[Federal Register Volume 61, Number 100 (Wednesday, May 22, 1996)]
[Notices]
[Pages 25696-25720]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X96-20522]



=======================================================================
-----------------------------------------------------------------------

UNITED STATES NUCLEAR REGULATORY COMMISSION

Biweekly Notice


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from April 27, 1996, through May 10, 1996. The 
last biweekly notice was published on May 8, 1996 (61 FR 20842).

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 Chief, Rules 
Review and Directives Branch, Division of Freedom of Information and 
Publications Services, Office of Administration, U.S. Nuclear 
Regulatory Commission, Washington, DC 20555-0001, 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 June 21, 1996, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with

[[Page 25697]]

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 20555-0001, 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 20555-0001, 
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.

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 amendments request: April 5, 1996
    Description of amendments request: Pursuant to 10 CFR 50.80 and 
50.90, the Baltimore Gas and Electric Company (BGE) hereby requests the 
transfer and amendment of Operating License Nos. DPR-53 and DPR-69 for 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2.
    The proposed license transfers and amendments are requested as part 
of the pending merger between BGE and Potomac Electric Power Company 
into Constellation Energy Corporation. The proposed license transfers 
would transfer authority to possess and operate Calvert Cliffs from BGE 
to Constellation Energy Corporation. The proposed amendments would 
change the licenses as well as the related Technical Specifications, to 
reflect this transfer by submitting Constellation Energy Corporation in 
place of BGE as the licensee for Calvert Cliffs.
    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. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed amendment will change the name of the licensee 
authorized to possess and operate Calvert Cliffs Nuclear Power Plant 
from Baltimore Gas and Electric Company (BGE) to Constellation 
Energy Corporation. This amendment request is necessary because of a 
proposed merger of BGE and Potomac Electric Power Company into 
Constellation Energy Corporation. As a result of the savings 
achieved through a reduction in operating costs due to the merger, 
Constellation Energy Corporation will have the financial resources 
to possess and operate Calvert Cliffs.
    In addition, Constellation Energy Corporation personnel will be 
technically qualified to operate the plant. Baltimore Gas and 
Electric Company nuclear personnel have been named to management 
positions in Constellation Energy Corporation, and will remain 
responsible for Calvert Cliffs operation and maintenance. The 
proposed amendment involves no changes in the training program or 
operating organization for Calvert Cliffs.
    The proposed amendment does not require any physical change to 
the facilities or substantive modifications to the Technical 
Specifications or to procedures. The proposed change does not 
increase the probability of an accident previously evaluated because 
it does not affect any initiators in any previously evaluated 
accidents. The proposed change does not increase the consequences of 
an accident previously evaluated because it does not affect any of 
the items on which the consequences depend.
    Therefore, the proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Would not create the possibility of a new or different kind 
of accident from any accident previously evaluated.
    The proposed amendment does not modify the plant's configuration 
or operations. As a result, no new accident initiators are 
introduced. Therefore, the proposed amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    This amendment request is necessary because of a proposed merger 
of BGE and Potomac Electric Power Company into Constellation Energy 
Corporation. As a result of the savings achieved through a reduction 
in operating costs due to the merger, Constellation Energy 
Corporation will have the financial resources to possess and operate 
Calvert Cliffs. Also, Constellation Energy Corporation personnel 
will be technically qualified to operate the plant. Baltimore Gas 
and Electric Company nuclear personnel have been named to management 
positions in Constellation Energy Corporation, and will remain 
responsible for Calvert Cliffs' operation and maintenance. The 
proposed amendment involves no changes in the training program or 
operating organization for Calvert Cliffs. In addition, the proposed 
amendment to substitute Constellation Energy Corporation for BGE 
does not result in any changes to the physical design or operation 
of the plant. Therefore, the proposed 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.

[[Page 25698]]

    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Susan Frant Shankman, Acting

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: April 2, 1996
    Description of amendments request: The proposed amendments revise 
the Brunswick Steam Electric Plant, Units 1 and 2, Technical 
Specifications (TS) to allow uprate of the units to 105 percent of 
rated thermal power.
    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:
    . May the proposed activity involve a significant increase in 
the probability or consequences of an accident evaluated previously 
in the Safety Analysis Report?
    The increase in power level, steam flow, feedwater flow and 
associated instrument setpoint changes will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    The probability (frequency of occurrence) of Design Basis 
Accidents occurring is not affected by the increase in power level, 
as plant equipment will remain in compliance with the applicable 
regulatory criteria (ASME Codes, IEEE Standards, NEMA Standards, 
Regulatory Guide criteria, etc.). The physical plant changes 
necessary to support power uprate include instrument setpoint 
changes, indicating meter scale changes for the RWCU [reactor water 
cleanup] System flow and Main Steam Flow indicators, Leak Detection, 
Process Computer, ERFIS [emergency response facility information 
system], and Feedwater System software changes, and SRV [safety/
relief valve] setpoint changes. The setpoints were calculated in 
accordance with the CP&L Setpoint Methodology. Utilizing this 
methodology ensures scram setpoints (instrument settings that 
initiate automatic plant shutdowns) will be established such that 
there is no significant increase in scram frequency due to uprate. 
No new challenges to safety related equipment will result from power 
uprate.
    The changes in consequences of hypothetical accidents which 
would occur from 102% of the uprated power (2609 MWt), compared to 
those previously evaluated from [greater than or equal to] 102% of 
the original power (2485 MWt), are not significant, because the 
accident evaluations at uprated power will not result in exceeding 
the NRC approved acceptance limits. The spectrum of hypothetical 
accidents and transients has been investigated, and those accidents/
transients currently evaluated in the UFSAR [Updated Final Safety 
Analysis Report] were shown to meet the plant's current regulatory 
criteria at uprated conditions (105%). In the area of core design, 
for example, the fuel operating limits will still be met at the 
uprated power level, and fuel reload analyses show plant transients 
will still meet the criteria accepted by the NRC as specified in 
NEDO-24011, ``GESTAR II.'' Challenges to fuel or ECCS [emergency 
core cooling system] performance have been evaluated and shown to 
meet the criteria of 10CFR50 Appendix K. Challenges to the 
containment have been evaluated and still meet 10CFR50 Appendix A 
Criterion 38, Long Term Cooling, and Criterion 50, Containment. 
Bounding events involving radiological releases have been evaluated 
and were shown to be well within the criteria of 10CFR100.
    2. May the proposed activity create the possibility of a new or 
different kind of accident from any accident previously evaluated in 
the Safety Analysis Report?
    The change in reactor thermal power will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Equipment that could be affected by power uprate has been 
evaluated. No new operating mode, safety related equipment lineup, 
accident scenario, or equipment failure mode was identified. The 
full spectrum of accident considerations defined in the BNP 
[Brunswick Nuclear Plant] UFSAR has been evaluated and no new or 
different kind of accident has been identified. Uprate uses 
developed 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 BWRs [Boiling Water Reactors] of 
higher power levels than the uprated power of any of the currently 
uprated BWR/4 fleet and has not identified new power dependent 
accidents.
    The changes to the Technical Specifications required to 
implement power uprate make little change to the plant's 
configuration. These changes fall into three major categories. The 
first includes those changes resulting from power uprate parameter 
changes. These parameter changes, such as the increase in vessel 
pressure, temperature and piping system flows are minor in nature. 
The evaluations have shown the plant is still within its design 
capabilities when operating under these conditions. The changes 
required as a result of power uprate will not affect the design 
function(s) of currently installed equipment; therefore, there is no 
possibility of a new or different kind of failure mode. The second 
set of changes is a result of applying setpoint methodology to 
calculate TS Allowable Values and Normal Trip Setpoints for 
instruments that are directly affected by the parameter changes due 
to power uprate. By using CP&L's methodology, the TS values were 
calculated to ensure adequate margin exists between the analytical 
limit and the TS Allowable Value. The third change include [sic] 
setpoints that were reconstituted by the power uprate project. 
Again, CP&L methodology was applied and the results show the 
setpoints have moved to a more conservative value. This will reduce 
the likelihood of spurious scrams and unnecessary challenges to 
safety systems while ensuring initiation/actuation equipment 
continues to function consistent with existing accident analyses.
    3. Does the proposed activity involve a significant reduction in 
a margin of safety defined in the basis of any Operating License 
Technical Specification?
    Power Uprate will not involve a significant reduction in a 
margin of safety. The bounding events which had been analyzed in the 
UFSAR were reevaluated to demonstrate that power uprate can be 
implemented without exceeding any analyzed limit. Because the 
applicable safety analysis criteria and limits are satisfied for 
power uprate, the margin of safety associated with the safety limits 
and other limits identified in the Technical Specifications will be 
maintained.
    As discussed in Section 5 of GE Nuclear Energy's License Topical 
Report NEDO-31984P ``Generic Evaluations of General Electric Boiling 
Water Reactor Power Uprate,'' the safety margins prescribed by the 
Code of Federal Regulations (CFR) have been maintained by meeting 
the appropriate regulatory criteria. Similarly, the margins provided 
by the application of the ASME design criteria have been maintained. 
The Brunswick unique analysis NEDC-32466P ``Power Uprate Safety 
Analysis Report for Brunswick Steam Electric Plant Units 1 and 2'' 
discusses the effects of power uprate on safety margins for (1) fuel 
thermal limits, (2) design basis accidents and the challenges for 
fuel, containment and radiological releases, (3) transient analysis, 
(4) non-LOCA radiological releases, and (5) environmental 
consequences. These evaluations conclude that applicable safety 
analysis criteria and limits are satisfied, and thus, the margins of 
safety will be maintained.
    The changes to the Technical Specification instrumentation will 
not involve a reduction in the margin of safety. The calculations 
performed for power uprate have established an analytical limit and 
calculated the TS Allowable Value and Nominal Trip Setpoint using 
formal setpoint methodology. This ensures the instrumentation 
functional requirements are met.
    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: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

[[Page 25699]]

