probable because the voltage setpoints for the new relaying will be 
higher than the settings for the existing relaying. The closer relay 
settings are to 100% bus voltage, the more frequently actual bus 
voltage can be expected to occur at or below the setpoint. The 
occurrence of a LOOP presents a challenge to safety systems. More 
probable (e.g., more frequent) LOOPs increase the frequency of 
safety system challenges, which increases the probability of 
malfunction of equipment important to safety. However, offsetting 
this probability increase is a probability decrease due to the 
protection of safety equipment from degraded voltage conditions, 
given by the added protective relaying. The EPC system is required 
to provide power for equipment used for accident mitigation and safe 
shutdown. The ability of the EPC system to perform its required 
safety functions will not be degraded by the implementation of this 
TS change. No common failure modes are created between redundant EPC 
system power trains. Therefore, the consequences of an accident or 
malfunction of equipment important to safety evaluated in the SAR 
are not increased.
    Criterion 2
    The requested amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated. No new failure modes are created by the implementation of 
this TS change. No accidents previously considered incredible are 
made credible. The added protective relaying is expected to be as 
reliable as the existing relaying. The added equipment is QA 
Condition 1, and qualifications of equipment enclosures have been 
maintained. Thus, the possibility of an accident or malfunction of 
equipment of a different type than evaluated in the SAR will not be 
created.
    Criterion 3
    The requested amendments will not involve a significant 
reduction in a margin of safety. The setpoints for the existing Loss 
of Power protective relays are lowered by this TS change. The new 
setpoints have been evaluated and will not prevent the protective 
relaying from performing its required safety function. The fission 
product barriers (RCS pressure boundary, containment, fuel pellets, 
and cladding) are not degraded. No assumptions made in any accident 
analysis are affected by the implementation of this TS change, 
except as previously discussed for probability of a Loss of Offsite 
Power. Therefore, the margin of safety as defined in the basis for 
any Technical Specification is not decreased.
    Based on the preceding analyses, Duke Power concludes that the 
requested amendments do not involve a significant hazards 
consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

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: August 31, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specifications (TSs), Section 6, Administrative 
Controls, and includes line-item improvements suggested by Generic 
Letter 93-07. The proposed changes include the following:
    1. Elimination of the references to specific frequencies for 
each of the Technical Specification required audits.
    2. Elimination of the references to reviews and audits of the 
Emergency Plan and Security Plan.
    3. Separation of the Inservice Inspection (ISI) and Inservice 
Testing (IST) Programs surveillance requirements and removal of the 
requirement that relief requests be granted before they are 
implemented for both IST and ISI.
    4. Editorial changes which were necessitated by a 
reorganization.
    5. Elimination of the reference to Appendix A of 10 CFR Part 55.
    6. Elimination of the requirement to perform an independent fire 
protection and loss prevention program inspection annually.
    7. Inclusion of the Offsite Dose Calculation Manual and Process 
Control Program and associated implementing procedures into the list 
of required audits.
    8. Updates of the Beaver Valley Power Station (BVPS) Unit 2 
License Conditions to reflect completion of activities.
    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 likelihood that an accident will occur is neither increased 
or decreased by this proposed Technical Specification change which 
only affects review and audit frequencies, removes redundancies in 
the audit program, corrects editorial information, and updates the 
Unit 2 license conditions. This Technical Specification change will 
not impact the function or method of operation of plant equipment. 
Thus, there is not a significant increase in the probability of a 
previously analyzed accident due to this change. No systems, 
equipment, or components are affected by the proposed change. Thus, 
the consequences of a malfunction of equipment important to safety 
previously evaluated in the Updated Final Safety Analysis Report are 
not increased by this change.
    The proposed change affects audit frequencies, types of audits 
listed in the technical specifications, references for some 
technical specification sections, the time frame for Inservice 
Testing (IST) and Inservice Inspection (ISI) relief request 
submittals, and editorial changes necessitated by an internal 
reorganization. As such, the proposed change has no impact on 
accident initiators or plant equipment, and therefore, does not 
affect the probabilities or consequences of an accident.
    Therefore, the proposed change will not involve a significant 
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?
    The proposed technical specification revisions do not involve 
changes to the physical plant or operations. Since program audits, 
organizational titles, and technical specification references do not 
contribute to accident initiation, a change related to the areas 
listed in the description section [***] cannot produce a new 
accident scenario or produce a new type of equipment malfunction. 
Therefore, this change does not alter any existing accident 
scenarios. The proposed change does not affect equipment or its 
operation.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The proposed change concerns the conduct of audits, technical 
specification references, ISI and IST relief request submittals, 
completed License conditions, and organizational title changes and 
does not directly affect plant equipment or operation. Safety limits 
and limiting safety system settings are not affected by this 
proposed change.
    Therefore, use of the proposed Technical Specification would 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 involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Alquippa, Pennsylvania 15001.
    Attorney for licensee: Gerald Charnoff, Esquire, Jay E. Silberg, 
Esquire, Shaw, Pittman, Potts & Trowbridge, 2300 N Street, NW., 
Washington, DC 20037.
    NRC Project Director: Walter R. Butler

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

    Date of amendment request: August 11, 1994, as supplemented 
December 2, 1994.
    Description of amendment request: The requested change would remove 
cycle-specific variables from the Waterford 3 Technical Specifications 
(TSs) and control them under a new document called the Core Operating 
Limits Report (COLR). All cycle-specific limits that are to be included 
in the COLR must be calculated using NRC approved methodologies. The 
proposed change is consistent with the TS line-item improvement 
guidelines provided by the NRC in Generic Letter (GL) 88-16, ``Removal 
of Cycle-Specific Parameter Limits From Technical Specifications,'' 
dated October 3, 1988.
    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. Removing cycle-specific variables from the TS and placing 
them into a COLR, is consistent with the NRC guidance provided in GL 
88-16. These changes are administrative in nature and have no impact 
on plant operation or accident analyses. The TS will continue to 
require operation within the core operational limits for each cycle 
reload calculated by the approved reload methodologies. If these 
limits are violated, Technical Specifications will continue to 
ensure that the appropriate actions are taken.
    The cycle-specific evaluation demonstrates that changes in the 
fuel cycle design and the corresponding COLR do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    Listing the NRC approved methodologies in the COLR as opposed to 
the TS Administrative Controls section is purely an administrative 
change in contrast to NUREG 1432. The proposed change requires the 
use of NRC approved methodologies. Listing the approved 
methodologies in the TS provides the potential for an increased 
licensee and NRC administrative burden without a commensurate 
increase in safety or control.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed changes, to relocate the cycle-specific 
variables from TS to the COLR, are administrative in nature. No 
change in the design, configuration, or method of operation of the 
plant is made by this amendment. The cycle-specific variables will 
continue to be calculated using NRC approved methods. TS will 
continue to require operation within the required core operating 
limits and appropriate actions will be taken if the limits are 
exceeded.
    Listing the NRC approved methodologies in the COLR as opposed to 
the TS Administrative Controls section is purely an administrative 
change in contrast to NUREG 1432. The proposed change requires the 
use of NRC approved methodologies. Listing the approved 
methodologies in the TS provides the potential for an increased 
licensee and NRC administrative burden without a commensurate 
increase in safety or control.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    3. The margin of safety presently provided is not affected by 
removing cycle-specific core operating limits from TS. The core 
limits contained in the COLR are obtained through analyses using NRC 
approved methodologies. The TS still: (1) require that the core be 
operated within these limits and (2) specify appropriate actions to 
be taken if the limits are violated. The cycle-
    specific COLR limits for future reload will also be developed 
based on NRC-approved methodologies. In addition, each reload will 
involve a 10CFR 50.59 safety review to assure that operation of the 
unit within the cycle-specific limits will not involve a reduction 
in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Safety and Significant Hazard Determination
    Based on the above safety analysis, it is concluded that: (1) 
the proposed change does not constitute a significant hazards 
consideration as defined by 10CFR50.92; and (2) there is a 
reasonable assurance that the health and safety of the public will 
not be endangered by the proposed change; and (3) this action will 
not result in a condition which significantly alters the impact of 
the station on the environment as described in the NRC Final 
Environmental Statement.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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 amendment request: October 13, 1994
    Description of amendment request: The proposed amendments would 
revise the Hatch Technical Specifications (TS) as follows:
    1. Lower the anticipated transient without scram-recirculation pump 
trip (ATWS-RPT) setpoint by approximately 2 feet 2 inches to minimize 
the potential for recirculation pump trips following reactor scrams.
    2. Allow restarting the recirculation pump following an RPT when 
the temperature differential between the coolant at the reactor bottom 
head and the reactor steam dome cannot be obtained, provided certain 
conditions are met.
    The licensee believes the above changes will aid in preventing 
thermal stratification and unnecessary thermal cycles resulting from 
the rapid cooldown of the bottom head region and the reduction in 
reactor pressure to atmospheric conditions.
    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:
    Proposed Change 1
    Proposed Change 1 does not involve a significant hazards 
consideration, because it does not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Lowering the ATWS-RPT trip will not increase the probability of 
occurrence of any design basis accident or transient, since this 
change does not physically affect any component of the reactor 
coolant pressure boundary (RCPB). Therefore, the probability of a 
LOCA event is not increased. Lowering the ATWS-RPT water level 
setpoint does not increase the probability of an ATWS event, since 
no component of the CRD system or the reactor protection system is 
being physically altered by this change. Also, the operation of 
these two systems is not affected.
    Reducing the setpoint may require installation of new slave trip 
units; however, this addition does not increase the probability of 
occurrence of accidents or transients. The new trip units will be 
functionally identical to other slave trip units already in use at 
Plant Hatch and are within the design capabilities of ATTS. In 
conclusion, no safety-related plant system or component is being 
affected in a manner that would render it more susceptible to 
failure.
    Lowering the setpoint does not result in an increase of the 
consequences of a previously evaluated accident. GE reviewed the 
proposed reduction and determined the results of the ATWS event with 
the lowered setpoint remain acceptable. An approved analytical 
method (REDY) was used to evaluate a bounding ATWS event -- LOFW 
[loss of feedwater]. The results indicate that reactor power with 
the new ATWS-RPT setpoint remains stable, with no unacceptable power 
spikes. Hot and cold reactor shutdowns can still be ultimately 
attained.
    The consequences of non-ATWS events are not increased. For LOCA 
events, reducing the recirculation pump low water level trip 
setpoint allows the recirculation pumps to run longer. The forced 
circulation provided by the recirculation pumps keeps the fuel 
cooler for a longer period of time. The ECCS-LOCA analysis assumes 
the pump trip and coastdown early in the event. Therefore, lowering 
the ATWS-RPT makes the ECCS-LOCA analysis more conservative and, as 
a result, it does not need to change.
    Based on the above discussion, Proposed Change 1 does not 
constitute an increase in the probability or consequences of a 
previously analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated.
    Lowering the ATWS-RPT trip will not alter the design or 
operation of any safety-related system. The change may require 
adding new slave trip units to ATTS; however, the new trip units 
will be functionally identical to the equipment already in use at 
Plant Hatch. Furthermore, the addition of this slave trip unit is 
within the design capabilities of ATTS.
    Since no new operation modes, accident scenarios, or failure 
modes are introduced, Proposed Change 1 does not create the 
possibility of a new type of accident.
    3. Involve a significant reduction in the margin of safety.
    As stated previously, reducing the ATWS-RPT low water level 
setpoint will not cause unacceptable results for ATWS events. 
Specifically, the LOFW event is bounding for all the ATWS events. An 
evaluation using approved analytical methods indicates that reducing 
the ATWS-RPT setpoint will not result in power instabilities or 
unacceptable power spikes, or prevent the mitigation of an ATWS 
event. (Reference Enclosure 1 [of the licensee's submittal], 
Proposed Change 1).
    The ATWS-RPT aids in maintaining the level above the top of the 
active fuel. The reduction of core flow reduces the neutron flux and 
thermal power and, therefore, the rate of coolant boil-off. However, 
the setpoint reduction does not significantly reduce the margin of 
safety since a substantial margin remains to the top of the active 
fuel.
    For non-ATWS events, delaying the RPT will provide a slight 
improvement in the current ECCS-LOCA analysis, thereby improving the 
margin of safety.
    The margin of safety for transients is not reduced because plant 
transient (MCPR) analyses do not take credit for the ATWS-RPT trip.
    Proposed Change 2
    Proposed Change 2 does not involve a significant hazards 
consideration, because it does not:
    1. Involve a significant increase in the probability of 
occurrence or the consequences of a previously analyzed accident.
    Allowing a recirculation pump restart within 30 minutes of a 
trip, when the temperature differential is unknown, will not 
increase the probability of occurrence of a previously analyzed 
accident because this change does not physically alter the RCPB. 
Additionally, the proposed change does not alter the design or 
function of any safety-related systems.
    Furthermore, no recirculation system equipment is being changed 
as a result of this amendment. The start circuitry and trip 
circuitry remain[s] unaffected. Operation of the recirculation 
system with the reactor at power is also unaffected. As a result, 
the probability of the chapter 14 and 15 events dealing with the 
recirculation system are not increased; i.e., trip of one or both 
recirculation pumps, recirculation pump seizure, recirculation flow 
controller failure, etc.
    The purpose of the 145 deg.F temperature differential 
requirement is to avoid thermal shock caused by hot water on the 
cold CRD stub tubes during recirculation pump restart. If the 
temperature differential is unable to be determined, restart within 
30 minutes of the trip will not increase the probability or severity 
of thermal fatigue on the stub tubes. As discussed in Enclosure 1, 
Basis for Proposed Change 2, stratification will not develop within 
a 30-minute period following pump trip, thus, the temperature 
differential will not exceed 145 deg.F. Additional caveats are 
provided to insure the required temperature differential is met. 
These involve certain conditions of ECCS injection, feedwater 
temperature, and drive flow.
    General Electric verified that this provision for recirculation 
pump restart will not affect any plant safety analysis, including 
radiological analysis. Therefore, the consequences of previously 
analyzed events are not increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed recirculation pump restart provisions do not 
introduce any new plant operating modes, accident scenarios, or 
equipment failure modes. All other requirements for recirculation 
pump restart; e.g., those addressing equipment protection and power 
oscillations, will continue to apply.
    3. Involve a significant reduction in the margin of safety.
    The 145 deg.F differential temperature requirement is in place 
to avoid thermal fatigue on the CRD stub tubes and the in-core 
housing welds. Allowing the early restart with the listed caveats, 
when temperature indication is not available, is acceptable because 
the conditions for re-start insure that a stratified condition has 
not yet developed. Thus, the cooler vessel structures at the vessel 
bottom will not experience a severe thermal shock resulting from 
exposure to hot water following the pump restart.
    This change will actually aid in preventing the development of a 
stratified condition, since the recirculation pumps will be 
restarted before a stratified condition can develop, thereby helping 
to maintain RCPB integrity. In the past, it has often been necessary 
to depressurize the RPV [reactor pressure vessel] to atmospheric 
pressure before the required temperature differential was met. 
Proposed Change 2 should reduce the number of times depressurization 
is required, thus avoiding unnecessary thermal cycles on the RPV. 
Therefore, the margin of safety regarding the protection of RPV 
components from severe thermal stresses, and the integrity of the 
RCPB has not been reduced, and may actually increase.
    The margin of safety in existing plant analyses is not reduced, 
because none of the analyses are adversely affected as a result of 
allowing the pump restart within 30 minutes of the RPT, as indicated 
in GE's review of plant transient and accident analyses.
    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: Appling County Public Library, 
301 City Hall Drive, Baxley, Georgia 31513
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Herbert N. Berkow