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

    Date of amendment request: March 29, 1996
    Description of amendment request: The proposed amendment would 
revise the technical specifications (TS) to add an allowance to 
complete a TS required surveillance within 24 hours of discovery of a 
missed surveillance in accordance with the guidance of Generic Letter 
(GL) 87-09, ``Sections 3.0 and 4.0 of the Standard Technical 
Specifications (STS) on the Applicability of Limiting Conditions for 
Operation and Surveillance Requirements.'' The wording specifying 
intervals for testing has been changed to reflect wording consistent 
the new STS. Typographical errors in the basis are also being 
corrected.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
1. The proposed changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes clarify and incorporates [sic] NRC guidance 
for application of extending or moving surveillance intervals by 
plus or minus 25%, by elimination of restrictive surveillance 
interval descriptions that conflict with NRC guidance, by allowing 
for an additional 24 hours to perform missed surveillances, and by 
providing a defined finite period for the term ``immediate'' for 
Technical Specification (TS) and Inservice Inspection (ISI) 
surveillances. The basis for extending or moving surveillances, as 
stated in GL 89-14, ``Line-Item Improvements in Technical 
Specifications - Removal of the 3.25 Limit on Extending Surveillance 
Intervals,'' is to provide plants flexibility for scheduling the 
performance of surveillances and to permit consideration of plant 
operating conditions that may not be suitable for conducting a 
surveillance at the specified time interval. Such operating 
conditions include transient plant operation or ongoing surveillance 
or maintenance activities. Extending surveillance intervals during 
plant operation can result in a benefit to safety when a scheduled 
surveillances [sic] is due at a time that is not suitable for 
conducting the scheduled surveillance. NUREG-1431, ``Standard 
Technical Specifications - Westinghouse Plants,'' states ``the 25% 
extension does not significantly degrade the reliability that 
results from performing the surveillance at its specified 
frequency.'' This is based on the recognition that the most probable 
result of any particular surveillance being performed is the 
verification of conformance with the surveillance requirements. The 
basis for the 24 hour delay period, as stated in the basis for 
NUREG-1431, includes consideration of unit conditions, adequate 
planning, availability of personnel, the time required to perform 
the surveillance, the recognition that the most probable result of 
any particular surveillance being performed is the verification of 
conformance with the requirements.'' The basis for defining the term 
``immediate'' is to provide guidance to plant personnel for 
conducting operability testing of the Steam Driven Auxiliary 
Feedwater pump after extended shutdown periods in order to minimize 
plant risks and not pose an unsafe operational transient during an 
unstable plant configuration (i.e., during plant startup). Since 
these changes do not affect plant design, operation, or the manner 
in which testing is performed, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes clarify and incorporates [sic] NRC guidance 
for application of extending or moving surveillance intervals by 
plus or minus 25%, by elimination of restrictive surveillance 
interval descriptions that conflict with NRC guidance, by allowing 
for an additional 24 hours to perform missed surveillances, and by 
providing a defined finite period for the term ``immediate'' for TS 
and ISI surveillances. Since these changes do not affect plant 
design, operation, or the manner in which testing is performed, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The changes proposed, with the exception of allowing an 
additional 24 hours to complete missed surveillances, are to clarify 
existing surveillance intervals and to provide more specific and 
detailed criteria without changing current surveillance scheduling 
methodologies. The NRC has determined that allowing an additional 24 
hours to complete missed surveillance tests minimizes additional 
challenges to plant operations such that there is a conservative 
balance between the risk associated with performing the surveillance 
during stable plant conditions and the risk of imposing a plant 
transient due to TS action statements or changing ``modes'' of 
operation. These extensions are current industry practices endorsed 
by the NRC which provide flexibility for scheduling and performing 
surveillances and permit consideration of plant operating conditions 
that may not be suitable for conducting a surveillance at either the 
specified time interval or inadvertently missing the surveillance 
interval. The risk to safety is low in contrast to the alternatives; 
therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Hartsville Memorial Library, 
147 West College Avenue, Hartsville, South Carolina 29550
    Attorney for licensee: William D. Johnson, Vice President and 
Senior Counsel, Carolina Power & Light Company, Post Office Box 1551, 
Raleigh, North Carolina 27602
    NRC Project Director: Eugene V. Imbro

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 8, 1996
    Description of amendment request: The proposed amendments would 
change various sections of the Technical Specifications (TS) to reflect 
the transition of fuel supplier from Generic Electric to Siemens Power 
Corporation (SPC). The amendments would revise the definitions and 
Limiting Conditions for Operation related to Linear Heat Generation 
Rate, Critical Power Ratio, Maximum Critical Power Ratio, and Fraction 
of Limiting Power Density to incorporate SPC terms and methodology or 
to make the TS vendor neutral. Section 6.0 of the TS would be revised 
to include SPC references. The proposed amendment also adds a 
requirement to adjust the Average Planar Linear Heat Generation Rate 
when the reactor is in single loop operation since SPC methodologies 
may require this reduction factor for SPC fuel. The SPC methodologies 
to be added to the TS have previously been approved by the NRC. The 
proposed amendment would also relocate requirements for the traversing 
in-core probe system from the TS to the Core Operating Limits Report 
and would upgrade the fuel description in Section 5.0 as a line item 
from the Improved 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. Involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant

[[Page 25700]]

systems designed to mitigate those consequences. Limits will be 
established consistent with NRC approved methods to ensure that fuel 
performance during normal, transient, and accident conditions is 
acceptable. The proposed Technical Specifications amendment reflects 
previously approved SPC methodology used to analyze normal 
operations, including anticipated operational occurrences (AOOs), 
and to determine the potential consequences of accidents.
    Licensing Methods and Models
    The proposed amendment is to support operation with NRC approved 
fuel and licensing methods supplied from Siemens Power Corporation. 
In accordance with FSAR Chapter 15, the same accidents and 
transients will be analyzed with the new fuel and methods as were 
analyzed by GE for GE fuel. The analysis methods and models are NRC 
approved (Note the mixed core treatment of critical power ratio is 
being addressed under separate correspondence). These approved 
methods and models are used to determine the fuel thermal limits. 
Traversing In-core Probe (TIP) uncertainty are assumptions in the 
approved Siemens core monitoring methodologies. The SPC core 
monitoring code enables the site to monitor keff as well as rod 
density to perform the reactivity anomaly surveillance. This is 
consistent with GE methodology. Therefore, the change in licensing 
analysis methods and models does not significantly increase the 
probability of an accident or the consequences of an accident 
previously identified. The support systems for minimizing the 
consequences of transients and accidents are not affected by the 
proposed amendment.
    New Fuel Design
    The use of ATRIUM 9B fuel at LaSalle does not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated in the FSAR. The ATRIUM-9B fuel is 
generically approved for use as a reload BWR fuel type. (See Boiling 
Water Reactor Licensing Methodology Summary, Siemens Power 
Corporation, EMF-94-217(NP)). Limiting postulated occurrences and 
normal operation have been analyzed using NRC-approved methods for 
the ATRIUM 9B fuel design to ensure that safety limits are protected 
and that acceptable transient and accident performance is 
maintained.
    The reload fuel has no adverse impact on the performance of in-
core neutron flux instrumentation or control rod drive response. The 
ATRIUM-9B fuel design will not adversely affect performance of 
neutron instrumentation nor will it adversely affect the movement of 
control blades. The exterior dimensions of the ATRIUM-9B fuel 
assembly are essentially identical to the GE9B; the ATRIUM-9B fuel 
assembly for LaSalle uses a standard fuel channel and normal control 
cell positioning (i.e., no offset). Thus, no adverse interactions 
with the adjacent control blade and nuclear instrumentation are 
anticipated. Additionally, given the above mentioned overall 
envelope similarities, no problems are anticipated with other 
station equipment such as the fuel storage racks, the new fuel 
inspection stand and the spent fuel pool fuel preparation machine.
    The ATRIUM 9B design is neutronically compatible with the 
existing fuel types and core components in the LaSalle core. SPC 
tests have demonstrated that the ATRIUM-9B fuel design is 
hydraulically compatible with the GE9 fuel. The bundle pressure drop 
characteristics of the ATRIUM 9B bundle are similar to those of the 
GE9 fuel design, hence core thermal-hydraulic stability 
characteristics are not adversely affected by the ATRIUM 9B design.
    An evaluation of the Emergency Procedures is being performed to 
ensure that the use of the ATRIUM-9B fuel at LaSalle does not alter 
any assumptions previously made in evaluating the radiological 
consequences of an accident at LaSalle Station.
    Methods approved by the NRC are being used in the evaluation of 
fuel performance during normal and abnormal operating conditions. 
The ComEd and SPC methods to be used for the cycle specific 
transient analyses have been previously NRC approved. The exception 
is the mixed core treatment of critical power ratio, which is being 
addressed under separate correspondence.
    The description of the fuel is expanded to be consistent with 
NUREG-1434. The description of the fuel materials, lead test 
assembly use, and stating that designs must have been analyzed with 
NRC Staff approved codes does not change existing methods; it only 
describes them.
    Review of the above concludes that the probability of occurrence 
and the consequences of an accident previously evaluated in the 
safety analysis report have not been significantly increased.
    * * * * *
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated:
    Creation of the possibility of a new or different kind of 
accident would require the creation of one or more new precursors of 
that accident. New accident precursors may be created by 
modifications of the plant configuration, including changes in 
allowable modes of operation.
    Licensing Methods and Models
    The proposed Technical Specification amendment reflects 
previously approved SPC methodology used to analyze normal 
operations, including AOOs, and to determine the potential 
consequences of accidents. As stated above, the proposed changes do 
not permit modes of reactor operation which differ from those 
currently permitted.
    New Fuel Design
    The basic design concept of a 9x9 fuel pin array with an 
internal water box has been used in various lead assembly programs 
and in reload quantities in Europe since 1986. WNP-2 has loaded 
reload quantities since 1991. Approximately 650 water box assemblies 
have been irradiated in the United States through 1995, with a 
substantially higher number being irradiated overseas. The NRC has 
reviewed and approved the ATRIUM-9B fuel design. (See Boiling Water 
Reactor Licensing Methodology Summary, Siemens Power Corporation, 
EMF-94-217(NP)). The similarities in fuel design and operation 
indicate there would be no expectation of introducing new or 
different types of accidents than have been considered for the 
existing fuel. Therefore, the use of ATRIUM-9B fuel at LaSalle does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    * * * * *
    3. Involve a significant reduction in the margin of safety for 
the following reasons:
    The existing margin to safety is provided by the existing 
acceptance criteria (e.g., 10CFR50.46 limits). The proposed 
Technical Specification amendment reflects previously approved SPC 
methodology used to demonstrate that the existing acceptance 
criteria are satisfied. The revised methodology has been previously 
reviewed and approved by the USNRC for application to reload cores 
of GE BWRs. References for the Licensing Topical Reports which 
document this methodology, and include the Safety Evaluation Reports 
prepared by the USNRC, are added to the Reference section of the 
Technical Specifications as part of this amendment.
    Licensing Methods and Models
    The proposed amendment does not involve changes to the existing 
operability criteria. NRC approved methods and established limits 
(implemented in the Core Operating Limits Report) ensure acceptable 
margin is maintained. The ComEd and SPC reload methodologies for the 
ATRIUM-9B reload design are consistent with the Technical 
Specification Bases. The Limiting Conditions for Operation are taken 
into consideration while performing the cycle specific and generic 
reload safety analyses. NRC approved methods are listed in 
Specification 6.0 of the Technical Specifications.
    Analyses performed with NRC-approved methodology have 
demonstrated that fuel design and licensing criteria will be met 
during normal and abnormal operating conditions. Therefore, there is 
not a significant reduction in the margin of safety.
    New Fuel Design
    The exterior dimensions of the ATRIUM-9B fuel assembly are 
essentially identical to the GE9B; the ATRIUM-9B fuel assembly for 
LaSalle uses a standard fuel channel and normal control cell 
positioning; i.e., no offset. Thus, no adverse interactions with the 
adjacent control blade and nuclear instrumentation are anticipated. 
The change does not adversely impact equipment important to safety 
and, therefore does not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One

[[Page 25701]]

First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 9, 1996
    Description of amendment request: The proposed amendments would 
eliminate the automatic reactor scram function and the group 1 and 3 
isolation valve closure functions associated with the Main Steam Line 
Radiation Monitoring (MSLRM) system high radiation setpoint. 
Elimination of these functions will eliminate potential spurious scrams 
and isolations caused by increased main steam line radiation levels 
during hydrogen injection. The licensee also proposes to raise the 
MSLRM system alarm setpoints which are not part of the Technical 
Specifications to include increased background radiation during 
hydrogen injection. The proposed amendment would also delete the 
surveillance requirements for the associated instruments.
    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 because:
    Redefining the full power radiation background, thus changing 
the MSLRM alarm setpoint, does not change the probability of 
occurrence of any accident which has been postulated and analyzed in 
the UFSAR, but will reduce the probability of the inadvertent MSIV 
closure transient which is an analyzed transient in the UFSAR. It 
does not change the probability of malfunction of any equipment 
important to safety associated with [loss of coolant accident] LOCA, 
fuel handling accident or [control rod drop accident] CRDA. It also 
does not change the resultant offsite radiological dose from the 
bounding design basis CRDA. This is based upon all radioactivity, 
resulting from the design basis CRDA, going to the condenser 
instantaneously (or independent of the actual MSLRM setpoint) in the 
offsite dose calculation.
    The elimination of reactor scram and isolation of MSIVs, 
isolation of main steam line drain valves and reactor water sample 
line valves, associated with the MSLRM system actuation do not 
introduce, mitigate, or reduce the probability of any design basis 
accident, or any accident, evaluated in the UFSAR. The topical 
report NEDO-31400A has shown that there is essentially no reasonable 
radiological consequence benefit in a design basis CRDA of retaining 
the MSLRM associated reactor scram and MSIV isolation function. In 
addition, the probability of inadvertent scram and isolation is 
reduced. The proposed change will not adversely impact the operation 
of the [reactor protection system] RPS or [primary containment 
isolation system] PCIS with respect to performing its other intended 
safety functions. The proposed change will not affect the operation 
of other plant systems or equipment important to safety. The 
consequences of eliminating the automatic closure of the main steam 
line drain isolation valves and reactor recirculation water sample 
line isolation valves along with the MSIVs has been evaluated to be 
negligible additions to the CRDA doses. A [LaSalle County Station] 
LSCS unique analysis has demonstrated that the radiological doses as 
a result of design basis CRDA are acceptable.
    The MSLRM system high radiation trip was intended to function in 
response to a CRDA which has been previously evaluated. No credit 
for MSIV closure was taken in the CRDA analysis since it postulates 
that all the radioactive material assumed to be released from the 
fuel is transported to the main condenser prior to MSIV closure. 
Furthermore, the probability of a fuel failure is independent of the 
operation of the MSLRM system.
    By eliminating the MSLRM induced MSIV closure, the Offgas system 
can be utilized to reduce potential offsite doses after a CRDA. The 
[mechanical vacuum pump] MVP is tripped no later than 15 minutes of 
a Hi-Hi radiation alarm but analytically results in acceptable 
offsite doses.
    Thus the proposed amendment will not increase the probability of 
any accident previously evaluated, and the elimination of the MSLRM 
isolation signal for MSIVs and other small containment valves will 
not significantly increase the consequences of a CRDA as previously 
evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    Redefining the full power radiation background, thus changing 
the actual MSLRM alarm setpoint, does not alter the configuration of 
the plant. It does not revise any logic or function of the MSLRM 
trip channels or add, replace, or delete any equipment important to 
safety. Therefore it does not introduce any new failure modes or 
create any possibility of a new accident which may challenge safety 
to the public and has not been previously analyzed. It also does not 
involve any equipment which either has not been evaluated 
previously, or may have any safety consequences to the public.
    The proposed Technical Specification changes involve eliminating 
the MSLRM system high radiation trip function for initiating an 
automatic reactor scram, and automatic isolations. The proposed 
changes will not affect the operation of other plant systems or 
equipment important to safety. The MSLRM system will continue to 
initiate alarms as before. Plant procedures will be in place to take 
appropriate mitigative measures in response to a high alarm.
    The isolation and reactor scram functions associated with the 
MSLRM system actuation were originally intended to mitigate, not 
prevent, a potential accident scenario such as a CRDA or gross fuel 
failure event. Adding or removing an electronic signal, such as the 
one from the MSLRM system, does not change system or hardware design 
within the reactor vessel pressure boundary, and therefore will not 
create the possibility of a new or different kind of accident from 
those evaluated in the UFSAR like a LOCA or CRDA during power 
operation. It also does not create the possibility of a new or 
different kind of accident outside the reactor vessel pressure 
boundary from those evaluated in the UFSAR, such as a LOCA or Fuel 
Handling Accident. Removing the isolation signal also reduces the 
probability of inadvertent scram and isolation.
    Therefore the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
analyzed.
    3) Involve a significant reduction in the margin of safety 
because:
    The current MSLRM trip Hi-Hi alarm setpoint (about 4 R/hour with 
full power background at 1.3 R/hour) is at 3 times the full power 
radiation background. As indicated in the plant unique analytical 
result for LSCS, the radiological reading at the MSLRMs for design 
basis CRDA is equivalent to over 1200 times the normal full power 
radiation background (1600 R/hour divided by 1.3 R/hour), or 150 
times the full power radiation background during peak HWC 
environment (since the radiation background is 8 times the normal 
background). Thus the safety margin was very large, and would still 
be quite large with the HWC background factored into the MSLRM 
actuation setpoint (3 x 8 x 1.3 = about 50). The Hi alarm setpoint 
of 1.5 times full power background likewise will have a higher 
safety margin. Thus there is basically no adverse consequence to the 
margin of safety in the basis for the LaSalle technical 
specifications.
    The proposed Technical Specification changes to eliminate the 
MSLRM system high radiation trip function for initiating an 
automatic reactor scram, and automatic closure of the MSIVs, main 
steam line drain isolation valves, and reactor recirculation water 
sample line isolation valves do not cause radiological dose 
consequences to exceed the limit established by SRP 15.4.9.
    Per NEDO-31400A, the elimination of MSLRM trip/scram signal will 
result in the reduction of potential inadvertent scrams, unnecessary 
safety-related actuations, undue vessel isolation, and duty 
challenges during normal plant operation. These can be interpreted 
to be a potential reduction in core damage frequency, which 
translates to an improvement in the margin of safety.
    Thus the margin of safety as defined in the basis of the 
technical specifications is essentially unaffected, and is therefore 
acceptable.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the

[[Page 25702]]

requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

Commonwealth Edison Company, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of amendment request: April 16, 1996
    Description of amendment request: The proposed amendments would 
eliminate the Technical Specification requirement to perform response 
time testing for selected instruments. The instruments affected are the 
sensors for selected reactor protection system instrumentation, main 
steam isolation actuation instrumentation, and all sensors for 
emergency core cooling system (ECCS) actuation instrumentation. The 
proposed changes are supported by analyses performed by the Boiling 
Water Reactor Owners' Group as documented in NEDO-32291-A which was 
approved by the NRC for use in license amendment applications.
    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 because:
    The purpose of the proposed Technical Specification (TS) change 
is to eliminate response time testing requirements for selected 
components in the Reactor Protection System (RPS), Isolation 
Actuation instrumentation and Emergency Core Cooling System (ECCS) 
actuation instrumentation. The Boiling Water Reactor Owners' Group 
(BWROG) has completed an evaluation which demonstrates that response 
time testing is redundant to the other TS-required testing. These 
other tests, in conjunction with actions taken in response to NRC 
Bulletin 90-01, ``Loss of Fill-Oil in Transmitters Manufactured by 
Rosemount,'' and Supplement 1, are sufficient to identify failure 
modes or degradations in instrument response time and ensure 
operation of the associated systems within acceptable limits. There 
are no known failure modes that can be detected by response time 
testing that cannot also be detected by the other TS-required 
testing. This evaluation was documented in NEDO-32291-A, ``System 
Analyses for the Elimination of Selected Response Time Testing 
Requirements,'' dated October 1995. LaSalle County Station, LaSalle, 
has confirmed the applicability of this evaluation to LaSalle. In 
addition, LaSalle will complete the actions identified in the NRC 
staffs safety evaluation of NEDO-32291-A.
    Because of the continued application of other existing TS-
required tests such as channel calibrations, channel checks, channel 
functional tests, and logic system functional tests, the response 
time of these systems will be maintained within the acceptance 
limits assumed in plant safety analyses and required for successful 
mitigation of an initiating event. The proposed changes do not 
affect the capability of the associated systems to perform their 
intended function within their required response time, nor do the 
proposed changes themselves affect the operation of any equipment. 
As a result, LaSalle has concluded that the proposed changes do not 
involve a significant increase in the probability or the 
consequences of an accident previously evaluated.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    The proposed changes only apply to the testing requirements for 
the components identified above and do not result in any physical 
change to these or other components or their operation. As a result 
no new failure modes are introduced. Therefore, the proposed changes 
do not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3) Involve a significant reduction in the margin of safety 
because:
    The current TS-required response times are based on the maximum 
allowable values assumed in the plant safety analyses. These 
analyses conservatively establish the margin of safety. As described 
above, the proposed changes do not affect the capability of the 
associated systems to perform their intended function within the 
allowed response time used as the basis for the plant safety 
analyses. The potential failure modes for the components within the 
scope of this request were evaluated for impact on instrument 
response time. This evaluation confirmed that, with the exception of 
loss of fill-oil of Rosemount transmitters, the remaining TS-
required testing is sufficient to identify failure modes or 
degradations in instrument response times and ensure that operation 
of the applicable instrumentation is within acceptable limits. The 
actions taken in response to NRC Bulletin 90-01 and Supplement 1 are 
adequate to identify loss of fill-oil failures of Rosemount 
transmitters. As a result, it has been concluded that plant and 
system response to an initiating event will remain in compliance 
with the assumptions of the safety analysis.
    Further, although not explicitly evaluated, the proposed changes 
will provide an improvement to plant safety and operation by the 
following:
    a. Reducing the time safety systems are unavailable,
    b. Reducing the potential for safety system actuations,
    c. Reducing plant shutdown risk,
    d. Limiting radiation exposure to plant personnel, and
    e. Eliminating the diversion of key personnel resources to 
conduct unnecessary testing.
    Therefore, LaSalle has concluded that this request will not 
significantly reduce the margin of safety, and may actually cause an 
increase in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
requested amendments involve no significant hazards consideration.
    Local Public Document Room location: Jacobs Memorial Library, 
Illinois Valley Community College, Oglesby, Illinois 61348.
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60603
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: November 2, 1994
    Description of amendment request: The proposed amendments would 
delete the content of Appendix B, ``Environmental Protection Plan'' 
(nonradiological), and modify License Condition 2.C.(2) to delete that 
portion which refers to the Environmental Protection Plan.
    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 amendments would not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated]:
    Deletion of the Environmental Protection Plan and modifying 
License Condition 2.C.(2) will have no impact on the probability or 
consequences of an accident previously evaluated because the changes 
will not have any impact upon the design or operation of any plant 
systems or components.
    2. [The proposed amendments would not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated]:
    The proposed revision will not create the possibility of a new 
or different kind of accident from any previously evaluated because 
the revision is administrative in nature and will not change the 
types and amounts of effluent that will be released.
    3. [The proposed amendments would not involve a significant 
reduction in a margin of safety]:

[[Page 25703]]

    The proposed revision will not reduce a margin of safety because 
it is administrative in nature and will not [a]ffect the margin of 
safety as defined in the basis for any Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Duquesne Light Company, et al., Docket No. 50-412, Beaver Valley 
Power Station, Unit 2, Shippingport, Pennsylvania

    Date of amendment request: April 29, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 5.3.1 to allow the use of ZIRCO as 
an alternate zirconium-based fuel rod material and remove the word clad 
since it has been eliminated from the text of the NRC's improved 
Standard Technical Specifications (NUREG-1431). Limited substitution of 
fuel rods by ZIRCO filler rods would also be permitted. The proposed 
amendment would revise Note 2 on TS Table 3.9-1 to specify that the 
maximum burnup in the peak fuel rod in a fuel assembly stored in Region 
2 spent fuel racks should not exceed the NRC-approved limit for WCAP-
12610 rather than the current maximum burnup limit of 60 GWD/MTU.
    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 methodologies used in the accident analyses remain 
unchanged. The proposed changes do not change or alter the design 
assumptions for the systems or components used to mitigate the 
consequences of an accident. Use of ZIRLO fuel rod material does not 
adversely affect fuel performance or impact nuclear design 
methodology. Therefore, accident analysis results are not impacted.
    The operating limits will not be changed and the analysis 
methods to demonstrate operation within the limits will remain in 
accordance with NRC approved methodologies. Other than the changes 
to the fuel assemblies, there are no physical changes to the plant 
associated with this technical specification change. A safety 
analysis will continue to be performed for each cycle to demonstrate 
compliance with all fuel safety design bases.
    VANTAGE 5 fuel assemblies with ZIRLO fuel rods meet the same 
fuel assembly and fuel rod design bases as other VANTAGE 5 fuel 
assemblies. In addition, the 10 CFR 50.46 criteria are applied to 
the ZIRLO fuel rods. The use of these fuel assemblies will not 
result in a change to the reload design and safety analysis limits. 
Since the original design criteria are met, the ZIRLO fuel rods will 
not be an initiator for any new accident. The fuel rod material is 
similar in chemical composition and has similar physical and 
mechanical properties as Zircaloy-4. Thus, the fuel rod integrity is 
maintained and the structural integrity of the fuel assembly is not 
affected. ZIRLO improves corrosion performance and dimensional 
stability. No concerns have been identified with respect to the use 
of an assembly containing a combination of Zircaloy-4 and ZIRLO fuel 
rods.
    The dose predictions in the safety analyses are not sensitive to 
the fuel rod material used; therefore, the radiological consequences 
of accidents previously evaluated in the safety analysis remain 
valid. A reload analysis is completed for each cycle, in accordance 
with NRC approved methodologies. Therefore, the proposed change does 
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.
    VANTAGE 5 fuel assemblies with ZIRLO fuel rods satisfy the same 
design bases as those used for other VANTAGE 5 fuel assemblies. All 
design and performance criteria continue to be met and no new 
failure mechanisms have been identified. The ZIRLO fuel rod material 
offers improved corrosion resistance and structural integrity.
    The proposed changes do not affect the design or operation of 
any system or component in the plant. The safety functions of the 
related structures, systems, or components are not changed in any 
manner, nor is the reliability of any structure, system, or 
component reduced. The changes do not affect the manner by which the 
facility is operated and do not change any facility design feature, 
structure, or system. No new or different type of equipment will be 
installed. Since there is no change to the facility or operating 
procedures, and the safety functions and reliability of structures, 
systems, or components are not affected, the proposed changes do not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    The use of Zircaloy-4, ZIRLO, or stainless steal filler rods in 
fuel assemblies will not involve a significant reduction in the 
margin of safety because analyses using NRC approved methodology 
will be performed for each configuration to demonstrate continued 
operation within the limits that assure acceptable plant response to 
accidents and transients. These analyses will be performed using NRC 
approved methods that have been approved for application to the fuel 
configuration.
    Use of ZIRLO as fuel rod material does not change the VANTAGE 5 
reload design and safety analysis limits. The use of these fuel 
assemblies will take into consideration the normal core operating 
conditions allowed in the technical specifications. For each reload 
core, the fuel assemblies will be evaluated using NRC approved 
reload design methods, including consideration of the core physics 
analysis peaking factors and core average linear heat rate effects.
    Based on the above, it is concluded that the proposed license 
amendment request does not result in a significant reduction in 
margin with respect to plant safety as defined in the UFSAR [Updated 
Final Safety Analysis Report] or any plant technical specification 
BASES.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 1500l.
    Attorney for licensee: Jay E. Silberg, Esquire, Shaw, Pittman, 
Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

Entergy Operations, Inc., Docket Nos. 50-313 and 50-368, Arkansas 
Nuclear One, Unit Nos. 1 and 2 (ANO-1&2), Pope County, Arkansas

    Date of amendment request: May 2, 1996
    Description of amendment request: The proposed technical 
specification amendments would extend the allowed outage times for 
emergency diesel generators at Arkansas Nuclear One, Units 1 and 2 to 7 
days with an additional, once per refueling cycle extension of 7 more 
days for each machine.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1 - Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The emergency diesel generators (EDGs) are backup alternating 
current power sources

[[Page 25704]]

designed to power essential safety systems in the event of a loss of 
offsite power. The EDGs are not accident initiators in any accident 
previously evaluated. Probabilistic Safety Analysis (PSA) methods 
were utilized in order to fully evaluate the EDG allowed outage time 
(AOT) extension proposed in this submittal. The results of these 
analyses indicate there is not a significant increase in the 
probability of an accident previously evaluated. Therefore, this 
change does not involve an increase in the probability of an 
accident previously evaluated.
    The EDGs provide backup power to components that mitigate the 
consequences of accidents. The current TSs allow for an EDG to be 
removed from service for an AOT. The proposed amendment extends the 
current AOT for an EDG. The proposed change does not allow any more 
equipment to be removed from service at one time. The proposed 
changes to the AOTs do not affect any of the assumptions used in 
deterministic safety analysis. By extending the EDG AOT, the 
consequences of an accident previously evaluated will remain 
unchanged.
    The proposed change removes redundant requirements associated 
with an inoperable emergency power supply from the TS for the 
pressurizer proportional heaters. The operability requirements for 
emergency power supplies and actions to be taken if an EDG is 
inoperable are already addressed in the ANO-2 TS 3.8.1.1.
    The associated changes that remove the requirements to test the 
EDGs if one or both offsite power supplies are inoperable, for an 
inoperable station battery, for an inoperable component in the two 
ESF electrical distribution systems, the accelerated testing 
requirements of the EDGs, and the daily testing requirements for the 
operable EDGs improve the reliability for the operable EDGs by 
reducing the number of unnecessary starts and stops. By improving 
the EDG reliability, this change will not increase the consequences 
of the accidents previously evaluated.
    The other changes in this submittal associated with the bases 
are considered administrative in nature and have no effect on the 
consequences of an accident previously evaluated.
    Therefore, this change does not involve a significant increase 
in the probability or consequences of any accident previously 
evaluated.
    Criterion 2 - Does Not Create the Possibility of a New or 
Different Kind of Accident from any Previously Evaluated.
    This proposed change does not alter the design, configuration, 
or method of operation of the plant. Therefore, this change does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    Criterion 3 - Does Not Involve a Significant Reduction in the 
Margin of Safety.
    The proposed changes do not affect the Technical Specification 
limiting conditions for operation or their bases which support the 
deterministic analyses used to establish the margin of safety.
    Calculations performed to analyze the change in risk based on 
these changes produced acceptable values which are included in the 
tables located in the description of changes section. These 
calculated changes in risk fall well within that which is normally 
considered acceptable. When the additional benefit of maintaining 
the Emergency Diesel Generators available during shutdown cooling 
operations associated with refueling outages in considered, the 
overall change in risk is further reduced.
    The remaining proposed changes are either associated with 
increasing EDG reliability or considered administrative in nature.
    Therefore, this change does not involve a significant reduction 
in the margin of safety
    The NRC staff has reviewed the licensee's analysis and, based on 
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: Tomlinson Library, Arkansas 
Tech University, Russellville, AR 72801
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