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 (VEGP), Units 1 
and 2, Burke County, Georgia

    Date of amendment request: October 3, 1994
    Description of amendment request: This amendment would replace the 
reactor coolant system heatup and cooldown limitations for VEGP Units 1 
and 2, contained in Technical Specification figures 3.4-2a through 3.4-
3b, and the maximum allowable nominal power-operated relief valve 
(PORV) setpoint for the cold overpressure protection system. These 
changes are the results of new analyses that account for the 
nonconservatisms identified in NRC Information Notice 93-58, the 
results of reactor pressure vessel surveillance capsule examinations, 
and recently issued ASME Code Case N-514.
    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:
    Conformance of the proposed amendment with the standards for a 
determination of no significant hazards, as defined in the three 
factor test of 10 CFR 50.92, is shown in that the proposed 
amendment:
    1. Does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The revised heatup and cooldown limits and PORV setpoints ensure 
that the Appendix G pressure/temperature limits are not exceeded and 
therefore, help ensure that RCS integrity is maintained. The changes 
do not result in a condition where the design, material, and 
construction standards of the RCS are altered. In addition, the 
safety function of the COMS (cold over-pressure mitigation system), 
which is related to accident mitigation, has not been degraded. 
Therefore, the probability of an accident is not increased by the 
PORV setpoint change.
    The changes do not adversely affect the integrity of the RCS 
such that its function in the control of radiological consequences 
is affected. In addition, the changes do not affect any fission 
barrier. The changes do not degrade or prevent the response of the 
COMS or other safety-related system to accident scenarios, as 
described in FSAR chapter 15. In addition, the changes do not alter 
any assumption previously made in the radiological consequence 
evaluations nor affect the mitigation of the radiological 
consequences of an accident described in the FSAR. Therefore, the 
consequences of an accident previously evaluated in the FSAR will 
not be increased.
    Thus, operation of VEGP Units 1 and 2 in accordance with the 
proposed license amendment, does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The changes do not cause the initiation of any accident nor 
create any new credible limiting single failure for safety-related 
systems and components. The changes do not result in any event 
previously deemed incredible being made credible. As such, it does 
not create the possibility of an accident different than any 
evaluated in the FSAR.
    The changes do not have any effect on the ability of the safety-
related systems to perform their intended safety functions. The 
changes do not create failure modes that could adversely impact 
safety-related equipment. Therefore, it will not create the 
possibility of a malfunction of equipment important to safety 
different than previously evaluated in the FSAR. Thus, the proposed 
license amendment does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    3. Does not involve a significant reduction in a margin of 
safety.
    The evaluation has shown that the PORV setpoints ensure that the 
Appendix G pressure/temperature limits are not exceeded. The 
analysis to support the proposed PORV setpoint change demonstrates 
that the appropriate criteria, including that of ASME Code Case N-
514, are met for the postulated RCS pressures and temperatures. An 
adequate margin of safety against vessel failure is assured, in 
part, by the safety factors identified in Appendix G to Section III 
of the ASME Boiler and Pressure Vessel Code, and [in] the basis for 
ASME Code Case N-514 as well as [in the] added margin to prevent 
lifting of the PORVs. The heatup and cooldown limits are designed to 
prevent nonductile failure of the reactor vessel and take into 
account the results of surveillance capsule Y on the reactor vessel 
materials for VEGP Unit 1. The actuation of the safety-related 
components and responses of the safety-related systems will remain 
as modeled in the safety analyses. The changes will have no adverse 
[effect] on the availability, operability, or performance of the 
COMS. Therefore, the changes will not reduce the margin of safety, 
as described in the bases to any Technical Specification.
    Thus, [this] proposed license amendment 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: Burke County Public Library, 
412 Fourth Street, Waynesboro, Georgia 30830.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: January 14, 1994, as supplemented by 
letter dated November 10, 1994.
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TSs) by removing component lists 
from the TSs in accordance with NRC Generic Letter (GL) 91-08 and by 
removing the schedule for withdrawal of reactor vessel material 
specimen capsules from the TSs in accordance with GL 91-01. This 
proposed amendment was originally noticed in the Federal Register on 
May 23, 1994, (59 FR 26675). The licensee's letter dated November 10, 
1994, provides clarification of the wording in the proposed TSs and 
does not change the proposed determination that the amendment request 
involves no significant hazards consideration. However, the notice is 
being repeated here.
    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:
    Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed change will not result in any hardware or operating 
changes. The proposed change is based upon Generic Letters 91-01 and 
91-08 and merely removes component lists, removes details relating 
to the component lists, provides clarifying information supporting 
the removal of the component listings, or removes details (which are 
considered administrative) that are no longer applicable to the 
Technical Specifications. The components listed in the affected 
Technical Specifications are assumed in the mitigation of accident 
and transient events. The removal of tabular component listings from 
the Technical Specifications does not impact affected component 
OPERABILITY requirements. Technical Specifications will continue to 
require the components to be OPERABLE. Action statements and 
surveillance requirements for the components will also remain in the 
Technical Specifications. The tabular component lists are relocated 
to the Technical Requirements Manual which will be in accordance 
with the change control provisions specified in the Administrative 
Controls Section of the Technical Specifications (Specification 
6.5.2). Therefore, this change is administrative in nature and does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes to parameters governing normal plant 
operation. The proposed change will not impose any different 
requirements and adequate control of information will be maintained. 
No new failure modes are introduced. Therefore, this proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
    The proposed change will not reduce a margin of safety because 
it has no impact on any safety analysis assumption. The proposed 
changes do not alter the scope of equipment currently required to be 
OPERABLE or subject to surveillance testing, nor do the proposed 
changes affect any instrument setpoints or equipment safety 
functions. Therefore the change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Documents 
Department, Louisiana State University, Baton Rouge, Louisiana 70803
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 10, 1994
    Description of amendment request: The proposed amendment revises 
the Duane Arnold Energy Center (DAEC) Technical Specification (TS) 
Section 3.2.A to refer to the Offsite Dose Assessment Manual (ODAM) for 
the setpoint of the Offgas Stack Radiation Monitor and makes the 
``Applicable Operating Mode'' and the ``Action'' statements for these 
instruments consistent with the required function. The Action statement 
for the other instruments which initiate Secondary Containment 
isolation is also revised to be consistent with the current practice 
and with the function of those instruments. The Basis is also revised 
to add further description of the function and 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 provided below:
    1) The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because the instruments will still be required 
to be operable to initiate an isolation at a setpoint which will 
assure that the offsite dose limits are preserved, as designed, or 
else administrative controls will be established for the venting of 
primary containment. Through either means, offsite releases will be 
maintained within the limits established in the ODAM. The change to 
the applicable operating mode simply will require that the 
instruments be operable when they are assumed to be operable in 
previously analyzed accidents. The change to the required action 
when the TS requirement cannot be met will assure that the flow path 
from containment is isolated or that positive control is established 
so that any offsite radioactive gaseous release is within the limits 
analyzed in the ODAM.
    2) The proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated 
because the affected instruments are inputs to the secondary 
containment isolation and the revised specification will assure that 
they are operable or adequately compensated when they are assumed to 
perform their function. The instruments initiate a secondary 
containment isolation in the event that high radiation levels are 
detected in the monitored effluent.
    3) The proposed amendment will not involve a significant 
reduction in a margin of safety because the revised applicability 
statement will assure that the instruments are operable when they 
are required to perform their function. The proposed compensatory 
action allows administrative control of the isolation valves when 
the instruments are inoperable and it is necessary to continue 
venting. This allowance recognizes that venting is a controlled 
evolution and that operator action would be adequate to prevent 
excessive releases in the event of high radioactivity in the offgas 
piping. The revision to the setpoint will not affect system 
operation, but will continue to assure that the gaseous effluents 
released are within the limits specified in the ODAM.
    In summary, the proposed changes do not change the probability 
or consequences of an accident previously evaluated, do not create 
the possibility for a new or different kind of accident and do not 
involve a reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
thisreview, 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.
    Attorney for licensee: Jack Newman, Kathleen H. Shea, Newman, 
Bouknight & Edgar, PC, 1615 L Street, NW., Washington, DC 20036.
    NRC Project Director: Leif J. Norrholm

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: October 7, 1994
    Description of amendment requests: The proposed amendments would 
remove the requirements for the Nuclear Safety and Design Review 
Committee (NSDRC) to audit, and for the Plant Nuclear Safety Review 
Committee (PNSRC) to review, the Emergency and Security plans and 
implementing procedures. The composition of the PNSRC and the NSDRC 
would also be revised to reflect organizational changes. Changes would 
be made to the delegation of responsibility by the Site Vice President/
Plant Manager, and title corrections would be made on all pages 
affected by the above changes.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:We [the licensee] have 
evaluated the proposed T/S changes and have determined that the changes 
should involve no significant hazards consideration. The proposed 
amendment involves changes to the administrative controls section of 
the T/Ss only. Because all changes reflect organizational/title changes 
only or guidance from GL 93-07, they do 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.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Maud Preston Palenske 
MemorialLibrary, 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

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: November 16, 1994
    Description of amendment requests: The proposed amendments would 
allow core offload 100 hours after core subcriticality instead of the 
168 hours currently required. Also included in this submittal are minor 
typographical corrections to Figure 5.6-1, ``Normal Storage Pattern 
(Mixed Three Zone), and Figure 5.6-2, ``Interim Storage Pattern 
(Checkerboard).
    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:
    We [the licensee] have evaluated the proposed T/S, editorial and 
clarification changes and have determined that they do not represent 
a significant hazards consideration based on the criteria 
established in 10 CFR 50.92(c). Operation of Cook Nuclear Plant in 
accordance with the proposed amendment will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Although one of the proposed changes results in initiation of 
core offload earlier after subcriticality than is currently allowed, 
it does not increase the probability or consequences of an accident 
previously evaluated. The bulk pool water temperatures, fuel rod 
clad temperatures, and pool wall concrete temperatures will be 
within acceptable limits as shown in Attachment 2 [of the November 
16, 1994, submittal]. In addition, the subject change will not 
result in an uncontrolled release of radiation to the environment 
and will not initiate an accident. The remaining changes are 
editorial in nature and have no [e]ffect on probability or 
consequences of a postulated accident.
    (2) Create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    As previously stated, the earlier fuel movement change will not 
result in bulk pool water, fuel rod clad, or concrete temperatures 
which would initiate bulk pool boiling, challenge fuel rod integrity 
or jeopardize the structural integrity of the pool. This change will 
also have no impact on the criticality, structural, seismic, or 
dropped assembly accident analysis previously performed and accepted 
by the NRC. Consequently, the proposed T/S change does not create 
the possibility of a new or different kind of accident from any 
previously analyzed. The remaining changes have no [e]ffect on [the] 
nature or probability of a postulated accident.
    (3) Involve a significant reduction in a margin of safety.
    The proposed change for earlier fuel movement will not result in 
bulk pool water temperatures, fuel rod clad temperatures or concrete 
temperatures which would initiate bulk pool boiling, challenge fuel 
rod integrity or jeopardize the structural integrity of the pool. 
This proposed change will not affect the results of any other 
analysis associated with the spent fuel pool. It is, therefore, 
concluded that this change poses no significant reduction in a 
margin of safety. The remaining changes have no [e]ffect on the 
nature or probability of a postulated accident.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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

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: November 18, 1994
    Description of amendment requests: The license amendment requests 
propose a change to Technical Specification (T/S) 4.0.5 for both units 
to delete the wording ``except where specific written relief has been 
granted by the Commission pursuant to 10 CFR 50, Section 
50.55a(g)(6)(i).'' This change, which is consistent with guidance in 
the November 1993 draft NUREG-1482, ``Guidelines for Inservice Testing 
at Nuclear Power Plants,'' would allow the licensee to implement 
certain 10 CFR 50.55a relief requests while the relief requests were 
being reviewed by the NRC.
    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:
    We [the licensee] have evaluated the proposed T/S change and 
have determined that the change involves no significant hazards 
consideration. Operation of Cook Nuclear Plant in accordance with 
the proposed amendment will not:
    (1) Involve a significant increase in the probability or 
consequences of an accident.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The inspections required under Section XI are intended to 
show the operational readiness of the applicable components, and 
exceptions to the Code are allowed. When taking relief from Code 
requirements, alternate requirements are developed which provide a 
high level of confidence that components will perform their intended 
function.
    The proposed change does not alter the Code requirements or 
lessen our obligations under existing regulations. Its only effect 
is to allow implementation of Code relief prior to obtaining NRC 
written approval. The proposed T/S change is consistent with NUREG-
1431, and, as such, has been found to be acceptable by the NRC. 
Therefore, we believe that implementation of this change will not 
involve a significant increase in the probability or consequences of 
a previously analyzed incident.
    (2) Create the possibility of a new or different kind of 
accident from any previously analyzed.
    The proposed amendment does not create the possibility of a new 
or different kind of accident from any previously evaluated. Typical 
relief requests involve using alternative testing methods or 
increasing the time interval between tests. Each proposed 
alternative must assure that the component will perform its intended 
function. The proposed change involves no physical changes to the 
plant; therefore, we believe that implementation of this change will 
not introduce a new of different kind of accident than previously 
analyzed.
    (3) Involve a significant reduction in a margin of safety.
    The proposed amendment does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The inspections required under Section XI are intended to 
show the operational readiness of the applicable components, and 
exceptions to the Code are allowed. When taking relief from Code 
requirements, alternate requirements are developed which provide a 
high level of confidence that components will perform their intended 
function.
    The proposed change does not altar the Code requirements or 
lessen our obligations under existing regulations. Its only effect 
is to allow implementation of Code relief prior to obtaining NRC 
written approval. The proposed T/S change is consistent with NUREG-
1431, and, as such, has been found to be acceptable by the NRC. 
Therefore, we believe that implementation of this change will not 
result in a significant reduction of the margin of safety.
    NRC staff has reviewed the licensee's analysis and, based on this 
review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: 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