Entergy Operations, Inc., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: November 20, 1995, as supplemented by 
the letter dated December 15, 1995.
    Description of amendment request: The licensee has proposed to 
revise the Grand Gulf Nuclear Station (GGNS), Unit 1, Technical 
Specifications (TSs) as follows for the drywell, the drywell airlock, 
and the drywell isolation valves:
    1. For the drywell in Limiting Condition of Operation (LCO) 
3.6.5.1, the surveillance frequency interval for the drywell bypass 
test in Surveillance Requirement (SR) 3.6.5.1.1 would be increased from 
18 months to 10 years. For this interval change, an increased testing 
frequency would be required if bypass performance degrades (i.e., the 
leakage is greater than the limit for two consecutive tests) and the 
application of SR 3.0.2, the allowance to extend the surveillance 
interval by 25 percent, would be restricted to 12 months on the 10-year 
interval. This includes deleting the Note in SR 3.6.5.1.1.
    2. For the drywell airlock in LCO 3.6.5.2, the following changes 
are requested: (a) the leak rate SR 3.6.5.2.2 would be transferred from 
the airlock LCO (3.6.5.2) to SR 3.6.5.1.3 in the drywell LCO (3.6.5.1), 
(b) the requirement in SR 3.6.5.2.2 for the air lock to meet a specific 
overall leakage limit would be deleted, (c) the Note in SR 3.6.5.2.2 
that stated that an inoperable air lock door does not invalidate the 
previous air lock leakage test would be deleted, (d) the test pressure 
for the air lock leakage test in SR 3.6.5.2.2 would be reduced from 
11.5 psig to 3 psid, and (e) the surveillance frequency interval for 
the air lock leakage and interlock testing, required in SRs 3.6.5.2.1 
and 3.6.5.2.2, would be increased from 18 months to 24 months.
    3. For the drywell airlock in LCO 3.6.5.2 and the drywell isolation 
valves in LCO 3.6.5.3, the Action Notes, which identify that the 
actions required by drywell LCO 3.6.5.1 must be taken when the drywell 
bypass leakage limit is not met, would be deleted. Action C.1 of LCO 
3.6.5.2 and its associated completion time would also be deleted.There 
would also be changes to the Bases of the TSs for the above LCOs and 
SRs, based on the proposed changes.
    Basis for proposed no significant hazards consideration 
determination: The amendment request dated November 20, 1995, applied 
to both the Grand Gulf Nuclear Station (GGNS) and the River Bend 
Station (RSB); however, not all of the proposed amendments apply to 
GGNS. This Notice only discusses the amendment request for GGNS. The 
reference below to proposed amendments which do not apply to GGNS are 
marked by ``[....]''.
    As required by 10 CFR 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration in its 
application dated November 20, 1995, which is presented below:
    Entergy Operations, Inc. proposes to change the current Grand 
Gulf Nuclear Station (GGNS) [....] Technical Specifications. The 
specific proposed changes are:
    1. The Surveillance Frequency [interval] for the drywell bypass 
test is changed [increased] from 18 months to 10 years with an 
increased testing frequency required if performance degrades.
    2. The following changes are requested for the drywell air lock 
testing: (a) the leakage rate surveillance is moved from the air 
lock Limiting Condition for Operation (LCO) to the drywell LCO, (b) 
the requirement for the air lock to meet a specific overall leakage 
limit is deleted, (c) the Note that an inoperable air lock door does 
not invalidate the previous air lock leakage test is deleted, (d) 
the GGNS test pressure for the air lock leakage test is changed 
[reduced] from 11.5 psig to 3 psid, [...,] and ([e]) the 
Surveillance Frequency [interval] for the air lock leakage test and 
interlock test is changed [increased] from 18 months to 24 months.
    3. The Actions Notes in the drywell air lock LCO and the drywell 
isolation valve LCO that identifies that the Actions required

[[Page 25705]]

by the drywell LCO must be taken when the drywell bypass leakage 
limit is not met is deleted. [Action C.1 of LCO 3.6.5.2 and its 
associated completion time would also be deleted.]
    [4. ...]
    The Commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 10 CFR 
50.92(c). The proposed changes involve the withdrawal of operating 
restrictions previously imposed because acceptable operation of the 
Mark III primary containment design had not been demonstrated at the 
time of licensing. As published in the Federal Register regarding no 
significant hazards consideration criteria, granting of a relief, 
based upon demonstration of acceptable operation from an operating 
restriction that was imposed because acceptable operation had not 
yet been demonstrated does not involve a significant hazards 
consideration (Ref. 48 FR 14870). Furthermore, a proposed amendment 
to an operating license involves no significant hazards 
consideration if 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.
    Entergy Operations, Inc. has evaluated the no significant 
hazards consideration in its request for this license amendment, 
even though the above-mentioned criterion is satisfied by this 
proposal. In accordance with 10 CFR 50.91(a), Entergy Operations, 
Inc. is providing the analysis of the proposed amendment against the 
three standards in 10 CFR 50.92(c). A description of the no 
significant hazards consideration determination follows:
    I. The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    The requested changes are either administrative changes which 
clarify the format of the requirement or change the requirement to 
match the design bases of the plant, a change which relocates the 
requirement to the Technical Specification Bases, or a change in 
[the] surveillance interval. Each of these types of change are 
discussed below:
    1. The administrative changes clarify the format of the 
requirement or change the requirement to match the design bases of 
the plant. Clarifying [the] administrative format of the Technical 
Specifications does not result in any changes to the Technical 
Specification requirements and, as a result, does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated. Also, changing the requirements of 
the Technical Specifications to more closely match the design bases 
of the plant will continue to assure that the plant will respond as 
assumed in the accident analyses and, as a result, does not involve 
a significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed changes relocate information to the Technical 
Specification Bases. In the Technical Specifications Bases the 
relocated information will be maintained in accordance with 10 CFR 
50.59 and subject to the change control provisions in Chapter 5 of 
Technical Specifications. Since any changes to the Technical 
Specifications Bases will be evaluated per the requirements of 10 
CFR 50.59, no increase (significant or insignificant) in the 
probability or consequences of an accident previously evaluated will 
be allowed. Therefore, this change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    3. The proposed changes in frequency for the drywell bypass 
leakage and drywell air lock surveillances will continue to ensure 
that no paths exist through passive drywell boundary components that 
would permit gross leakage from the drywell to the primary 
containment air space and result in bypassing the primary 
containment pressure-suppression feature beyond the design basis 
limit. The Mark III primary containment system satisfies General 
Design Criterion 16 of Appendix A to 10 CFR Part 50. Maximum drywell 
bypass leakage was determined previously by reviewing the full range 
of postulated primary system break sizes. The limiting case was a 
primary system small break loss of coolant accident (LOCA) and 
yielded a design allowable drywell bypass leakage rate limit of 
approximately 35,000 scfm for GGNS [....]. The Technical 
Specifications acceptable limit for the bypass leakage following a 
surveillance is less than 10% of this design basis value. The most 
recent bypass leakage value was approximately 2.5% for GGNS [....] 
of the design allowable leakage rate limit for the limiting event. 
EOI [Entergy Operations, Inc.] is committed to maintaining 
programmatic and oversight controls that ensure that drywell bypass 
leakage remains a small fraction of the design allowable leakage 
limit.
    The drywell is typically exposed to essentially 0 psig during 
normal plant operation and 3 psig during drywell bypass leak rate 
testing. These pressures are considerably lower than the structural 
integrity test pressure and are less likely to initiate a crack or 
cause an existing crack to grow. Visual inspections of the 
accessible drywell surfaces that have been performed since the 
structural integrity tests have not revealed the presence of 
additional cracking or other abnormalities. Therefore, additional 
cracking of the drywell structure is not expected due to testing or 
operation and, similar to the justification for the ten year 10 CFR 
50 Appendix J Type A test interval, it is not considered credible 
for the passive drywell structure to begin to leak sufficiently to 
impact the design drywell bypass leakage limit.
    The primary containment's ability to perform its safety function 
is fairly insensitive to the amount of drywell leakage, thereby 
providing a margin to loss of the drywell safety function that is 
not normally available for systems. This insensitivity is 
demonstrated by the extremely high limiting event design basis 
allowable leakage for the drywell (e.g., 35,000 scfm for GGNS 
[....]). The limiting leakage is almost an order of magnitude higher 
for other events. Additionally, an even higher allowable leakage can 
be realistically accommodated by the primary containment due to the 
margins in the containment design. Because of the margins available, 
it will take valves in multiple penetration flow paths leaking 
excessively to cause the primary containment to fail as a result of 
overpressurization, the probability that drywell isolation valve 
leakage will result in primary containment failure due to excessive 
drywell leakage is not considered significant and this drywell/
primary containment failure mode is not considered credible.
    The proposed Technical Specification changes have no significant 
impact on the GGNS Individual Plant Examination (IPE) [....] 
conducted per NRC Generic Letter 88-20. The IPEs considered 
overpressurization failure of primary containment as part of the 
primary containment performance assessment. Due to the magnitude of 
acceptable drywell leakage and the extremely low probabilities of 
achieving such leakage, primary containment failure due to 
preexisting excessive drywell leakage was considered a non 
significant contributor to primary containment failure. Primary 
containment overpressurization failure can occur with or without 
preexisting excessive drywell leakage in a severe accident. This is 
due to physical phenomena associated with potentially extreme 
environmental conditions inside primary containment following a 
severe accident. However, the calculated frequency of such extreme 
conditions is very small. The proposed changes do not impact the IPE 
evaluated phenomena causing primary containment overpressurization 
failure nor significantly increase the probability that the drywell 
has preexisting excessive leakage and therefore would not contribute 
to these accident scenarios.
    For the reasons discussed above, the proposed changes do not 
have any significant risk impact to accidents previously evaluated 
and do not significantly increase the consequences of an accident 
previously evaluated. Additionally, drywell leakage is not the 
initiator of any accident evaluated; therefore, changes in the 
frequency of the surveillance for drywell leakage does not increase 
the probability of any accident evaluated.
    Therefore, the proposed changes do not significantly increase 
the probability or consequences of an accident previously evaluated.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested changes are either administrative changes which 
clarify the format of the requirement or change the requirement to 
match the design bases of the plant, a change which relocates the 
requirement to the Technical Specification Bases, or a change in 
surveillance interval. Each of these types of change are discussed 
below:
    1. The administrative changes in the Technical Specification 
requirements do not