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

    Date of amendment request: November 14, 1994
    Description of amendment request: The proposed license amendment 
would revise Technical Specification 4.5.1.e.2.e) to reduce the leak 
rate test pressure for the Automatic Depressurization System (ADS) 
nitrogen receiving tanks from 385 psig to 365 psig. This pressure 
reduction would be made to reduce potential degradation of the rupture 
disk installed on each ADS nitrogen receiving tank during periodic leak 
testing of the receiving tanks. Plant operating experience has shown 
that leak rate testing at 385 psig occasionally results in inadvertent 
failure of the rupture disks. Testing at the reduced pressure would be 
consistent with the manufacturer's recommendations.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The ADS is required to effect or support the safe shutdown of 
the reactor. This is accomplished by the blowdown of steam from a 
group of seven designated ADS SRVs [safety/relief valves] to the 
suppression pool. The proposed change to the test pressure does not 
affect any accident precursors. Therefore, the proposed change 
cannot increase the probability of an accident previously evaluated.
    In the event the nitrogen gas supply from the nitrogen gas 
storage tanks is lost, a minimum nitrogen pressure of 334 psig in 
the ADS nitrogen receiver tanks assures a five-day supply of 
nitrogen to the ADS accumulators. The proposed change to 
Surveillance Requirement 4.5.1.e.2.e) would decrease the leak rate 
test pressure of the ADS nitrogen receiver tanks from 385 psig to 
365 psig. Since the proposed test pressure remains well above the 
design minimum pressure of 334 psig, the surveillance test continues 
to ensure that the actual leakage of the safety related ADS 
accumulator pneumatic supply system is bounded by the leakage 
assumptions contained in the system design. In addition, the 
surveillance test continues to ensure that the ADS nitrogen receiver 
tanks are capable of providing a 5-day supply of nitrogen to the ADS 
accumulators. Therefore, the proposed change does not significantly 
increase the consequences of a previously evaluated accident.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed leak rate test pressure of 365 psig for the ADS 
nitrogen receiver tanks is above the minimum design pressure to 
assure a 5-day supply of nitrogen is available to the ADS 
accumulators if makeup from the high pressure nitrogen gas storage 
tanks is lost. With the proposed change, the ADS will continue to 
perform its safety function of effecting and supporting the safe 
shutdown of the reactor. The nitrogen receiving tank test pressure 
is not a precursor for any new or different accident and the change 
does not affect the operation of the system in any way.
    Accordingly, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    The operation of the ADS SRVs, in conjunction with the LPCI [low 
pressure coolant injection] mode of RHR [residual heat removal 
system] and/or the LPCS [low pressure core spray] system, functions 
as an alternative to the HPCS [high pressure core spray system] for 
protection against fuel cladding damage upon a small break loss-of-
coolant accident. The blowdown of steam by these SRVs depressurizes 
the reactor, allowing injection by the low-pressure coolant 
injection sources. With the proposed change, the ADS will continue 
to perform its intended safety function of effecting and supporting 
the safe shutdown of the reactor as an alternate to the HPCS. The 
proposed test pressure of 365 psig remains well above the minimum 
acceptable pressure for the nitrogen receiver tanks of 334 psig. 
Therefore, the change 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Michael J. Case

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

    Date of amendment request: November 30, 1994
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications by adding a footnote to Limiting 
Conditions for Operation (LCOs) 3.8.1.1.b and 3.8.1.2.b which will 
denote that 24,000 gallons of fuel oil is capable of supporting the 
operation of one emergency diesel generator (EDG) for at least 4 days 
and the other EDG for 1 hour with the EDGs loaded to the continuous 
rated load of 2750 kW.
    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 because the changes would not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases 
Section 3/4.8 will revise the Millstone Unit No. 2 design 
requirements regarding the volume of EDG fuel oil which is required 
to be stored onsite. The new rationale indicates that 24,000 gallons 
of safety-related fuel oil would support the operation of one EDG 
for at least four days with the other EDG running for at least one 
hour. These run-times assume the EDGs are loaded to the continuous 
rated loading of 2750 kW.
    The proposed changes have no effect on EDG operation and 
reliability. They provide additional operational flexibility, 
because the EDG loading can be varied without the EDG minimum run-
time being altered. Also, an EDG run-time of at least four days 
provides significant time to replenish fuel oil from onsite and 
offsite sources even in the event of a hurricane or seismic event.
    Based on the above, there is no effect on the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The EDGs are required to operate in response to a loss of 
offsite power. The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b 
and Bases Section 3/4.8 do not change the manner in which the EDGs 
respond to a design basis accident. Also, the proposed changes do 
not introduce any new failure mechanisms. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. Involve a significant reduction in a margin of safety.
    The proposed changes to LCOs 3.8.1.1.b and 3.8.1.2.b and Bases 
Section 3/4.8 have no effect on EDG operation and reliability. They 
provide additional operational flexibility, because the EDG loading 
can be varied without the EDG minimum run-time being altered.
    An EDG run-time of at least four days provides significant time 
to replenish EDG fuel oil via onsite or offsite sources even in the 
event of a hurricane of seismic event. EPIP 4400 requires that the 
need to order EDG fuel oil be evaluated within four hours of a loss 
of offsite power event. Also, the high reliability of the electrical 
grid and the high probability that offsite power would be restored 
within 24 hours reduces the need to rely on extended EDG operation.
    Millstone Unit No. 2 has more margin than is indicated by the 
new design requirements. The EDG run-time will be significantly 
greater than four days, because the electrical loading on the EDGs 
will be less than the continuous rated loading, and electrical loads 
will be shed through normal recovery actions following a design 
basis accident.
    Based on the above, the proposed changes do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.
    Attorney for licensee: Ms. L. M. Cuoco, Senior Nuclear Counsel, 
Northeast Utilities Service Company, Post Office Box 270, Hartford, CT 
06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment requests: July 11, 1994
    Description of amendment requests: The proposed amendments would 
change license condition 2.C.4 of each license to conform to the 
standard fire protection license condition as stated in Generic Letter 
(GL) 86-10, ``Implementation of Fire Protection Requirements.'' In 
addition, the amendments would delete the fire protection program 
elements from the Technical Specifications and incorporate, by 
reference, the NRC-approved Fire Protection Program and major 
commitments, including the fire hazards analysis, into the Updated 
Safety Analysis Report. Guidance for these proposed changes is also 
provided in GL 88-12, ``Removal of Fire Protection Requirements from 
Technical Specifications.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) The proposed amendment[s] will not involve a significant 
increase in the probability or consequences of [an] accident 
previously evaluated.
    The requested changes are administrative in nature in that they 
move fire protection requirements from the Technical Specifications 
to the Fire Protection Program and associated implementing 
procedures following the guidance provided in GL 86-10 and GL 88-12. 
The requested changes will not revise the requirements for fire 
protection equipment operability, testing or inspections. The 
amendment would give added responsibility to the Operations 
Committee for review of the Fire Protection Program in accordance 
with the guidance given in GL 86-10 and 88-12 including special 
reporting requirements associated with limiting conditions for 
operation for fire protection systems.
    The proposed changes do 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 do they affect 
any assumptions or conditions in any of the accident analyses. Since 
the accident analyses remain bounding, their radiological 
consequences are not adversely affected.
    Therefore, the probability or consequences of an accident 
previously evaluated are not affected.
    (2) The proposed amendment[s] will not create the possibility of 
a new or different kind of accident from any accident previously 
analyzed.
    The requested changes are administrative in nature in that they 
move fire protection requirements from the Technical Specifications 
to the Fire Protection Program and associated implementing 
procedures following the guidance provided in GL 86-10 and GL 88-12. 
The requested changes will not revise the requirements for fire 
protection equipment operability, testing or inspections. The 
amendment would give added responsibility to the Operations 
Committee for review of the Fire Protection Program in accordance 
with the guidance given in GL 86-10 and 88-12 including special 
reporting requirements associated with limiting conditions for 
operation for fire protection systems.
    The proposed changes do 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.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be 
created.
    (3) The proposed amendment[s] will not involve a significant 
reduction in the margin of safety.
    The requested changes are administrative in nature in that they 
move fire protection requirements from the Technical Specifications 
to the Fire Protection Program and associated implementing 
procedures following the guidance provided in GL 86-10 and GL 88-12. 
The requested changes will not revise the requirements for fire 
protection equipment operability, testing or inspections. The 
amendment would give added responsibility to the Operations 
Committee for review of the Fire Protection Program in accordance 
with the guidance given in GL 86-10 and 88-12 including special 
reporting requirements associated with limiting conditions for 
operation for fire protection systems.
    Therefore, a significant reduction in the margin of safety would 
not be involved.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: John N. Hannon

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

    Date of amendment request: November 11, 1994
    Description of amendment request: The proposed amendment to the 
Technical Specifications (TSs) would make administrative changes to TS 
5.2 and 5.5. These changes reflect organizational changes in OPPD 
senior management, delete specific titles of personnel on the Safety 
Audit and Review Committee (SARC) and Plant Review Committee (PRC), and 
make changes to SARC reviews and audits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    (1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes are administrative changes to reflect 
organizational changes in Omaha Public Power District (OPPD) Senior 
Management, remove specific titles from the membership of the Plant 
Review Committee (PRC) and the Safety Audit and Review Committee 
(SARC), add minor clarifications to SARC reviews and audits and 
delete statements concerning the frequency of SARC audits from the 
Technical Specifications (TS).
    The proposed change to revise the overall corporate 
responsibility for plant nuclear safety from the Senior Vice 
President to Vice President is administrative in nature as it only 
reflects an organizational change. Section 12 of the Updated Safety 
Analysis Report describes the management structure and reporting 
responsibilities of OPPD. Section 12 provides an organizational 
chart to differentiate the Vice President in charge of nuclear 
activities from other Vice Presidents within OPPD. Therefore, 
changing the corporate reporting responsibility does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The proposed changes to the membership of the PRC and SARC are 
administrative in nature since only the specific titles of the 
members are being removed from the TS. The management level and 
expertise of personnel who are PRC or SARC members is not being 
changed. The review of plant operations is still required to be in 
compliance with ANSI N18.7-1976 and Regulatory Guide 1.33, Revision 
2, as committed to in the Fort Calhoun Station Quality Assurance 
(QA) Program. Any changes in the QA Program which reduce the 
effectiveness of the program must be approved by the NRC in 
accordance with 10 CFR 50.54(a)(3). Therefore, the proposed changes 
to the membership of the PRC and SARC do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Clarifications of SARC reviews and audits and the deletion of 
SARC audit frequencies from the TS are administrative changes. The 
audit frequencies are required by the NRC approved QA Program and 
any changes that could reduce the effectiveness of the QA Program 
must be approved by the NRC in accordance with 10 CFR 50.54(a)(3). 
Therefore, the clarifications and deletion of the specific audit 
frequencies do not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The proposed changes are administrative in nature to reflect 
organizational changes in OPPD Senior Management, remove specific 
titles from the membership of the PRC and SARC, provide minor 
clarifications of SARC reviews and audits and delete statements 
concerning the frequency of SARC audits from the TS. The proposed 
changes do not revise any equipment setpoints, change the manner in 
which any plant equipment is operated, or propose any new operating 
modes. Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (3) Involve a significant reduction in a margin of safety.
    The proposed changes revise organizational and administrative 
requirements contained within the Administrative Controls section of 
the TS. The proposed changes do not revise any equipment setpoints, 
change the manner in which any plant equipment is operated, or 
propose any new operating modes. 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: W. Dale Clark Library, 215 
South 15th Street, Omaha, Nebraska 68102
    Attorney for licensee: LeBoeuf, Lamb, Leiby, and MacRae, 1875 
Connecticut Avenue, NW., Washington, DC 20009-5728
    NRC Project Director: Theodore R. Quay

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

    Date of amendment request: October 28, 1994
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) for the two units would add reference 
20 (Unit 1) and reference 18 (Unit 2) to Section 
6.9.3.2 as ``PL-NF-90-001, Supplement 1, 'Application of Reactor 
Analysis Methods for BWR Design and Analysis: Loss of Feedwater Heating 
Changes and Use of RETRAN MOD 5.1', September 1994''. These changes 
would add changes to the methodology that the licensee is using to 
perform its nuclear fuel reload analysis for the two units.
    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:
    I. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Incorporation of these proposed minor changes into PP&L's NRC 
approved methodology for performing reload licensing analysis is 
considered to be an enhancement to the currently approved 
methodology. Upgrading of the RETRAN code allows for taking 
advantage of state-of-the-art technology, while utilization of the 
generic correlation for the LOFWH event supports consistency in 
licensing analysis performance. Results of incorporating these 
changes will not significantly increase the probability or the 
consequences of an accident previously evaluated.
    II. Create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    As stated above, the incorporation of these minor changes are 
considered enhancements, allowing PP&L to more efficiently and cost 
effectively continue to perform future reload licensing analysis. 
Therefore, the incorporation of these changes will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    III. Involve a significant reduction in a margin of safety.
    In addition to the extensive testing perform by EPRI, PP&L has 
performed its own comparison tests utilizing RETRAN MOD005.1 in 
place of MOD004 for four licensing transients that use the RETRAN 
code. Results of this comparison were essentially the same for both 
codes and support this proposed change. Also, the Loss of Feedwater 
Heating event is not a limiting event for establishing MCPR 
Operating Limits for Susquehanna. Therefore, the incorporation of 
these changes will have no impact on current safety margins, nor 
will they involve a significant reduction in the margin to 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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: October 28, 1994
    Description of amendment request: The proposed changes to the 
Technical Specifications (TS) for the two units would make a number of 
administrative changes. These would include changing the title of the 
positions of Superintendent of Plant to the Vice President-Nuclear 
Operations, and changing the title of Vice-President-Nuclear Operations 
to Senior Vice President-Nuclear for the listing of the assignment of 
certain duties in various sub-sections of Section 6.0 of the TS. Other 
proposed changes would be the deletion of a number of footnotes 
indicating times, dates, and events that are no longer applicable, the 
addition of a footnote to Section 6.5.1.2 indicating that the Station 
Duty Manager shall act as a PORC [Plant Operations Review Committee] 
chairman in the absence of the Vice President-Nuclear Operations, and 
the change of the Semiannual Radioactive Effluent Release Report to 
Annual Radioactive Effluent Release Report in Table 4.11.2.1.2-1 
footnote g.
    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:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposal to change the title of Superintendent of Plant to 
Vice President-Nuclear Operations and Vice President-Nuclear 
Operations to Senior Vice President-Nuclear (for certain duties) is 
administrative in nature and does not compromise the minimum 
qualifications or training required for these positions. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change for the removal of footnotes that reference 
periods of time, dates, and events that have since past is justified 
based on the fact that they are no longer applicable. Because 
operators must perform unnecessary applicability reviews on these no 
longer applicable footnotes, removing the footnotes decreases the 
potential for confusion and incorrect actions. Therefore, the 
proposed change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to add a footnote indicating the Station 
Duty Manager shall act as PORC chairman in the absence of the Vice 
President-Nuclear Operations will ensure continuous leadership of 
PORC and will enhance performance by ensuring a responsible 
individual is available during all shifts. Therefore, the proposed 
change does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Changing the Semiannual Radioactive Effluent Release Report to 
Annual Radioactive Effluent Release Report was previously approved 
in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating 
this change into footnote g of Table 4.11.2.1.2-1 will maintain 
accuracy and consistency. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposal to change the title of Superintendent of Plant to 
Vice President-Nuclear Operations and Vice President-Nuclear 
Operations to Senior Vice President-Nuclear (for certain duties) is 
administrative in nature and does not compromise the minimum 
qualifications or training required for these positions. Also, the 
change does not diminish the responsibilities or functions of these 
positions. Therefore, the proposed change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed change for the removal of footnotes that reference 
periods of time, dates, and events that have since past is justified 
based on the fact that they are no longer applicable. Because 
operators must perform unnecessary applicability reviews on them, 
removing the no longer applicable footnotes decreases the potential 
for confusion and incorrect actions, thereby enhancing the safe 
operation of Susquehanna SES. Therefore, the proposed change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed change to add a footnote indicating the Station 
Duty Manager shall act as PORC chairman in the absence of the Vice 
President-Nuclear Operations will ensure continuous leadership of 
PORC and will enhance performance by ensuring a responsible 
individual is available during all shifts. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Changing Semiannual Radioactive Effluent Release Report to 
Annual Radioactive Effluent Release Report was previously approved 
in Amendment 128 to Unit 1 and Amendment 97 to Unit 2. Incorporating 
this change into footnote g of Table 4.11.2.1.2-1 will maintain 
accuracy and consistency. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    For the reasons discussed in items I and II above, as well as 
the enclosed Safety Assessment, the proposed change does not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: October 28, 1994
    Description of amendment request: The amendment would delete the 
requirements for chlorine detection and the associated Bases from the 
Technical Specifications for each unit as a result of the removal of 
bulk quantities of gaseous chlorine from the Susquehanna Steam Electric 
Station. Specifically, Sections 3.3.7.8 and the associated Surveillance 
Requirements in Section 4.3.7.8 would be deleted. In addition, Bases 3/
4.3.7.8 would also be deleted.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Review of the various design basis accidents identified in 
Chapter 15 of the Susquehanna SES Final Safety Analysis Report 
(FSAR) concluded that none of these accidents are affected by 
deletion of the chlorine detection requirements from Technical 
Specifications. With the elimination of bulk quantities of gaseous 
chlorine from use at Susquehanna SES the probability of control room 
inhabitability due to a gaseous chlorine release has actually 
decreased. Therefore, this proposed change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change involves only the deletion of the chlorine 
detection system Technical Specifications based upon a plant 
modification to remove gaseous chlorine as a biocide from 
Susquehanna SES and replace it with a nonoxidizing biocide. The 
release of chlorine from an off-site source is bounded by Reg. Guide 
1.95 in that manual isolation capability for the control room 
ventilation system is acceptable. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    The proposed change would not alter the margins of safety 
provided in the existing FSAR analysis (Sections 2.2.3.1.3 and 6.4) 
for chlorine release events since the basis for the existing margin 
of safety, which are the Reg. Guide 1.95 requirements, are not 
altered by the change. As stated above, since gaseous chlorine is no 
longer used for open cooling water treatment at Susquehanna SES and 
since the nonoxidizing biocide is relatively nontoxic to humans, 
safety margin has actually increased. Therefore, the proposed change 
does not involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 11, 1994
    Description of amendment request: The amendment would extend the 
Main Turbine Valve surveillance test interval from a weekly basis to no 
greater than 92 days for all Main Turbine Stop, Control, and Combined 
Intermediate 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:
    I. This proposal does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed Technical Specification change to a quarterly 
turbine inlet valve surveillance test interval is based on 
maintaining the turbine missile generation probability within the 
NRC criteria as stated in Table 3.1 of NUREG-1048. This, combined 
with the NRC acceptable strike-and-damage probability as specified 
in NUREG-1048, will keep the probability of unacceptable damage to 
safety-related structures, systems, and components from turbine 
missiles acceptably low (i.e., <10-7) . Thus, the NRC 
acceptable risk rate of <10-7/yr. is not changed and there is 
no increase in the probability of an accident previously evaluated.
    The proposed Technical Specification change to the turbine inlet 
valve surveillance interval does not effect the sequence of events 
or the consequences of an accident previously evaluated. The 
surveillance interval does not affect the strike and damage scenario 
of an accident previously evaluated. Thus, the radiological 
consequences of an accident previously evaluated will not be 
increased.
    II. This proposal does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed Technical Specification change to the turbine inlet 
valve surveillance interval does not affect the surveillance test 
characteristics. There are no new surveillance testing requirements. 
Surveillance testing of these valves does not create the possibility 
for a new or different kind of accident from any accident previously 
evaluated.
    III. This change does not involve a significant reduction in a 
margin of safety.
    The proposed Technical Specification change to the turbine inlet 
valve surveillance interval is based on maintaining the same margin 
of safety as previously determined by the NRC and does not reduce 
the margin of safety. In fact, the reduction in the testing rate 
will reduce the potential for testing related transients, which have 
been credited with causing 18 reactor scrams in the period 1985 
through 1992.
    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: Osterhout Free Library, 
Reference Department, 71 South Franklin Street, Wilkes-Barre, 
Pennsylvania 18701
    Attorney for licensee: Jay Silberg, Esquire, Shaw, Pittman, Potts 
and Trowbridge, 2300 N Street NW., Washington, DC 20037
    NRC Project Director: John F. Stolz

Philadelphia Electric Company, Public Service Electric and Gas 
Company, Delmarva Power and Light Company, and Atlantic City 
Electric Company, Dockets Nos. 50-277 and 50-278, Peach Bottom 
Atomic Power Station, Units Nos. 2 and 3, York County, Pennsylvania 
Date of application for amendments: November 14, 1994

    Description of amendment request: The proposed changes relocate 
audit topics and frequencies, Nuclear Review Board review requirements 
and requirements associated with the independent Safety Engineering 
Group function from the Peach Bottom Atomic Power Station, Units 2 and 
3 Technical Specifications to licensee controlled documents.
    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 
because the changes relocate requirements from the TS to licensee 
controlled documents consistent with the NRC Final Policy Statement 
on TS Improvements. Any changes to the licensee controlled documents 
will be evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59 
as appropriate. Therefore, these changes will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2) The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously evaluated 
because the changes will not alter the plant or the manner in which 
the plant is operated. The changes will not involve a design change 
or introduce any new failure modes. The changes will not alter 
assumptions made in the safety analysis and licensing basis. 
Adequate control of information will be maintained. Therefore, these 
changes will not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    2) The proposed changes do not involve a significant reduction 
in a margin of safety because they have no impact on any safety 
analysis assumptions. The requirements to be transposed from the TS 
to licensee controlled documents are the same as the existing TS. 
Any future changes to licensee controlled documents will be 
evaluated in accordance with 10 CFR 50.54(a) or 10 CFR 50.59 as 
appropriate. Because the proposed changes are consistent with NUREG-
1433, as modified by approved generic change BWOG-09, and the change 
controls for proposed relocated details and requirements provide an 
equivalent level of regulatory authority, revising the TS to reflect 
the approved level of detail and requirements ensures no significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, (REGIONAL DEPOSITORY) Education 
Building, Walnut Street and Commonwealth Avenue, Box 1601, Harrisburg, 
Pennsylvania 17105.
    Attorney for licensee: J. W. Durham, Sr., Esquire, Sr. V. P. and 
General Counsel, Philadelphia Electric Company, 2301 Market Street, 
Philadelphia, Pennsylvania 19101
    NRC Project Director: John F. Stolz

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

    Date of amendment request: November 15, 1994 (TS 350)
    Description of amendment request: The proposed change would remove 
the frequency for each of the audits specified in the administrative 
controls section of the technical specifications (TS). The requirements 
to perform the audits would be retained, but the frequency for their 
performance would be controlled by a requirement to be added to the 
Nuclear Quality Assurance Plan. This would require that the audits 
listed in the TS be performed on a biennial frequency. In addition, the 
proposed change would remove the requirement to perform site 
Radiological Emergency Plan and Physical Security/Safeguard Contingency 
Plan reviews and audits from the TS, since these requirements presently 
exist in the respective Plans.
    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 concluded that operation of BFN units 1, 2, and 3 in 
accordance with the proposed change to the technical specifications 
does not involve a significant hazards consideration. TVA's 
conclusion is based on its evaluation in accordance with 10 CFR 
50.91(a)(1), of the three standards set forth in 10 CFR 50.92(c). 
TVA's conclusion is based on the following:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The likelihood that an accident will occur is neither increased 
or decreased by this Technical Specification change which only 
affects review and audit frequencies. This Technical Specification 
change will not impact the function or method of operation of plant 
equipment. Thus, there is not a significant increase in the 
probability of a previously analyzed accident due to this change. No 
systems, equipment, or components are affected by the proposed 
change. Thus, the consequences of a malfunction of equipment 
important to safety previously evaluated in the UFSAR are not 
increased by this change.
    The proposed change only affects review and audit frequencies. 
As such, the proposed change has no impact on accident initiators or 
plant equipment, and thus, does not affect the probabilities or 
consequences of an accident.
    Therefore, we conclude that this change does not significantly 
increase the probabilities or consequences of an accident.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve changes to the physical 
plant or operations. Since program audits do not contribute to 
accident initiation, a change related to audit functions cannot 
produce a new accident scenario or produce a new type of equipment 
malfunction. Also, this change does not alter any existing accident 
scenarios. The proposed change does not affect equipment or its 
operation, and, thus, does not create the possibility of a new or 
different kind of accident. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The proposed change concerning conduct of reviews and audits 
does not directly affect plant equipment or operation. Safety limits 
and limiting safety system settings are no affected by this proposed 
change.
    Therefore, use of the proposed Technical Specification would not 
involve any 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: Athens Public Library, South 
Street, Athens, Alabama 35611
    Attorney for licensee: General Counsel, Tennessee Valley Authority, 
400 West Summit Hill Drive, ET llH, Knoxville, Tennessee 37902
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: November 15, 1994 (TS 94-12)
    Description of amendment request: The proposed change would remove 
the frequency for each of the audits specified in the administrative 
controls section of the technical specifications (TS). The requirements 
to perform the audits would be retained, but the frequency for their 
performance would be controlled by a requirement to be added to the 
Nuclear Quality Assurance Plan. This would require that the audits 
listed in the TS be performed on a biennial frequency. In addition, the 
proposed change would remove the requirement to perform site 
Radiological Emergency Plan and Physical Security/Safeguard Contingency 
Plan reviews and audits from the TS, since these requirements presently 
exist in the respective Plans.
    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 standards used to arrive at a determination that a Technical 
Specification change request involves no significant hazards 
consideration are included in the Commission's regulations, 10 CFR 
50.92, which states that no significant hazards considerations are 
involved if the 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. Each standard is addressed as 
follows:
    1. Operation of the facility in accordance with the proposed 
technical specifications would not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The likelihood that an accident will occur is neither increased 
or decreased by the Technical Specification change which only 
affects review and audit frequencies. This Technical Specification 
change will not impact the function or method of operation of plant 
equipment. Thus, there is not a significant increase in the 
probability of a previously analyzed accident due to this change. No 
systems, equipment, or components are affected by the proposed 
changes. Thus, the consequences of a malfunction of equipment 
important to safety previously evaluated in the FSAR are not 
increased by this change.
    The proposed change only affects review and audit frequencies. 
As such, the proposed change has no impact on accident initiators or 
plant equipment, and thus, does not affect the probabilities or 
consequences of an accident.
    Therefore, we conclude that this change does not significantly 
increase the probabilities or consequences of an accident.
    2. Operation of the facility in accordance with the proposed 
technical specifications would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not involve changes to the physical 
plant or operations. Since program audits do not contribute to 
accident initiation, a change related to audit functions cannot 
produce a new accident scenario or produce a new type of equipment 
malfunction. Also, this change does not alter any existing accident 
scenarios. The proposed change does not affect equipment or its 
operation, and, thus, does not create the possibility of a new or 
different kind of accident. Therefore, the proposed change does not 
create the possibility of a new or different kind of accident.
    3. Operation of the facility in accordance with the proposed 
technical specifications would not involve a significant reduction 
in a margin of safety.
    The proposed change concerning conduct of reviews and audits 
does not directly affect plant equipment or operation. Safety limits 
and limiting safety system settings are no affected by this proposed 
change.
    Therefore, use of the proposed Technical Specification would not 
involve any reduction in the margin of safety.
    The NRC has reviewed the licensee's analysis and, based on 
thisreview, 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