[[Page 25706]]

involve a physical alteration of the plant (no new or different type 
of equipment will be installed) nor does it change the methods 
governing normal plant operation. Thus, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    2. The proposed relocation of requirements does not involve a 
physical alteration of the plant (no new or different type of 
equipment will be installed) nor does it change the methods 
governing normal plant operation. The proposed change will not 
impose or eliminate any requirements. Adequate control of the 
information will be maintained in the Technical Specification Bases. 
Thus, the change proposed does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed change modifies the surveillance frequency for 
drywell bypass leakage and drywell air lock surveillances. The 
changes only impact the test frequency and do not result in any 
change in the response of the equipment to an accident. The changes 
do not alter equipment design or capabilities. The changes do not 
present any new or additional failure mechanisms. The drywell is 
passive in nature and the surveillance will continue to verify that 
its integrity has not deteriorated. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    The requested changes are either administrative changes which 
clarify the format of the requirement or change the requirement to 
match the design bases of the plant, a change which relocates the 
requirement to the Technical Specification Bases, or a change in 
surveillance interval. Each of these types of changes are discussed 
below:
    1. The administrative changes in the Technical Specification 
requirements do not involve a physical alteration of the plant (no 
new or different type of equipment will be installed) nor does it 
change the methods governing normal plant operation. Thus, this 
change does not cause a significant reduction in the margin of 
safety.
    2. The relocation of requirements will not reduce a margin of 
safety because it has no impact on any safety analysis assumptions. 
In addition, the requirements to be transferred from the Technical 
Specifications to the Technical Specifications Bases are the same as 
the existing Technical Specifications. Since any future changes to 
these requirements in the Technical Specifications Bases will be 
evaluated per the requirements of 10 CFR 50.59, no reduction 
(significant or insignificant) in a margin of safety will be 
allowed.
    3. The proposed change modifies the surveillance frequency for 
drywell bypass leakage and associated air lock surveillances. 
Reliability of drywell integrity is evidenced by the measured 
leakage rate during past drywell bypass leakage surveillances. 
Appropriate design basis assumptions will be upheld, even when 
combined with the complementary bypass leakage surveillances as 
proposed. Drywell integrity will continue to be tested by means of 
the proposed periodic drywell bypass leakage test, performance of 
the drywell air lock door latching and interlock mechanism 
surveillance, and performance of additional surveillances including 
exercising of drywell isolation valves. The combination of these 
surveillances will provide adequate assurance that drywell bypass 
leakage will not exceed the design basis limit. Margins of safety 
would not be reduced unless leakage rates exceeded the design 
allowable drywell bypass leakage limit. Therefore, the proposed 
change does not cause a significant reduction in the margin of 
safety.
    Therefore, the proposed changes do not cause a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Judge George W. Armstrong 
Library, 220 S. Commerce Street, Natchez, MS 39120
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, N.W., 12th Floor, Washington, DC 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: August 11, 1995, as supplemented by 
letter dated February 12, 1996.
    Description of amendment request: The proposed change will reduce 
the minimum reactor coolant cold leg temperature from 544 Degrees F to 
541 degrees F in Technical Specification Section 3.2.6, ``Reactor 
Coolant Cold Leg Temperature.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change involves a 3 deg.F reduction in the minimum 
core inlet temperature. This change will not have any impact on the 
probability of occurrence of any accident documented in the FSAR.
    The impact of this change on the consequences of events 
documented in the FSAR has been evaluated. The evaluation 
demonstrated that most events are insensitive to the core inlet 
temperature. The events that are impacted by lower core inlet 
temperature are:
    Loss of condenser vacuum (LOCV),
    Part length CEA drop,
    Single CEA withdrawal within deadband, and
    CEA ejection.
    The LOCV event has been reanalyzed for the upcoming Cycle (Cycle 
8) and the results indicate that the peak RCS pressure remains below 
the acceptable limit (110% of the design pressure, i.e., 2750 psia). 
The reactivity anomaly events (remaining events) will be reanalyzed 
as part of COLSS/CPC setpoint calculations. These calculations will 
be performed prior to Cycle 8 startup and will address the impact of 
the 3 deg.F reduction on the minimum core inlet temperature. The 
CPC/COLSS databases and/or addressable constants will be modified, 
as needed due to proposed change, prior to cycle startup.
    A qualitative assessment of the impact of the proposed change on 
the calculated LOCA blowdown loads that are applied to the major 
NSSS components, their supports and the reactor vessel internals was 
also performed. This assessment consisted of an evaluation of the 
design margins on the major components and a determination of the 
impact this lower temperature would have on those margins. The 
evaluation concluded that the impact of a 3 deg.F cold leg 
temperature reduction will be well within the current design 
margins. Therefore, the proposed change will not involve a 
significant increase in the probability or consequences of any 
accident previously evaluated.
    The proposed change to the minimum core inlet temperature does 
not involve any change to any equipment or the manner in which the 
plant will be operated. Since no hardware modifications or changes 
in operation procedures will be made, the proposed change would not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated. Therefore, the proposed change 
will not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The impact of the proposed change on the Waterford 3 FSAR 
analyses have been evaluated. The evaluation showed that the events 
that were impacted were important with respect to RCS pressure and 
fuel thermal limits. One of the events that was impacted by the 
proposed change was the LOCV event. This event was analyzed and the 
results showed that the peak RCS pressure remained below the 
acceptable limit. The impact of this change on other events 
(reactivity anomaly events) will be evaluated as part of the COLSS/
CPC setpoint calculations and the COLSS/CPC databases and/or 
addressable constants will modified as needed to account for any 
adverse impact on the results of these events due to the proposed 
change.
    The impact of this change on the Linear Heat Generation Rate 
limits which varies as a function of the cold leg temperature, is 
accounted for by Technical Specification 3.2.1, ``Linear Heat 
Rate''. The impact of this change on LOCA blowdown loads were 
evaluated to be insignificant compared to the

[[Page 25707]]

current design margins. Therefore, the proposed change will not 
involve a significant reduction in a margin of safety, specifically 
fuel thermal limits and RCS pressure limit.
    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, LA 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

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

    Dates of amendment request: March 20, 1996, and April 23, 1996
    Description of amendment request: The licensee proposed to change 
the Turkey Point Units 3 and 4 Technical Specifications (TS) to 
relocate the requirements for surveillance testing of the water level 
and pressure channel instrumentation for the reactor coolant system 
accumulators and clarify the remaining TS surveillance tests. These 
amendments also modify the existing action statements of TS 3.5.1 for 
accumulators to reflect the requirements of NUREG-1431 by requiring a 
72-hour period to restore boron concentration if it is not within the 
limits, and a 1-hour period to restore any other condition rendering 
the accumulators inoperable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below.
    (1) Operation of the facility in accordance with the proposed 
amendments would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendments do not involve a significant increase in 
the probability or consequences of an accident previously evaluated 
because the proposed amendments conform to the guidance given in 
Enclosure 1 of the NRC GL [Generic Letter] 93-05. The overall 
functional capabilities of the Emergency Core Cooling System (ECCS) 
accumulators will not be modified by the proposed change. This 
amendment will not involve a significant increase in the probability 
or consequences of an accident previously evaluated for the 
following reasons:
    1) The Water Level and Pressure Channel Instrumentation does not 
perform a specific safety function, and merely provides an 
indicating function. The instrumentation in no way affects the 
capability of the accumulators to perform their respective safety 
function.
    2) The changes in most of the ACTION statements are more 
restrictive than current TS requirements due to the one hour vice 
four hour completion time, and therefore will not increase the 
probability or consequences of a previously evaluated accident. If 
one accumulator is inoperable for a reason other than boron 
concentration, the accumulator must be returned to OPERABLE status 
within 1 hour. In this condition, the required contents of three 
accumulators cannot be assumed to reach the core during a Loss Of 
Coolant Accident (LOCA). Due to the severity of the consequences 
should a LOCA occur in these conditions, the 1 hour completion time 
to open the valve, remove power to the valve, or restore the proper 
water volume or nitrogen cover pressure ensures that prompt action 
will be taken to return the inoperable accumulator to OPERABLE 
status. The completion time minimizes the potential for exposure of 
the plant to a LOCA under these conditions. The 1 hour requirement 
for restoring a closed isolation valve is merely a clarification of 
the existing ``immediate'' time requirement.
    3) In the case of low-out-of-specification boron concentration 
in one accumulator, it must be returned to within the limits within 
72 hours. In this condition, ability to maintain subcriticality or 
minimum boron precipitation time may be reduced. The boron in the 
accumulators contributes to the assumption that the combined ECCS 
water in the partially recovered core during the early reflooding 
phase of a large break LOCA is sufficient to keep that portion of 
the core subcritical. One accumulator below the minimum boron 
concentration limit, however, will have no effect on available ECCS 
water and an insignificant effect on core subcriticality during 
reflood. Boiling of ECCS water in the core during reflood 
concentrates boron in the saturated liquid that remains in the core. 
In addition, current Turkey Point analysis demonstrate that the 
accumulators discharge only a small amount following a large main 
steam line break. Therefore, their impact on boron concentration in 
the reactor coolant system is minor and not a design limiting event. 
Thus, 72 hours is allowed to return the boron concentration to 
within limits and does not increase the probability or consequences 
of an accident previously evaluated.
    (2) Operation of the facility in accordance with the proposed 
amendments would not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    The use of the modified specifications can not create the 
possibility of a new or different kind of accident from any 
previously evaluated since the proposed amendments will not change 
the physical plant or the modes of plant operation defined in the 
facility operating license. No new failure mode is introduced due to 
the surveillance changes and clarifications, since the proposed 
changes do not involve the addition or modification of equipment nor 
do they alter the design or operation of affected plant systems.
    (3) Operation of the facility in accordance with the proposed 
amendments would not involve a significant reduction in a margin of 
safety.
    The operating limits and functional capabilities of the affected 
system are unchanged by the proposed amendment. The modified 
specifications which remove surveillance requirements from the TS to 
plant procedures are consistent with the NRC GL 93-05 line-item 
improvement guidance do not significantly reduce any of the margins 
of safety even though the amount of surveillances is decreased. The 
modification of the existing ACTION Statements do not have an 
adverse on [sic] affect on the margin of safety for the following 
reasons:
    1) The SI [Safety Injection] Accumulator Water Level and 
Pressure Channel instrumentation performs no safety function.
    2) The changes in ACTION statements a) and b) are for the most 
part more restrictive than existing TS requirements, the reason 
being the removal of instrumentation requirements for operability.
    3) In the case of low-out-of-specification boron concentration 
in one accumulator, the requirement will be less restrictive, but 
the low boron concentration in one accumulator will have no effect 
on available ECCS water and an insignificant effect on core 
subcriticality during reflood and therefore will not significantly 
reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Florida International 
University, University Park, Miami, Florida 33199
    Attorney for licensee: J. R. Newman, Esquire, Morgan, Lewis & 
Bockius, 1800 M Street, NW., Washington, DC 20036
    NRC Project Director: Frederick J. Hebdon

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

    Date of amendment request: April 19, 1996
    Description of amendment request: The proposed amendment would 
include revisions to Technical Specification (TS) 3.3.6.1, ``Primary 
Containment and Drywell Isolation Instrumentation; TS 
3.3.6.2, ``Secondary Containment Isolation Instrumentation; 
TS 3.3.7.1, ``Control Room Ventilation System 
Instrumentation; TS 3.6.1.2, ``Primary Containment Air 
Locks; TS 3.6.1.3,

[[Page 25708]]

``Primary Containment Isolation Valves; TS 3.6.4.1, 
``Secondary Containment; TS 3.6.4.2, ``Secondary Containment 
Isolation Dampers; TS 3.6.4.3, ``Standby Gas 
Treatment; TS 3.7.3, ``Control Room Ventilation; 
and TS 3.7.4, ``Control Room AC System.'' These TSs would be revised to 
eliminate CORE ALTERATIONS as an applicable condition for which the 
associated Limiting Conditions for Operation (LCO) must be met. 
Consistent changes are also proposed for the associated ACTIONS in each 
of these LCOs, to reflect the changes in the applicable conditions. The 
intent of these proposed changes is to allow certain activities such as 
control rod venting, which is considered a CORE ALTERATION in MODE 5, 
to be performed without the requirements of the identified LCOs being 
met.
    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 eliminate CORE ALTERATIONS as an 
applicable condition requiring operability of the primary and 
secondary containment and control room ventilation system. As stated 
in the BASES for the associated Technical Specifications, 
operability of these systems is primarily required for mitigation of 
the design basis accident - fuel handling accident (DBA-FHA) and 
design basis accident - loss of coolant accident (DBA-LOCA). The 
performance of CORE ALTERATIONS alone is neither a precursor to, nor 
a condition during which these DBAs are postulated to occur. The 
proposed changes only delete CORE ALTERATIONS as an applicable 
condition for the affected Technical Specifications. All other 
applicable MODES or specified conditions, including operations with 
the potential for draining the reactor vessels (OPDRVs) and the 
movement of irradiated fuel assemblies within the primary or 
secondary containment, remain unchanged. Further, the limitations 
placed on the handling of light loads are also unchanged. The 
Technical Specifications (and the separate requirements imposed on 
the handling of light loads) will thus continue to require that 
systems or functions designed to mitigate design-basis/previously 
evaluated accidents are OPERABLE during the relevant operating MODES 
or conditions. On the basis of the above, it is concluded that the 
requested amendment will not increase the probability or 
consequences of any accident previously evaluated.
    2. The proposed changes do not involve any modification to the 
plant design or to the operation of plant systems (except to 
determine when certain analyzed accident-mitigating systems or 
features are required to be OPERABLE). The failure modes considered 
for the proposed changes are the same as those previously 
considered, therefore, it can be concluded that no new failure modes 
will be created. On this basis, the proposed amendment will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The changes being made to eliminate CORE ALTERATIONS as an 
applicable condition for which certain LCOs must be met, do not 
eliminate the requirements for operability of those systems or 
features assumed to mitigate design-basis or analyzed accidents 
during the applicable MODES when such systems or features are 
assumed to be available for performing their mitigating function. 
The safety margins assumed or established by the accident analyses 
for those design-basis events (as described in the accident analyses 
of the Clinton Power Station Updated Final Safety Analysis Report) 
therefore remain unchanged. Further, the proposed changes do not 
impact the controls imposed on the handling of light loads 
(including unirradiated fuel assemblies) for ensuring that such 
activities cannot result in an event that yields consequences more 
severe than those calculated for the DBA-FHA. With respect to 
reactivity concerns during refueling operations (MODE 5), all 
systems or features required to be OPERABLE for precluding 
inadvertent criticality and monitoring reactivity changes will 
continue to be required OPERABLE as per the current Technical 
Specification requirements. The deletion of CORE ALTERATIONS as an 
applicable condition only applies to the noted systems which do not 
contribute to precluding reactivity events. 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: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetener, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendment would 
revise the Clinton Power Station (CPS) Operating License and Technical 
Specifications (TS) to implement 10 CFR Part 50, Appendix J - Option B, 
by referring to Regulatory Guide 1.163, ``Performance-Based Containment 
Leak-Test Program.'' Specifically, changes would be made to paragraph 
2.D of the Operating License; TS Section 1.1, ``Definitions;'' TS 
3.6.1.1, ``Primary Containment;'' TS 3.6.1.1, ``Primary Containment Air 
Locks;'' TS 3.6.1.3, ``Primary Containment Isolation Valves (PCIVs);'' 
and TS Section 5.5, ``Programs and Manuals.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. The proposed change implements new Option B of 10 CFR 50 
Appendix J for performance-based primary containment leakage 
testing. The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any parameters or conditions that contribute to the 
initiation of any accidents previously evaluated. Thus, the proposed 
change cannot increase the probability of any accident previously 
evaluated.
    The proposed change potentially affects the leak-tight integrity 
of the primary containment structure which is designed to mitigate 
the consequences of a loss-of-coolant accident (LOCA) by limiting 
the release of fission products contained in the post-LOCA primary 
containment atmosphere. Functional integrity of the primary 
containment must be maintained during and following the peak 
transient pressures and temperatures that may result from a LOCA. 
Because the proposed change does not alter the plant design, 
including the primary containment and primary containment 
penetrations, and because it only affects the frequency of measuring 
Type A, B, and C leakage without changing the acceptance criteria 
for the Type A, B, and C leakage rate tests, the proposed change 
does not directly result in an increase in the primary containment 
leakage. However, decreasing the test frequency can increase the 
probability that an increase in primary containment leakage could go 
undetected for an extended period of time. To minimize that 
probability, test intervals will be established based on the 
performance history of components being tested.
    NUREG-1493, ``Performance-Based Containment Leak-Test Program,'' 
provides the technical basis for the NRC's rulemaking to revise 
primary containment leakage testing requirements for nuclear power 
reactors in 10 CFR 50, Appendix J. NUREG-1493 documents the NRC's 
determination that the effect of primary containment leakage on 
overall accident risk is minimal since risk is dominated by accident 
sequences that result in failure of bypass of primary containment. 
NUREG-1493 also documents that increasing the Type A leakage test 
intervals would have a minimal impact on public risk, and that Type 
B and C tests can identify the vast majority (greater than ninety 
five percent) of all leakage paths. Therefore, performance-based 
alternatives to current local leakage-testing requirements are 
feasible without significant risk impacts.

[[Page 25709]]

    Based on the above, IP has concluded that the proposed change 
will not result in a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. The proposed change does not involve a change to the plant 
design or operation. As a result, the proposed change does not 
affect any of the parameters or conditions that could contribute to 
initiation of any accidents. This change involves the reduction of 
Type A, B, and C test frequency. Except for the method of defining 
the test frequency, the methods for performing the actual tests are 
not changed. No new accident modes are created by extending the 
testing intervals. No safety-related equipment or safety functions 
are altered as a result of this change. Thus, extending the test 
frequency has no influence on, nor does it contribute to the 
possibility of a new or different kind of accident or malfunction 
from those previously analyzed.
    Based on the above, IP has concluded that the proposed change 
will not create the possibility of a new or different kind of 
accident not previously evaluated.
    3. The request does not involve a significant reduction in a 
margin to safety. The proposed change only affects the frequency of 
the Type A, B, and C testing. Except for the method of defining the 
test frequency, the methods for performing the actual tests are not 
changed. However, the proposed change can increase the probability 
that an increase in primary containment leakage could go undetected 
for an extended period of time. NUREG-1493 has determined that under 
several different accident scenarios, the increased risk of 
radioactivity release from primary containment is negligible with 
the implementation of these proposed changes.
    The margin of safety that has the potential of being impacted by 
the proposed change involves the offsite dose consequences of 
postulated accidents which are directly related to the rate of 
primary containment leakage. The primary containment isolation 
system is designed to limit leakage to La, which is defined by 
the CPS Technical Specifications to be 0.65% of primary containment 
air weight per day at the calculated peak containment internal 
pressure for the design basis loss of coolant accident (Pa). 
The limitation on the rate of primary containment leakage is 
designed to ensure that the total leakage volume will not exceed the 
value assumed in the accident analyses at the peak accident pressure 
(Pa). The margin of safety for the offsite dose consequences of 
postulated accidents directly related to the primary containment 
leakage rate is maintained by continuing to meet the 1.0 La 
acceptance criteria. The La value is not being modified by this 
proposed change.
    Except for the method of defining the test frequency, no change 
in the method of testing is being proposed. The Type A, B, and C 
tests will continue to be done at full pressure (Pa) or 
greater. Other programs are in place to ensure that proper 
maintenance and repairs are performed during the service life of the 
primary containment and systems and components penetrating the 
primary containment.
    As a result, IP has concluded that the proposed change will not 
result in a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Vespasian Warner Public 
Library, 120 West Johnson Street, Clinton, Illinois 61727
    Attorney for licensee: Leah Manning Stetener, Vice President, 
General Counsel, and Corporate Secretary, 500 South 27th Street, 
Decatur, Illinois 62525
    NRC Project Director: Gail H. Marcus

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: January 25, 1996
    Description of amendment request: The amendment proposes to extend 
instrumentation and miscellaneous surveillance test intervals (STI) to 
support 24-month operating cycles. Additionally, this application 
proposes: (1) to revise the Trip Level Settings for Emergency Bus Loss 
of Voltage and Degraded Voltage Instrumentation, (2) to revise the 
Reactor Protection System (RPS) Normal Supply Electrical Protection 
Assembly (EPA) Undervoltage Trip Setpoint, and (3) to make editorial 
revisions, clarification and Bases 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:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment would not involve a significant hazards 
consideration as defined in 10 CFR 50.92, since it would not:
    1. involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed STI changes evaluated in Section IV.A do not 
involve any physical changes to the plant, do not alter the way 
these systems function, and will not degrade the performance of the 
plant safety systems. Proposed instrument setpoint changes ensure 
that plant safety limits are not exceeded due to instrument drift 
predicted for the longer calibration interval. The type of testing 
and the corrective actions required if the subject surveillances 
fail remains the same. The proposed changes do not adversely affect 
the reliability of these systems or affect the ability of the 
systems to meet their design objectives. A historical review of 
surveillance test results supports these conclusions.
    The Trip Level Setpoint changes evaluated in Section IV.B ensure 
that the related systems perform as assumed in the transient and 
accident analysis by ensuring that plant safety limits are not 
exceeded due to instrument drift predicted for the longer 
calibration interval. The changes do not alter the system function, 
and will not degrade the performance of plant safety systems. The 
proposed Trip Level Setting changes do not adversely affect the 
reliability of these systems or adversely affect the ability of 
these systems to meet their design objectives.
    The editorial, clarification and Bases changes evaluated in 
Section IV.C propose enhancements that clarify the Technical 
Specifications requirements and are editorial in nature. These 
changes do not alter any Technical Specification requirement, do not 
involve physical changes to the plant, or alter any operational 
setpoints. There are no safety implications in these proposed 
changes.
    2. create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed STI changes evaluated in Section IV.A do not modify 
the design or operation of the plant, therefore, no new failure 
modes are introduced. Proposed instrument setpoint changes ensure 
that plant safety limits are not exceeded due to instrument drift 
resulting from the longer calibration interval. No changes are 
proposed to the type and method of testing performed, only to the 
length of the surveillance test interval. Past equipment performance 
and on-line testing indicate that longer test intervals will not 
degrade these systems. A historical review of surveillance test 
results supports these conclusions.
    The Trip Level Setpoint changes evaluated in Section IV.B ensure 
that the related systems perform as assumed in the transient and 
accident analysis by ensuring that plant safety limits are not 
exceeded due to instrument drift predicted for the longer 
calibration interval. The changes do not alter the system function, 
introduce any new failure modes, and will not degrade the 
performance of plant safety systems. The proposed Trip Level Setting 
changes do not adversely affect the reliability of these systems or 
adversely affect the ability of these systems to meet their design 
objectives.
    The editorial, clarification and Bases changes evaluated in 
Section IV.C propose enhancements that clarify the Technical 
Specifications requirements and are editorial in nature. These 
changes do not alter any Technical Specification requirement, do not 
involve physical changes to the plant, or alter any operational 
setpoints. There are no safety implications in these proposed 
changes.
    3. involve a significant reduction in a margin of safety.
    Although the proposed STI changes evaluated in Section IV.A will 
result in an increase in the interval between surveillance tests, 
the impact on system reliability is minimal. This is based on more 
frequent on-line testing and the redundant design of the evaluated 
systems. A review of past surveillance history has shown no evidence