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

    Date of amendment request: September 8, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 4.2.2.2, 4.2.2.4, and 6.9.19 to 
incorporate a penalty in the Core Operating Limit Report (COLR) to 
account for FQ increases greater than 2 percent between 
measurements.
    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 to the Technical Specifications do not 
involve a significant hazards consideration because operation of 
Callaway Plant in accordance with these changes would not:
    1) Involve a significant increase in the probability of 
occurrence or the consequences of an accident or malfunction of 
equipment important to safety previously evaluated in the safety 
analysis report. There is no increase in the probability of 
occurrence or the consequences of an accident. The removal of 
FQ(Z) penalty values from the Callaway Plant Technical 
Specifications and the creation of cycle-specific FQ(Z) values 
in the COLR has no influence or impact on the probability or 
consequences of any accident previously evaluated. The cycle-
specific FQ(Z) values, although not in Technical 
Specifications, will be followed in the operation of the Callaway 
Plant. The proposed amendment still requires exactly the same 
actions to be taken when or if FQ(Z) limits are exceeded as is 
required by current Technical Specifications.
    2) Create a possibility of a new or different kind of accident 
from any previously evaluated in the safety analysis report. There 
is no new type of accident or malfunction created and the method and 
manner of plant operation will not change. As stated earlier, the 
removal of the cycle-specific FQ(Z) value has no influence or 
impact, nor does it contribute in any way to the probability or 
consequences of an accident. No safety-related equipment, safety 
function, or plant operation will be altered as a result of this 
proposed change. The cycle-specific FQ(Z) values are calculated 
using NRC approved methods. The Technical Specifications will 
continue to require operation within the required FQ(Z) limits 
and appropriate actions will be taken when or if limits are 
exceeded.
    3) Involve a significant reduction in a margin of safety. This 
is based on the fact that no plant design changes are involved and 
the method and manner of plant operation remains the same. The 
margin of safety is not affected by change and removal of FQ(Z) 
penalty values from the Technical Specifications. The margin of 
safety presently provided by current Technical Specifications 
remains unchanged. The current FQ(Z) limits remain unchanged 
and the current safety analysis limits remain valid and unaffected 
by this change. The proposed amendment continues to require 
operation within the core limits as obtained from the NRC approved 
design methodology and appropriate actions to be taken when or if 
FQ(Z) limits are violated remain unchanged.
    Given the above discussions as well as those presented in the 
Safety Evaluation, the proposed change does not adversely affect or 
endanger the health or safety of the general public or involve a 
significant safety hazard.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: Leif J. Norrholm

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

    Date of amendment request: September 12, 1994
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3.7.1.1 (Tables 3.7-1 and 3.7-2). Tables 
2.2-1 and 3.3-2, and Bases 3/4.7. Tables 3.7-1, 3.7-2, 3.3-2, and 2.2-1 
would be revised to provide appropriate margin to relax main steam line 
safety valve setpoint tolerance. Bases 3/4.7 would be revised to 
incorporate the methodology used to determine the maximum allowable 
power level associated with inoperable main steam line safety 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:
    The proposed changes to the Technical Specifications do not 
involve a significant hazards consideration because operation of 
Callaway Plant in accordance with these changes would not:
    1. Involve a significant increase in the probability of 
occurrence or the consequences of an accident or malfunction of 
equipment important to safety previously evaluated in the safety 
analysis report. The main steam line safety valves are designed to 
mitigate transients by preventing overpressurization of the main 
steam system. The proposed change does not alter this design basis. 
The revised analysis shows that the probability or
    consequences of all previously analyzed accidents are not 
changed by increasing the setpoint tolerance of the safety valves. 
Therefore, there is no increase in the probability of occurrence or 
the consequences of any accident.
    2. Create the possibility of a new or different kind of accident 
from any previously evaluated in the safety analysis report. There 
is no new type of accident or malfunction created, the method and 
manner of plant operation will not change nor is there a change in 
the method in which any safety related system performs its function. 
Any main steam safety valve lifting at the extremes of the proposed 
tolerance will not result in low lift setpoint that is less than the 
normal no load system pressure or a high lift setpoint that allows 
main steam system overpressurization.
    3. Involve a significant reduction in a margin of safety. This 
is based on the fact that no plant design changes are involved and 
the method and manner of plant operation remains the same. With the 
increased setpoint tolerance, the main steam safety valves will 
still prevent pressure from exceeding 110 percent of design pressure 
in accordance with the ASME code. All FSAR accident analysis 
conclusions remain valid and unaffected by this change.
    Given the above discussions as well as those presented in the 
Safety Evaluation, the proposed change does not adversely affect or 
endanger the health or safety of the general public or involve a 
significant safety hazard.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, DC 20037
    NRC Project Director: Leif J. Norrholm

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

    Date of amendment request: July 9, 1993, with supplemental 
information provided October 8, 1993, October 25, 1993, January 6, 
1994, February 2, 1994, May 3, 1994, May 13, 1994, September 26, 1994, 
and October 12, 1994.
    Description of amendment request: The proposed amendment would 
modify the operating license and several Technical Specifications (TS) 
to allow an increase in licensed power level by 4.9%. This would raise 
the licensed power level from the current 3323 MWt to 3486 MWt, but 
will not affect the basic fuel design or fuel operating limits. The 
uprate in power will be accomplished by expanding the existing power-
to-flow map to allow an increase in core flow along the flow control 
lines, with an associated increase in core power. The increased flow 
and power will also cause an increase in operating reactor vessel steam 
dome pressure that will require an associated TS change.
    Other TS changes proposed to reflect necessary modifications to 
address the power uprate are (1) an increase in the average power range 
monitor (APRM) and flow biased scram and rod block monitor (RBM) 
setpoints, (2) an increase in reactor steam pressure limits, (3) an 
increase in main steam line (MSL) isolation valve and tunnel high 
differential temperature to reflect the increased operating pressure, 
(4) revised temperature/pressure limit curves to reflect the higher 
neutron flux over vessel life, (5) an increase in calculated peak 
containment pressure, (6) an increase in the pressure at which reactor 
core isolation cooling (RCIC) testing occurs, and (7) an increase in 
the safety relief valve (SRV) setpoints.

    The licensee, in the above reference letters, also proposed changes 
to the TS that are not associated with the power uprate. These proposed 
changes are (1) a change of the reactor protection system and End-of-
Cycle/Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a 
variable setpoint based on power level, (2) an increase in the SRV 
setpoint tolerance, and (3) an increase in the number of automatic 
depressurization system (ADS) valves that are allowed out of service.

    Although not part of the change to the license or associated TS, 
the licensee is also revising the bases to reflect the TS changes and 
power uprate.

    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 10 CFR 50.92 (c). The NRC staff's review is 
presented below:

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

    * [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]Plant 
operation at a higher power level is accomplished by increasing reactor 
core flow along flow control lines to achieve the desired increase in 
steam flow to the turbine/generator. The maximum allowable reactor 
recirculation flow rate remains unchanged from the original plant 
design analysis. The increased reactor core flow requires a 
corresponding increase in feedwater flow that remains well within the 
design of the feedwater system components. The increased power also 
requires a small increase in reactor pressure. Safety relief valve and 
power, pressure, and flow-related instrumentation trip setpoints are 
increased slightly to accommodate the power uprate, to maintain 
approximately the same level of trip avoidance and safety system 
challenges as before the uprated power condition.
    The plant is operated in the same manner at uprated power as it is 
at the currently licensed power level, including the methods and 
sequences of system and component operation. Since the level of trip 
avoidance and safety system challenges remains approximately the same, 
the frequency of operational responses to these events is not 
increased. Reactor fuel operating limits, designed to protect the fuel 
cladding, are maintained and thus provide the same level of protection 
as before the uprated power condition. The original design and 
regulatory criteria established for plant equipment, including ASME 
code, IEEE standards, NEMA standards, and Regulatory Guide criteria, 
are still imposed and met for operation at the uprated power level. In 
addition, the reactor vessel and internals, reactor connecting piping, 
balance of plant piping, primary containment, and related systems and 
components still meet the pre-uprate design and licensing criteria. The 
power uprate does not change the likelihood of failure of these systems 
or components. Thus, the probability of an accident previously 
evaluated is not significantly affected by the proposed power uprate.
    The consequences of postulated, power-dependent accidents are 
proportional to the power level assumed in the safety analysis. This is 
because potential offsite doses increase proportionately to reactor 
power since the radiological source term is directly proportional to 
reactor power. The meteorological factors are unaffected by the 
proposed power uprate. The current accident analyses are based on 
104.4% of original rated power. The accident analyses for power uprate 
are based on 1.02 x 104.9% (or 107%) of original rated power. Thus, 
power uprate increases postulated consequences of an accident by less 
than 3% over previous postulated consequences, while still remaining 
well within the 10 CFR Part 100 limits. In addition, a spectrum of 
hypothetical accidents and transients were investigated for power 
uprate, and the bounding events met the same regulatory criteria to 
which WNP-2 is currently licensed, including Maximum Average Planar 
Linear Heat Generation Rate (MAPLHGR), Operating Limit Minimum Critical 
Power Ratio (OLMCPR), 10 CFR 50.46 and 10 CFR Part 50-Appendix K, for 
fuel cladding integrity, and containment criteria in 10 CFR Part 50-
Appendix A, Criterion 38 and Criterion 50.
    The results of these analyses demonstrate that operation at the 
proposed power uprate level does not significantly increase the 
probability or consequences of any accident previously evaluated.
    * [Increase in the maximum allowable reactor steam dome operating 
pressure limit]
    The operating pressure limit is increased by the same amount as the 
nominal operating pressure increase for power uprate. This change to 
the dome operating pressure limit is consistent with and meets the 
current design criteria used for evaluation of steady state operating 
conditions and for the most limiting event for vessel overpressure 
protection. Operation at a higher pressure results in the plant 
operating closer to the criteria used in the design analysis. However, 
since operation of the plant within design limits is considered to 
result in a very low probability of failure of systems or components, 
this small increase in maximum operating pressure is considered to have 
a negligible increase in failure probability. Thus the proposed change 
does not significantly increase the likelihood of failure of existing 
systems or components, and does not significantly affect the 
probability of an accident previously evaluated.
    The power uprate overpressure protection analysis results show the 
peak reactor pressure vessel (RPV) pressure will remain below the ASME 
code limit, keeping any postulated radiological consequences within the 
bounds of existing analyses. Thus the consequences of an accident 
previously evaluated are not significantly affected.
    * [Increase in the average power range monitor (APRM) and flow 
biased scram and rod block monitor (RBM) setpoints]
    These scrams and rod blocks are designed to prevent fuel damage due 
to power during postulated events or anticipated operational 
occurrences. The setpoints for these scrams and rod blocks are 
increased by the same amount of the proposed increase in licensed 
power. The change in these setpoints does not affect the operation of 
any system or component, nor does it affect the circuitry that provides 
the protective function. Thus the increase in setpoints does not affect 
the probability of any accident previously evaluated.
    The increased setpoints maintain the same difference between 
licensed power level and scram setpoints, and between the scram line 
and rod block lines on the extended load line limit curve. The 
increased setpoints result in higher postulated source terms for 
accidents previously analyzed. The proposed change does not affect the 
response or operation of mitigative equipment. The licensee's analyses 
supporting power uprate demonstrate that the increased values do not 
significantly affect the consequences of an accident. The proposed 
change also maintains the current level of scram avoidance with 
associated avoidance of unnecessary challenges to plant equipment, 
while providing protection for the fuel, reactor systems, and 
containment to meet current design requirements. Thus the proposed 
change does not significantly increase the consequences of accident 
previously analyzed.
    * [Increase in main steam line (MSL) high flow differential 
pressure setpoint to reflect the increased operating pressure]
    The MSL high flow differential pressure setpoint is increased to 
reflect the higher steam flows necessary to operate the plant at the 
higher power. The proposed change does not affect the design, 
construction, or operation of existing plant equipment, nor does it add 
new equipment. The proposed change does not, therefore, affect the 
probability of accidents previously analyzed. The increased setpoint 
does not affect the maximum closure time for the main steam isolation 
valves, thus the release of fission products during postulated accident 
is not changed from current design analyses. The proposed change does 
not, therefore, affect the consequences of accidents previously 
analyzed.
    * [Revised temperature/pressure limit curves to reflect the higher 
neutron flux over vessel life]
    The temperature/pressure limit curves are being modified to reflect 
the increased exposure to neutron flux over the life of the vessel, to 
retain the existing margin to brittle fracture over vessel life. The 
plant will continue to be operated in conformance to the pressure/
temperature limits, retaining the existing probability of overcooling 
events and associated postulated brittle fracture of the reactor 
vessel. If a failure were to occur, the mode of failure is unaffected 
by the proposed change, thus the consequences of a postulated failure 
of the reactor vessel is unchanged by the proposed amendment.
    * [Increase in the pressure at which reactor core isolation cooling 
(RCIC) testing occurs]
    The pressure at which the RCIC system is tested is being increased 
to ensure adequate system operation with the increased reactor system 
pressure required for power uprate. No credit for RCIC system operation 
is taken in any accident analysis nor is RCIC included as an accident 
initiator, thus this change does not affect the probability or 
consequences of an accident previously evaluated.
    * [Change of the reactor protection system and End-of-Cycle/ 
Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a 
variable setpoint based on power level]
    This setpoint is being changed from a specific value of main 
turbine first stage pressure to a value to be calculated based on 
variable plant conditions. The basis for the setpoint will remain 
unchanged at a value equivalent to thermal power less than 30% of rated 
thermal power. This change is proposed because turbine first stage 
pressure can vary for a given thermal power level depending on the 
amount of subcooling in the core, which itself is highly dependent on 
feedwater temperatures. This proposed change does not affect the 
reliability of operation of the trip circuitry, nor does it affect the 
design or operation of plant equipment or safety systems. Thus the 
change does not affect the probability of accidents previously 
evaluated. In addition, the proposed change retains the same basis for 
establishing the setpoint as currently stated in the TS, thus the 
setpoint will be established at the level (30% rated thermal power) 
assumed in the accident analysis. Thus the proposed change does not 
affect the consequences of any accidents previously evaluated.
    * [Increase in the safety relief valve (SRV) setpoints]
    The licensee proposes to increase the setpoints of the two lowest-
set SRVs to accommodate the change in maximum operating pressure after 
power uprate. This increase in SRV setpoints maintains approximately 
the same difference between maximum operating RPV pressure and the 
lowest SRV setpoint as currently exists. As discussed in the above 
section regarding increased operating pressure, increasing the pressure 
at which the lowest SRV actuates would allow operation at a slightly 
higher pressure than currently allowed. This results in the plant 
operating closer to the criteria used in the design analysis. However, 
since operation of the plant within design limits is considered to 
result in a very low probability of failure of systems or components, 
this small increase in maximum operating pressure is considered to have 
a negligible increase in failure probability. Thus the proposed change 
does significantly increase the likelihood of failure of existing 
systems or components, and does not affect the probability of an 
accident previously evaluated.
    With the increased SRV setpoint, the analysis results show that the 
peak RPV pressure will remain below the ASME code limit, keeping any 
postulated radiological consequences within the bounds of existing 
analyses. Thus the consequences of an accident previously evaluated are 
not affected.*
    [Increase the value of Pa (the pressure at which primary 
containment is tested). Add a new definition for Pa in the 
``Definition'' section of the TS that gives a specific value of 
Pa, and simplify the TS by deleting the specific value of Pa 
and 1.10Pa from other locations in the TS]
    The analysis to support the power uprate resulted in a higher 
calculated peak containment pressure in response to postulated 
accidents. To maintain the validity of the radiological analysis, the 
containment leak rate testing must be based on a pressure greater than 
or equal to the peak calculated containment pressure. The proposed 
change does not affect the design, construction, or operation of 
existing plant systems or components, and therefore does not affect the 
probability of accidents previously evaluated.
    The increase in test pressure does not affect the acceptance 
criteria for the test, which is based on acceptable leakage. Offsite 
dose projections are based in part on the leakage criteria, and since 
the proposed change does not affect the leakage limits, the change does 
not affect the consequences of accidents previously evaluated.
    The addition of a definition of Pa, the delineation of the 
specific value in the definition and associated deletion of the 
specific value in individual TS, is a purely administrative change that 
does not affect plant design, construction, or operation. This part of 
the proposed change does not, therefore, affect the probability or 
consequences of an accident previously evaluated.
    * [Increase in the SRV setpoint tolerance]
    The licensee is proposing to increase the setpoint tolerance for 
the SRVs from +1/-3% to plus or minus 3% of the setpoint. This proposed 
change does not affect how the SRVs operate in response to accidents or 
abnormal operating occurrences, and thus does not affect the 
probability of an accident previously evaluated.
    The proposed increase in setpoint tolerance results in a 
potentially higher pressure at which an SRV would lift. This increase 
in pressure is approximately 23 psig, based on the proposed maximum 
system pressure and proposed setpoint tolerance. This 2% increase in 
maximum lift pressure would result in a proportional increase in 
possible offsite dose due to a postulated event. The dose increase 
would be something less than 2%, since the increased pressure would 
result in a corresponding increase in differential pressure, with 
attendant flow losses and subsequent filtration through accident 
mitigation systems accounting for the lower dose. This small increase 
in postulated offsite dose is not considered a significant increase in 
the consequences of accidents previously evaluated.
    * [Increase in the number of automatic depressurization system 
(ADS) valves that are allowed out of service]
    The proposed change would allow one ADS valve to be out of service 
without time limit, and changing the minimum number of ADS valves 
required to be in service from seven to six. The change would allow two 
ADS valves (compared to the current one) to be out of service for up to 
14 days, and would require plant shutdown within 12 hours for three or 
more ADS valves (compared to the current two or more) out of service. 
The proposed change would reduce the likelihood of an inadvertent 
opening ADS/SRV as an initiating event if one SRV were out of service. 
Having more than one ADS/SRV out of service would also reduce this 
likelihood, although the reduction would have minimal effect since the 
probability of multiple ADS/SRVs being out of service with a concurrent 
lift and failure to close of an SRV is small. Thus the proposed change 
would have little effect on the probability of accidents previously 
evaluated.
    The ADS/SRVs serve to limit overpressurization of the RPV and 
connected primary piping. Having ADS valves out of service would 
increase the consequences compared to the current TS which limit any 
ADS valves out of service for more than 14 days. Reanalysis of the 
effect of ADS valves on protection of the RPV demonstrated that five 
ADS valves would prevent RPV overpressurization as the previous 
analysis using six ADS valves. Thus the proposed change would not 
affect the consequences of accidents previously evaluated.
    2. Does the amendment create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    * [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]
    Equipment that could be impacted by power uprate has been evaluated 
by the licensee. The proposed change has not introduced any new 
operating modes, equipment lineups, accident scenarios, or equipment 
failure modes. The full spectrum of accident considerations defined in 
Regulatory Guide 1.70 has been reviewed, and no new or different kind 
of accident has been identified. Power uprate uses existing technology 
and applies it within the capabilities of existing plant equipment in 
accordance with existing regulatory criteria including NRC-approved 
codes, standards, and methods. General Electric has designed to higher 
power levels than the uprated power of WNP-2, and no new power-
dependent accidents have been identified. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    * [Increase in the maximum allowable reactor steam dome operating 
pressure limit]
    The proposed change does not involve the addition of new equipment, 
nor has it introduced any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Increase in the average power range monitor (APRM) and flow 
biased scram and rod block monitor (RBM) setpoints]
    The proposed setpoint changes do not involve the addition of new 
equipment, nor do they introduce any new operating modes, equipment 
lineups, accident scenarios, or equipment failure modes. The proposed 
changes do not, therefore, create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    * [Increase in main steam line (MSL) high flow differential 
pressure setpoint to reflect the increased operating pressure]
    The proposed setpoint change does not involve the addition of new 
equipment, nor does it introduce any new operating modes, equipment 
lineups, accident scenarios, or equipment failure modes. The proposed 
change does not, therefore, create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    * [Revised temperature/pressure limit curves to reflect the higher 
neutron flux over vessel life]
    The proposed change does not involve the addition of new equipment, 
nor does it introduce any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Increase in the pressure at which reactor core isolation cooling 
(RCIC) testing occurs]
    The proposed change does not involve the addition of new equipment, 
nor does it introduce any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Change of the reactor protection system and End-of-Cycle/ 
Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a 
variable setpoint based on power level]
    The proposed change does not involve the addition of new equipment, 
nor does it introduce any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Increase in the safety relief valve (SRV) setpoints]
    The proposed change does not involve the addition of new equipment, 
nor does it introduce any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Increase the value of Pa (the pressure at which primary 
containment is tested). Add a new definition for Pa in the 
``Definition'' section of the TS that gives a specific value of 
Pa, and simplify the TS by deleting the specific value of Pa 
and 1.10 Pa from other locations in the TS]
    The proposed change does not involve the addition of new equipment, 
nor do it introduce any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Increase in the SRV setpoint tolerance]
    The proposed change does not involve the addition of new equipment, 
nor does it introduce any new operating modes, equipment lineups, 
accident scenarios, or equipment failure modes. The proposed change 
does not, therefore, create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    * [Increase in the number of automatic depressurization system 
(ADS) valves that are allowed out of service]
    The proposed change does not involve the addition of new equipment, 
nor does it introduce any new operating modes, accident sequences, or 
equipment failure modes. The proposed change does allow a new equipment 
lineup by allowing operation with one ADS valve out of service with no 
limitation, which is not allowed by the current TS. In addition, two 
ADS valves can be out of service for an extended period of time (14 
days), with shutdown within 12 hours required only for three or more 
valves inoperable compared to the current two valves. This allowed 
equipment lineup does not alter the operation of the ADS valves or 
their impact as potential event initiators as discussed in the FSAR. 
The proposed change does not, therefore, create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.

    3. Does the amendment involve a significant reduction in a margin 
of safety?*
    [Increase in Rated Thermal Power from 3323 MWt to 3486 MWt]
    The plant was originally designed for operation at 105% rated steam 
flow. The proposed change therefore does not affect the fuel design or 
safety limits. The Maximum Axial Power Linear Heat Generation Rate 
(MAPLHGR) limits remain the same. The Operating Limit Minimum Critical 
Power Ratio (OLMCPR) is expected to increase by approximately 2%, which 
will ensure that the margin to the Safety Limit MCPR is not affected.
    The entire plant design has been reviewed to ensure that plant 
equipment will perform properly and will still meet original design and 
licensing criteria. The safety margins prescribed by the Code of 
Federal Regulations have been maintained by meeting the appropriate 
regulatory criteria. The margins provided by the application of the 
ASME design acceptance criteria have been maintained, as well as other 
margin-assuring acceptance criteria.
    The emergency core cooling system-loss of coolant accident analysis 
was conservatively performed based on two ADS valves out of service and 
power level corresponding to 110% of original rated steam flow (plus 2% 
power uncertainty factor). The analyzed results remain well below the 
safety margin established at the 2200 deg.F peak centerline temperature 
regulatory limit.
    The overpressurization and containment analyses were repeated based 
on 110% of original rated steam flow, resulting in a slightly higher 
peak reactor vessel pressure. The increased pressure remains below the 
acceptance criteria for the design basis, which the licensee identified 
as below the ASME code limit and below the applicable TS limit.
    From the containment analysis, the peak containment pressure 
increases from 34.7 psig to 35.1 psig. This remains below the 
containment design pressure of 45 psig.
    The postulated radiological doses of design basis events, including 
the bounding analysis involving loss of coolant accident, were 
calculated based on the uprate power level. The results remain within 
the design basis established by 10 CFR Part 100.
    Based on these considerations, the proposed increase in licensed 
power level does not significantly reduce any margins of safety.
    * [Increase in the maximum allowable reactor steam dome operating 
pressure limit]
    The maximum pressure is increased by the same amount as the nominal 
operating pressure increase for power uprate. The increased pressure 
remains below the acceptance criteria for the design basis, which the 
licensee identified as below the ASME code limit and below the 
applicable TS limit. Thus the proposed TS change does not reduce the 
margin to safety.
    * [Increase in the average power range monitor (APRM) flow biased 
scram and rod block monitor (RBM) setpoints]
    The APRM flow biased scram setpoints were increase by the same 
percentage of power as the uprated power. This maintains the same 
margin to the trip setpoint while maintaining the same scram avoidance 
as originally designed. The current margin (9%) between the scram line 
and rod block line is maintained for power uprate. Thus the proposed TS 
changes do not reduce any margin to safety.
    * [Increase in main steam line (MSL) high flow isolation 
differential pressure setpoint to reflect the increased operating 
pressure]
    The revised safety analysis retains the current analytic basis of 
140% of rated steam flow to ensure the same level of scram avoidance is 
maintained. This results in approximately a 10% increase in the 
differential pressure required to trip the plant. This change does not, 
however, affect the assumed closure times for the main steam isolation 
valves in design analyses. The closure times determine the analyzed 
release of fission products during postulated accidents. Since the 
closure time is unaffected by the proposed change, the change does not 
affect the margin of safety.
    * [Revised temperature/pressure limit curves to reflect the higher 
neutron flux over vessel life]
    The proposed increase in licensed power also increases the neutron 
fluence on the reactor pressure vessel over the license period (40 
years). The analysis for the pressure/temperature limit curves was 
updated to incorporate the increased fluence, and the revised curves 
maintain the same margin to postulated brittle fracture of the reactor 
vessel as the current TS curves. The proposed change does not, 
therefore, change a margin of safety.
    * [Increase in the pressure at which reactor core isolation cooling 
(RCIC) testing occurs]
    The proposed change would increase the test pressure for RCIC to 
conform to the increased system pressure resulting from the power 
uprate. Increasing the test pressure periodically verifies that RCIC 
will operate at the increased pressure resulting from power uprate. In 
addition, RCIC is not credited in safety analyses for assuring that 
margins of safety are maintained. The proposed change, therefore, does 
not affect any margin of safety.
    * [Change of the reactor protection system and End-of-Cycle/ 
Recirculation Pump Trip (EOC/RPT) trip setpoint from a fixed to a 
variable setpoint based on power level]
    The safety analyses assume an EOC/RPT at 30% power. This proposed 
change will allow the setpoint to vary relative to turbine first stage 
pressure (the current TS setpoint), but will assure that the setpoint 
is set based on a fixed 30% power. This assures that the EOC/RPT 
maintains the existing margin of safety.
    * [Increase in the safety relief valve (SRV) setpoints]
    The two low set SRV setpoints are being increased by the same 
amount of operating system pressure increase resulting from the power 
uprate. The SRVs at the new setpoints will relieve pressure to ensure 
the reactor coolant system remains below the acceptance criteria for 
the design basis, which the licensee identified as below the ASME code 
limit and below the applicable TS limit. Thus the proposed TS change 
does not reduce the margin to safety.
    * [Increase the value of Pa (the pressure at which primary 
containment is tested). Add a new definition for Pa in the 
``Definition'' section of the TS that gives a specific value of 
Pa, and simplify the TS by deleting the specific value of Pa 
and 1.10Pa from other locations in the TS]
    The pressure at which primary containment is to be tested is being 
increased to reflect the analyzed results of the power uprate. The 
increase will assure that the periodic verification of containment 
integrity is performed at a pressure that assures that postulated 
radioactive release rates remain within analyzed assumptions. This 
assures that existing margins of safety are maintained.
    * [Increase in the SRV setpoint tolerance]
    The proposed increase in the SRV setpoint tolerance from +1/-3% to 
plus or minus 3% of the setpoint has the potential to increase the 
actual pressure at which the SRV would lift. This increase in pressure 
is approximately 23 psig, or 2% of the design lift pressure. This 
pressure is still well within the RPV design pressure that establishes 
the safety margin. The proposed change would not, therefore, affect a 
margin of safety.
    * [Increase in the number of automatic depressurization system 
(ADS) valves that are allowed out of service]
    The proposed change results in an increase in the calculated peak 
centerline temperature (PCT) for postulated accidents, specifically, 
the loss of coolant accident (LOCA). The margin of safety for this 
parameter at WNP-2 is 2200 deg.F, which is also the regulatory limit. 
Maintaining PCT less than 2200 deg.F ensures cladding integrity, 
including the safety margin established by regulation. The proposed 
change does not, therefore, affect a margin of safety.
    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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Theodore R. Quay

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


    Date of amendment request: October 31, 1994
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications (TS) to: (1) add two action 
statements that would provide allowed outage times for either one or 
both of the scram discharge volume (SDV) vent or drain valves less 
stringent that the current requirements of TS 3.0.3.; and (2) change 
the surveillance requirements for the SDV vent and drain valves to 
conduct the testing during shutdown conditions rather than at power as 
currently required.
    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?
     [Adding action statements for allowed outage times]
    The primary functions of the SDV vent and drain valves are to 
isolate the SDV following a scram to stop leakage of reactor coolant 
past the CRD seals and to reopen following a scram to reset to drain 
the reactor coolant from the SDV to the reactor building equipment 
drain sump. The SDV is sized to accept CRD over piston discharge 
water from all 185 CRDs. The SDV vent and drain valves reopen when 
the scram signal is reset to provide assurance that there is 
sufficient SDV volume available to accept the CRD discharge in the 
event of another scram. Each vent and drain line contains two 
redundant valves in series, which close to isolate the SDV on a 
scram signal.
    Proposed Action Statement 3.1.3.1.d allows 7 days to repair an 
inoperable valve when the redundant valve is still operable, or 
isolate the affected line. The reliability of the isolation function 
is reduced during the period of the 7 day Allowed Outage Time (AOT) 
since a single failure could prevent a redundant valve from 
isolating a SDV vent or drain line. As a result, the proposed 7 day 
AOT introduces an increase in the risk of an unisolated path for 
reactor coolant release to the reactor building floor drain or 
equipment drain sump. However, the reduction in reliability can only 
affect plant safety if the redundant valve should fail to close 
during an accident involving core damage. The probability of two 
valves in series failing to isolate the SDV upon demand is about 4.9 
x 10-5. The redundant valve is designed to automatically close 
in response to a scram signal or upon loss of air or electrical 
power. The probability of having one valve fail a surveillance test, 
combined with a subsequent scram and redundant valve failure during 
the 7 day AOT following the surveillance is about 8.2 x 10-7 
per year. The risk of failure to isolate the SDV is increased by 
about 2% (as determined by the WNP-2 Individual Plant Evaluation 
(IPE)) by proposed Action Statement 3.1.3.1.d, which is not 
considered significant.
    A failure of the SDV to isolate during a scram generally does 
not pose a hazard because the reactor coolant from CRD discharge and 
seal leakage is routed to the reactor building drain sumps. The 
release of reactor coolant can be terminated by resetting the scram 
from the control room, which would close the scram outlet valves, or 
by manually closing the isolation valves located in the reactor 
building. Failure of the SDV to isolate can pose a risk of higher 
consequences if a core damage accident occurs simultaneously. The 
increased risk of activity release during a core damage accident due 
to addition of the 7 day AOT is less than 1 x 10-9 per year.
    Proposed Action Statement 3.1.3.1.e allows 8 hours for repair 
when both valves in a line are inoperable. If a scram should occur 
during this 8 hour period, both valves in a SDV vent or drain line 
could fail to isolate the line, creating a path for reactor coolant 
release to the reactor building floor drain or equipment drain sump. 
The probability of having to enter the proposed action statement is 
low, at about 4.9 x 10-5 per year. The probability of entering 
the action statement, combined with a scram occurring in the 
following 8 hours is about 2.6 x 10-7 per year. This represents 
less than a 1% increase in the probability of SDV isolation failure 
upon demand during normal operation, which is not considered 
significant.
    As stated above, a failure of the SDV to isolate during a scram 
generally does not pose a hazard. The reactor coolant from CRD 
discharge and seal leakage is routed to the reactor building 
equipment drain sump. The release of reactor coolant can be 
terminated by resetting the scram from the control room or by 
manually closing the isolation valves located in the reactor 
building. Failure of the SDV to isolate can pose a risk of higher 
consequences if a core damage accident occurs simultaneously. The 
increased risk of radioactivity release during a core damage 
accident due to addition of the 8 hour AOT is less than 1 x 10-
9 per year. To date, there have not been any instances at WNP-2 
where both valves in a SDV vent or drain line were inoperable at the 
same time during plant operation. If this unlikely event were to 
occur, the proposed action statement would require the affected line 
to be isolated within 8 hours. With a SDV vent or drain line 
isolated, normal operational leakage from the scram outlet valves 
would cause the SDV instrument volume level to increase. However, 
ample time would exist after receipt of a SDV high level alarm in 
the control room to drain the SDV instrument volume to prevent 
actuation of an automatic scram on high SDV level.
    Since it is unlikely that both valves in an SDV vent or drain 
line would be inoperable and isolated, the periodic opening of an 
isolated line under administrative control for SDV instrument volume 
venting and draining in accordance with proposed Note 
 is expected to be very infrequent. In addition, 
Proposed Note  does not adversely affect plant safety by 
permitting separate action statement entry for each vent and drain 
line since the probability of entering the proposed action 
statements is low.
    Based on the information presented above, it is concluded that 
the AOTs proposed in Action Statements 3.1.3.1.d and 3.1.3.1.e and 
the associated Notes do not represent changes that involve a 
significant increase in the probability of an accident previously 
evaluated.
    As discussed above, the proposed AOTs and the associated Notes 
introduce a small increase in the risk of an unisolated path for 
reactor coolant release to the reactor building floor drain or 
equipment drain sump through an unisolated SDV vent or drain line. 
However, this event is bounded by the *NUREG-0803 evaluation of the 
consequences of a postulated reactor scram and SDV rupture. Based on 
the NUREG safety evaluation, the volume of coolant lost via the 
bounding leakage pathway is relatively small (approximately 550 gpm 
or the equivalent of a 1.008 inch break), and adequate core cooling 
would be maintained such that no fuel failures are predicted for the 
event. The NRC staff concluded in the NUREG safety evaluation that 
resulting reactor building flooding for this event did not adversely 
impact safety-related equipment. The NRC staff also concluded that 
the area of the reactor building where the leak occurs will become 
contaminated only to the activity level normally present in the 
reactor coolant, and offsite doses would be well within the 10 CFR 
Part 100 reference values for plants operating with Standard 
Technical Specification (STS) coolant activity limits (0.2 
microCuries/gm). The release of reactor coolant through an 
unisolated SDV vent or drain line can be terminated by resetting the 
scram from the control room, which would close the scram outlet 
valves, or by manually closing the isolation valves located in the 
reactor building. The NUREG-0803 evaluation determined that the 
reactor building would be accessible to terminate leakage during a 
postulated reactor scram and SDV rupture with appropriate 
radiological precautions. In addition, the SDV vent and drain lines 
route the release in a controlled manner to the reactor building 
floor drain and equipment drain sumps. Thus, the consequences are 
significantly less than those of the SDV rupture analysis.
    The WNP-2 FSAR Accident Analyses, Section 15.6, ``Decrease in 
Reactor Coolant Inventory,'' acceptance limits for radiological 
consequences are based on the guidance set forth in 10 CFR Part 100. 
Since WNP-2 operates in accordance with the STS coolant activity 
limits and the offsite dose reference values of 10 CFR Part 100, the 
consequences established in the NUREG-0803 safety evaluation bound 
the consequences of an unisolated SDV vent or drain line event. 
Therefore, the AOTs proposed in Action Statements 3.1.3.1.d and 
3.1.3.1.e and the associated Notes do not represent changes that 
involve a significant increase in the consequences of an accident 
previously evaluated.
     [Changing surveillance requirements]
    The primary functions of the SDV vent and drain valves are to 
isolate the SDV following a scram to stop leakage of reactor coolant 
past the CRD seals and to reopen following a scram reset to drain 
the reactor coolant from the SDV to the reactor building equipment 
drain sump. The basis for Surveillance Requirement 4.1.3.1.4.a is to 
verify SDV vent and drain valve operability so that the SDV will be 
available when needed to accept CRD over piston discharge water and 
so that the reactor coolant collected in the SDV will be isolated 
from the secondary containment (reactor building). Performance of 
Surveillance Requirement 4.1.3.1.4.a, with the proposed deletion of 
the control rod configuration and density requirements and 
associated Note *, will still ensure that the safety functions and 
operability requirements are met.
    Valve operability can be demonstrated from shutdown conditions 
even though the surveillance test conditions of nearly ambient 
temperature and pressure and reduced CRD discharge flow do not match 
power conditions. Maximum SDV back pressure and CRD discharge flow 
will not significantly affect the SDV vent and drain valves closure 
rates. As verified by testing at LaSalle County Station and WNP-2, 
there is only approximately a 1 second difference in SDV vent and 
drain valve closing time from a scram at less than or equal to 50% 
rod density versus the closure time from either the test pushbuttons 
or a cold shutdown scram with all rods full in. These test results 
show that the differences in temperatures, pressures, and CRD 
discharge flows between power and cold shutdown conditions have a 
negligible effect on SDV vent and drain valves closing times. [In 
addition, the valves will be tested in the open direction at cold 
shutdown conditions during the same test conducted for valve 
closure.] Although the ability of the valves to open against full 
reactor pressure cannot be demonstrated during shutdown conditions, 
[additional verification of the valves' ability to open will be] 
verified [sic] as part of the scram recovery procedure. Thus, the 
ability of the valves to open against full reactor pressure will 
still be demonstrated after each reactor scram during operation.
    Since the operability of the SDV vent and drain valves can be 
demonstrated by performing Surveillance Requirement 4.1.3.1.4.a 
during shutdown conditions, the change does not represent a 
significant 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?
     [Adding action statements for allowed outage times]
    The AOTs proposed in Action Statements 3.1.3.1.d and 3.1.3.1.e 
and the associated Notes do not involve any changes to the facility 
or operation of the facility as described in the WNP-2 FSAR. The 
isolation function of the SDV vent and drain valves to prevent the 
discharge of reactor coolant into the reactor building floor drain 
and equipment drain sumps following a scram is maintained. The 
release of reactor coolant through a SDV vent or drain line during a 
scram can be isolated either by automatic closure of the redundant 
valve in response to the scram signal or by manual isolation of the 
affected line. The valves will also close automatically upon loss of 
air to the valves or electrical power to the associated solenoid 
pilot valves. The alarm, control rod withdrawal block, and reactor 
scram functions on increasing water level in the SDV instrument 
volume are unaffected by the proposed amendments. Although the 
proposed AOTs will change the method of plant operation, potentially 
resulting in a reduction in SDV vent or drain line isolation 
reliability, the potential release to the reactor building drain 
sumps has been previously evaluated in NUREG-0803, and as shown in 
(1) above, the associated probabilities of the event are acceptably 
low.
    Since the AOTs proposed in Action Statements 3.1.3.1.d and 
3.1.3.1.e and the associated Notes do not involve a change to the 
facility or the method of operation that has not been previously 
evaluated, the change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
     [Changing surveillance requirements]
    The deletion of the control rod configuration and density 
requirements and associated Note * proposed for Surveillance 
Requirement 4.1.3.1.4.a only change the conditions under which the 
surveillance is performed. As such, the change does not involve a 
change to the facility or method of operation as describe in the 
WNP-2 FSAR. As discussed in (1) above, performance of the 
surveillance with the proposed changes will still demonstrate SDV 
vent and drain valve operability to ensure that the SDV will perform 
as evaluated in the FSAR accident analysis.
    Since the performance of Surveillance Requirement 4.1.3.1.4.a 
during shutdown conditions does not involve a change to the facility 
or the method of operation, the change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Does the change involve a significant reduction in a margin of 
safety?
     [Adding action statements for allowed outage times]
    As discussed in (1) above, the AOTs proposed in Action 
Statements 3.1.3.1.d and 3.1.3.1.e and the associated Notes 
introduce a small increase in the risk of an unisolated path for 
reactor coolant release to the reactor building floor drain and 
equipment drain sumps through an unisolated SDV vent or drain line. 
The increased risk is due to the reduction in isolation function 
reliability when SDV vent and drain valves are inoperable. Since 
this reduction in reliability can only affect plant safety during a 
scram, it is expected that the increased risk of release due to a 
scram attributed to the AOTs will be offset by the reduced risk of a 
scram that will result from a reduction in the number of manual 
plant shutdowns. Currently, if one or more of the SDV vent or drain 
valves is discovered to be inoperable, an immediate plant shutdown 
is required. Establishment of the AOTs will eliminate these 
unnecessary plant shutdowns that limit plant operational flexibility 
and increase the risk of a plant scram and challenges to safety 
systems. Moreover, there is only a small probability of a scram 
occurring during the AOTs coincident with the failure of a SDV vent 
or drain line to isolate. The release of reactor coolant to the 
reactor building floor drain and equipment drain sumps through an 
unisolated line can be terminated by resetting the scram from the 
control room or by manually closing the isolation valves. 
Furthermore, the consequences of such an event are bounded by the 
consequences of the postulated reactor scram and SDV rupture event 
evaluated in NUREG-0803 and would be well within the 10 CFR Part 100 
reference values.
    Since the AOTs proposed in Action Statement 3.1.3.1.d and 
3.1.3.1.e and the associated Notes do not change the assumptions or 
increase the consequences of the bounding NUREG-0803 accident 
analysis, the margin to the 10 CFR Part 100 reference values is not 
changed. Therefore, the change does not involve a significant 
reduction in the margin of safety.
     [Changing surveillance requirements]
    Performance of Surveillance Requirement 4.1.3.1.4.a with the 
proposed deletion of the control rod configuration and density 
requirements and associated Note * will still ensure that the SDV 
vent and drain valve safety functions and operability requirements 
are met. Valve operability can be demonstrated from shutdown 
conditions even though the surveillance test conditions of nearly 
ambient temperature and pressure at shutdown and reduced CRD 
discharge flow do not match power conditions. As discussed in (1) 
above, the difference in test conditions represents only 
approximately a 1 second difference in the 30 second (as specified 
in Surveillance Requirement 4.1.3.1.4.a.1) SDV vent and drain valve 
closing times.
    The potential reduction in safety margin is related to the 
reliability of the SDV vent and drain valves to close within the 
required time to contain the reactor coolant leakage past the CRD 
seals following a scram. The consequences of the valves failing to 
close to isolate a line are bounded by the postulated reactor scram 
and SDV rupture event evaluated in NUREG-0803, which assumes a 
constant leakage rate for 4 hours. Since the NUREG evaluation 
concluded that the consequences of such an event would be well 
within the 10 CFR Part 100 reference values, the potential one 
second difference in valve closing time is relatively insignificant. 
In addition, the proposed surveillance requirement would eliminate 
approximately 20 scrams at power over the remaining life of the 
plant and prevent the concomitant transients and challenges to 
safety systems. This would be expected to increase the reliability 
of the SDV vent and drain valves and mitigate any reduction in 
safety margin. Although performance of the proposed surveillance 
requirement during shutdown conditions will not demonstrate the 
ability of the valves to open against a back pressure equal to full 
reactor pressure, the valves are verified to be open as part of [the 
test conducted at shutdown conditions as well] the scram recovery 
procedure. Thus, the ability of the valves to open against full 
reactor pressure will still be demonstrated after each reactor scram 
during operation.
    Since the performance of Surveillance Requirement 4.1.3.1.4.a 
during shutdown conditions does not change the assumptions or 
increase the consequences of the bounding NUREG-0803 accident 
analysis, the margin to the 10 CFR Part 100 reference values is not 
changed. Therefore, the change does not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, N.W., Washington, D.C. 20005-3502
    NRC Project Director: Theodore R. Quay

Washington Public Power Supply System, Docket No. 50-397, Nuclear 
Project No. 2, Benton County, Washington Date of application for 
amendment: October 31, 1994

    Brief description of amendment: The proposed amendment would revise 
the technical specifications to remove the requirements related to 
operability and surveillance testing of the safety/relief valve (SRV) 
position indication instrumentation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the standards of 50.92(c). The NRC staff's review is presented 
below.
    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change would relocate the SRV position indication 
instrumentation TS requirements to other licensee-controlled documents. 
This is an administrative change which does not involve any 
modification to plant equipment or plant operation as described in the 
WNP-2 Final Safety Analysis Report (FSAR). The instrumentation would 
continue to be available to provide SRV position indication to the 
operators. Therefore, the proposed amendment does not involve an 
increase in the probability or consequences of a previously evaluated 
accident.
    2. Does the proposed amendment create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change does not involve any physical alteration to 
plant equipment and results in no changes in the performance of any 
safety-related system. Therefore, the proposed amendment does not 
create the possibility of a new or different kind of accident from any 
previously evaluated accident.
    3. Does the proposed amendment involve a significant reduction in 
the margin of safety?
    As noted above, the proposed change does not involve any 
modification to plant equipment or plant operation. The proposed change 
does not affect any accident analyses contained in the WNP-2 FSAR. 
Therefore, the proposed amendment does not involve a significant 
reduction in 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: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Attorney for licensee: M. H. Philips, Jr., Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Theodore R. Quay

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 Nos. 1 and 2, Calvert 
County, Maryland
    Date of application for amendments:  August 2, 1994
    Brief description of amendments: The amendments revise Technical 
Specifications 3.9.1 and 3.1.2.7 to clarify the requirements when the 
required boron concentration is greater than 2300 ppm.
    Date of issuance: November 29, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment Nos.:  Unit 1 - 201 - Unit 2 - 179
    Facility Operating License Nos. DPR-53 and DPR-69: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: September 14, 1994 (59 
FR 47164). The Commission's related evaluation of these amendments is 
contained in a Safety Evaluation dated November 29, 1994. No 
significant hazards consideration comments received: No

    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

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

    Date of application for amendment: August 25, 1994

    Brief description of amendment: The amendment will revise the Total 
Allowance, Z, S and Allowable Values for the steam generator (SG) level 
instrument calculations. However, the trip setpoints relating to SG 
level will not be changed.

    Date of issuance: December 9, 1994

    Effective date: December 9, 1994
    Amendment No. 52

    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications

    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49425). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 9, 1994. No significant 
hazards consideration comments received: No

    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

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

    Date of application for amendments: September 19, 1994
    Brief description of amendments: The amendments revise the 
Technical Specifications by reducing the frequency for testing the 
containment spray system spray nozzles.
    Date of issuance: November 28, 1994
    Effective date: November 28, 1994
    Amendment Nos.: 159 and 147
    Facility Operating License Nos. DPR-39 and DPR-48. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51618). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 28, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Waukegan Public Library, 128 
N. County Street, Waukegan, Illinois 60085.

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

    Date of application for amendments: August 25, 1994
    Brief description of amendments: The amendments revise the testing 
interval for auxiliary feedwater (AFW) system pumps from monthly to 
quarterly on a staggered test basis. The amendments are consistent with 
the guidance in NUREG-1366, ``Improvements to Technical Specifications 
Surveillance Requirements'' and Generic Letter 93-05, ``Line-Item 
Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.'' In addition, a note 
is incorporated from NUREG-1431, ``Revised Standard Technical 
Specifications, Westinghouse Plants'' into the TS clarifying that the 
turbine-driven AFW pump cannot be tested until the required pressure 
exists in the secondary side of the steam generator.
    Date of issuance: December 8, 1994
    Effective date: To be implemented within 30 days from the date of 
issuance.
    Amendment Nos.: 126 and 120
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: 59 FR 51619 dated 
October 12, 1994. The Commission's related evaluation of the amendments 
is contained in a Safety Evaluation dated December 8, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

GPU Nuclear Corporation, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: September 26, 1994
    Brief description of amendment: The amendment revises the license 
condition regarding the ``Plan for the Long Range Planning Program.'' 
The amendment revises the Plan by changing the semi-annual reporting 
period to annual, and to reflect refined evaluation criteria and 
assessment methodology, and to incorporate necessary changes to the 
license condition wording.
    Date of issuance: November 28, 1994
    Effective date:  As of the date of issuance.
    Amendment No.: 173
    Facility Operating License No. DPR-16. Amendment revised the 
license.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53840). The Commission's related evaluation of this amendment is 
contained in a Safety Evaluation dated November 28, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753.

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

    Date of application for amendment: September 26, 1994
    Brief description of amendment: The amendment revises the plant 
operating license for Three Mile Island, Unit 1 (TMI-1) by changing the 
license condition regarding the ``Plan for the Long Range Planning 
Program.'' The amendment revises the Plan by changing the semi-annual 
reporting frequency to annual, reflects refined evaluation criteria and 
assessment methodology, and incorporates necessary changes to the 
license condition wording.
    Date of issuance: November 28, 1994
    Effective date: As of its date of issuance.
    Amendment No.: 192
    Facility Operating License No. DPR-50. Amendment revised License 
Condition No. 2.C.9.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53841) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location:  Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, PA 17105.

Indiana Michigan Power Company, Docket No. 50-316, Donald C. Cook, 
Nuclear Plant, Unit No. 2, Berrien County, Michigan

    Date of application for amendment: February 22, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) for pressure vessel heatup and cooldown curves and 
extends the applicability from 12 effective full power years (EFPYs) of 
operation to 15 EFPYs. The TS changes are based on an analysis of the 
Cook Unit 2 surveillance capsule U which was removed after exposure of 
8.65 EFPYs.
    Date of issuance:  November 25, 1994
    Effective date: November 25, 1994
    Amendment No.: 171
    Facility Operating License No. DPR-74. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 30, 1994 (59 FR 
14891) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 25, 1994.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085.

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

    Date of application for amendment: August 26, 1994
    Brief description of amendment: The amendment revises the Action 
statements for TS 3.6.1.3, ``Primary Containment Air Locks,'' to allow 
continued plant operation if the containment air lock interlock 
mechanism becomes inoperable, provided an operable door of the air lock 
is locked shut and is verified locked shut at least once per 31 days.
    Date of issuance: November 29, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 59
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 28, 1994 (59 
FR 49431) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 29, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment:  September 9, 1994
    Brief description of amendment: The amendment to the Technical 
Specifications deletes the requirement for a special test of the 
alternate train when one train is inoperable.
    Date of issuance: November 28, 1994
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 76
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53842) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resource Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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 30, 1994
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to change the trip setpoint for the 4kV bus 
undervoltage relay (for the grid degraded voltage) from its current 
value of greater than or equal to 3710 volts to its new setting of 
greater than or equal to 3730 volts.
    Date of issuance: November 30, 1994
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 98
    Facility Operating License No. NPF-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 3, 1994 (59 FR 
39594) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 30, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, Thames Valley Campus, 574 New 
London Turnpike, Norwich, CT 06360.

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: January 14, 1994, as 
supplemented by letters dated March 22, July 14, September 1, October 
21, and November 22, 1994
    Brief description of amendments: The amendments revise TS sections 
5.5.1.1 and 5.5.3, to permit a modification to install new high density 
spent fuel storage racks in each of the spent fuel pools at Limerick. 
The new high density spent fuel storage racks will increase the spent 
fuel pool storage capacity in each spent fuel pool to 4117 fuel 
assemblies.
    Date of issuance:  November 29, 1994
    Effective date: November 29, 1994
    Amendment Nos. 82 and 43
    Facility Operating License Nos. NPF-39 and NPF-85. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 8, 1994 (59 FR 
40376) The supplemental letters provided clarifying information that 
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 November 29, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

Philadelphia Electric 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: April 15, 1994
    Brief description of amendments: These amendments 1) correct a 
typographical error in the Unit 3 TS, 2) reflect the name change of 
Philadelphia Electric Company to PECO Energy Company, and 3) implement 
line-item TS improvements recommended by Generic Letter 93-05, ``Line-
Item Technical Specifications Improvements to Reduce Surveillance 
Requirements for Testing During Power Operation.''
    Date of issuance: November 29, 1994
    Effective date: November 29, 1994
    Amendments Nos.: 199 and 201
    Facility Operating License Nos. DPR-44 and DPR-56: Amendments 
revised the Facility Operating License, Technical Specifications (TS), 
and Environmental TS.
    Date of initial notice in Federal Register: May 25, 1994 (59 FR 
27064) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated November 29, 1994. 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.

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: July 20, 1994, as supplemented 
September 20, 1994
    Brief description of amendment: The amendment changes the TS to to 
allow alternative, equivalent testing of diesel fuel used in the 
emergency diesel generators (EDG). These alternative methods are 
necessary due to recent changes in Environmental Protection Agency 
(EPA) Regulations that are designed to limit the use of high sulfur 
fuels.
    The licensee also proposes to modify the TS by changing the 
revision level of WCAP-10216-P-A, ``Relaxation of Constant Axial Offset 
Control - FQ Surveillance Technical Specification,'' referenced in TS 
6.9.1.11. This pertains to the FQ(z) TS (TS 3.2.1 and 3.2.2) and is 
necessary since Westinghouse revised their methodology in determining 
FQ(z).
    Date of issuance: November 29, 1994
    Effective date: November 29, 1994
    Amendment No.: 121
    Facility Operating License No. NPF-12. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 12, 1994 (59 FR 
51626) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 29, 1994. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

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

    Date of application for amendments: March 18, 1993 (TS 331)
    Brief description of amendments: The proposed changes consist of 
administrative changes to the Technical Specifications for the Browns 
Ferry Nuclear Plant (BFN), Units 1, 2, and 3. The changes include 
deletion of requirements applicable only to BFN Unit 2 Cycle 6 
operation, various administrative error corrections, eliminating 
discrepancies between the Technical Specification Bases and the BFN 
Final Safety Analysis Report, and clarification of certain requirements 
to ensure consistency in application.
    Date of issuance: December 7, 1994
    Effective date: December 7, 1994
    Amendment Nos.: 213, 229 and 186
    Facility Operating License Nos. DPR-33, DPR-52 and DPR-68: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 1, 1993 (58 FR 
17296) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 7, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public Library, South 
Street, Athens, Alabama 35611

Tennessee Valley Authority, Docket No. 50-260, Browns Ferry Nuclear 
Plant, Unit 2, Limestone County, Alabama

    Date of application for amendment: May 11, 1994 (TS 347T)
    Brief description of amendment: The amendment provides a temporary 
extension of the allowed outage time from 5 to 45 days for 250 volt dc 
power supplies which provide control power to the plant shutdown 
boards. This extension will be in effect from January 1, 1995 to 
December 31, 1995 to permit the licensee to upgrade the control power 
supplies with new, higher capacity components.
    Date of issuance: December 7, 1994
    Effective date: December 7, 1994
    Amendment No.: 228
    Facility Operating License No. DPR-52: Amendments revised the 
Technical Specifications.
    Date of initial notice in Federal Register: August 17, 1994 (59 FR 
42347) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 7, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Athens Public library, South 
Street, Athens, Alabama 35611

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

    Date of application for amendments: December 10, 1992, as 
supplemented on March 8, 1994.
    Brief description of amendments: These amendments revised Technical 
Specifications (TS) Section 15.3.5, ``Instrumentation System,'' and 
Section 15.4.1, ``Operational Safety Review.'' Specifically, extensive 
additions and modifications were made to various tables which specify 
requirements for the instrumentation and safety circuits necessary to 
ensure reactor safety and provide for the automatic initiation of the 
engineered safety features.
    Date of issuance: December 8, 1994
    Effective date: December 8, 1994
    Amendment Nos.: 157 & 161
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 25, 1993 (58 FR 
16236) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 8, 1994. The March 8, 
1994, submittal, provided additional supplemental information that did 
not change the initial proposed no significant hazards consideration 
determination. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241

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

    Date of application for amendment: September 17, 1993, as 
supplemented on August 31, 1994.
    Brief description of amendment: The amendment revises the Kewaunee 
Nuclear Power Plant (KNPP) Technical Specifications (TS) by 
incorporating technical and administrative changes to TS 4.5, Emergency 
Core Cooling System and Containment Air Cooling System Tests; TS 4.7, 
Main Steam Isolation Valves; and Table TS 4.1-3, Minimum Frequencies 
for Equipment Tests. Changes have been made to the safety injection 
(SI) system automatic initiation test; the internal containment spray 
system (ICS) flow blockage test; the SI, ICS and residual heat removal 
pumps' periodic tests; the main steam isolation valves' test; and the 
periodic control rod functional test.
    Date of issuance: November 30, 1994
    Effective date: November 30, 1994
    Amendment No.: 114
    Facility Operating License No. DPR-43. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 19, 1994 (59 FR 
2874) The August 31, 1994, submittal, changed the wording of TS 
4.5.a.2.B to specify the number of spray nozzles required for the 
system to function properly. This modification was not outside the 
scope of the original notice and did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated November 30, 1994. No significant hazards consideration comments 
received: No.
    Local Public Document Room location:  University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.

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


    Date of amendment request: July 15, 1994
    Brief description of amendment: This amendment revises Technical 
Specification Table 4.3-3, Radiation Monitoring Instrumentation for 
Plant Operations Surveillance Requirements, to change the analog 
channel operational test interval for selected radiation monitors from 
monthly to quarterly.
    Date of issuance: November 28, 1994
    Effective date: November 28, 1994, to be implemented with 30 days 
of the date of issuance
    Amendment No.: 80
    Facility Operating License No. NPF-42. The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 26, 1994 (59 FR 
53845) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated November 28, 1994. No 
significant hazards consideration comments received: No.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 14th day of December 1994.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[FR Doc. 94-31196 Filed 12-20-94; 8:45 am]
BILLING CODE 7590-01-F