[[Page 25710]]

of failures which would significantly impact the reliability of 
these systems. Operation of the plant remains unchanged by these 
proposed STI extensions. The assumptions in the Plant Licensing 
Basis are not adversely impacted. Therefore, the proposed changes do 
not result in a significant reduction in the margin of safety.
    The Trip Level Setpoint changes evaluated in Section IV.B ensure 
that the related systems perform as assumed in the transient and 
accident analysis by ensuring that plant safety limits are not 
exceeded due to instrument drift predicted for the longer 
calibration interval. The changes do not alter the system function, 
introduce any new failure modes, and will not degrade the 
performance of plant safety systems. The proposed Trip Level Setting 
changes do not adversely affect the reliability of these systems or 
adversely affect the ability of these systems to meet their design 
objectives.
    The editorial, clarification and Bases changes evaluated in 
Section IV.C propose enhancements that clarify the Technical 
Specifications requirements and are editorial in nature. These 
changes do not alter any Technical Specification requirement, do not 
involve physical changes to the plant, or alter any operational 
setpoints. There are no safety implications in these proposed 
changes.
    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: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Susan Frant Shankman, Acting

Power Authority of the State of New York, Docket No. 50-333, James 
A. FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of amendment request: April 24, 1996
    Description of amendment request: This amendment proposes to 
relocate Technical Specification (TS) 3.11.B/4.11.B ``Crescent Area 
Ventilation'' and associated Bases from the TS to an Authority 
controlled procedure.
    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:
    Operation of the FitzPatrick plant in accordance with the 
proposed Amendment will not involve a significant hazards 
consideration as defined in 10 CFR 50.92, based on the following:
    (1) These changes do not involve a significant increase in the 
probability or consequences of an accident previously evaluated 
because:
    No modifications, no changes to operating procedure 
requirements, and no reduction in equipment reliability are being 
made as a result of these changes. Operating limitations will 
continue to be imposed, and required surveillance will continue to 
be performed in accordance with regulations, and written procedures 
and instructions that are auditable by the [Nuclear Regulatory 
Commission] NRC. Crescent Area Ventilation operability and testing 
requirements will continue to be an integral part of FitzPatrick 
plant operation.
    Although future changes to the Crescent Area Ventilation system 
will no longer be controlled by 10 CFR 50.90, proposed changes will 
be evaluated under 10 CFR 50.59 and plant procedures. Programmatic 
controls will continue to assure that Crescent Area Ventilation 
system changes will not adversely affect [Emergency Core Cooling 
System] ECCS or [Reactor Core Isolation Cooling] RCIC system 
operability. As such, there is no significant increase in the 
probability or consequences of an accident previously evaluated.
    (2) These changes do not create the possibility of a new or 
different type of accident previously evaluated because:
    No modifications, no changes to operating procedure 
requirements, and no reduction in equipment reliability are being 
made as a result of these changes. Compliance with Crescent Area 
Ventilation system operability and surveillance requirements will be 
assured by maintaining them in an Authority controlled procedure. 
Changes to the Crescent Area Ventilation system will be subject to 
the requirements of 10 CFR 50.59. Therefore, the proposed changes do 
not introduce any failure mechanism of a different type than those 
previously evaluated since there are no changes being made to the 
facility and do not create the possibility of a new or different 
type of accident previously evaluated.
    (3) The proposed amendment does not involve a reduction in a 
margin of safety because:
    The Crescent Area Ventilation system supports Core Spray, [Low 
Pressure Coolant Injection] LPCI mode of [Residual Heat Removal] 
RHR, containment cooling mode of RHR, [High Pressure Coolant 
Injection] HPCI, and RCIC operability, and Crescent Area Ventilation 
system inoperability does affect these systems. As a result, the 
requirement for Crescent Area Ventilation to be operable for these 
systems to be considered operable is implicit in TS Sections 3.5.A, 
3.5.B, 3.5.C, 3.5.E, and the definition of OPERABLE contained in TS 
Section 1.0.J. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment 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: Mr. Charles M. Pratt, 1633 Broadway, New 
York, New York 10019.
    NRC Project Director: Susan Frant Shankman, Acting

Public Service Electric & Gas Company, Docket No. 50-311, Salem 
Nuclear Generating Station, Unit No. 2, Salem County, New Jersey

    Date of amendment request: May 7, 1996
    Description of amendment request: The proposed amendment involves a 
one-time change to Technical Specification (TS) 3/4.7.6, ``Control Room 
Emergency Air Conditioning System.'' The change would permit refueling 
of Salem, Unit 2, with the Control Room Emergency Air Conditioning 
System (CREACS) inoperable in Modes 5 and 6. The change will expire 
after the completion of the Control Room and CREACS upgrade, which is 
currently in progress, and the restart and entry into Mode 4 of Unit 2 
from the current outage.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The CREACS is not an accident initiator. CREACS functions post-
accident to provide cooling for Control Room equipment and 
habitability for operations personnel. Therefore, CREACS has no 
influence on the probability of any of the previously evaluated 
accidents or the other events evaluated as listed below.
    Event
    Fuel Handling Accident (Salem)
    Waste Gas or Volume Control Tank Failures
    Uncontrolled Boron Dilution
    Loss of Offsite Power
    Fuel Handling Accident (Hope Creek)
    Liquid and Gaseous Waste Releases (Hope Creek)
    Loss of Coolant Accident (LOCA) (Hope Creek)
    Chemical Storage
    Barge Collision
    Control Room Internal and External Fire
    Loss of Spent Fuel Pool Cooling
    Loss of Decay Heat Removal
    The Control Area Air Conditioning System (CAACS) and other 
measures will be

[[Page 25711]]

available to maintain Control Room Envelope (CRE) ambient 
temperatures and habitability.
    The proposed one-time change does not impact the consequences of 
an accident previously evaluated based on the following discussions.
    The fuel has decayed to such low levels for more than six months 
that doses associated with the fuel handling accident are well 
within the limits of GDC [General Design Criteria] 19. There is 
insufficient activity remaining in either gaseous waste storage or 
liquid waste storage to force a Control Room evacuation. In the 
event of a Loss of Offsite Power (LOOP), uncontrolled boron dilution 
event, loss of spent fuel pool cooling or loss of decay heat 
removal, CREACS is not required in Modes 5 or 6 to mitigate the 
consequences of this event and CRE habitability will be maintained.
    For a Hope Creek fuel handling accident, gaseous radwaste 
release of LOCA, dose to Salem Control Room personnel will not 
exceed GDC 19 limits. PSE&G [Public Service Electric & Gas] will 
maintain the CAACS [Control Area Air Conditioning System] outside 
air intakes either isolated or capable of being isolated in the 
event of a Hope Creek LOCA. The Hope Creek Event Classification 
Guide (ECG) requires notification of the Salem Control Room in the 
event of an emergency that has the potential to result in a 
radioactive release. The Salem Control Room will isolate the outside 
air intakes if isolation has not already been accomplished.
    For the other events evaluated, the need for evacuation is not 
considered credible for any event with the exception of an internal 
or external fire. However, the possibility of evacuation of the CRE 
in the event of an internal or external fire would be no different 
whether or not CREACS is operating. In the event of an internal 
fire, CAACS will remain in operation to provide purging of the CRE. 
For the case of a possible external fire, the need for evacuation is 
not considered credible because of the short duration of the CREACS 
outage and improbability of the factors which are necessary to 
require an evacuation of the Control Room (i.e. wind direction, wind 
speed, amount of smoke). If an external fire is detected, operator 
action will be taken to isolate the CRE from outside air while CAACS 
remains available. In the unlikely event that the Control Room would 
become uninhabitable due to smoke in the atmosphere, evacuation 
procedures would be followed as in the case of the internal fire.
    The one chemical storage type event which might impact the 
Control Room, rupture of an ammonium hydroxide tanker, is precluded 
by administrative controls such that no ammonium hydroxide tanker 
deliveries will be allowed during the system upgrade period.
    The CAACS will maintain the current design function and TS Bases 
requirements of the CREACS that the ambient air temperature does not 
exceed the allowable temperature for continuous duty rating for 
equipment and instrumentation cooled by the system for the combined 
CRE. The CAACS will be maintained functional while modification to 
the CREACS is ongoing to provide cooling during normal operation and 
under postulated accident conditions. Should the temperature in the 
CRE exceed allowable levels (85 Degrees F), administrative controls 
will be in place to require restoration of the temperature to within 
acceptable levels using CAACS, and prevent any Core Alteration 
activities or positive reactivity changes until the temperature is 
restored to acceptable levels.
    Therefore, the proposed one-time TS change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The CREACS is not an accident initiator. CREACS functions post-
accident to provide cooling for Control Room equipment and 
habitability for operations personnel. Therefore, CREACS 
inoperability during Modes 5 and 6 will not result in the creation 
of a new or different kind of accident from any accident previously 
evaluated. All pertinent accidents have been assessed and no other 
scenarios dealing with fuel movement, or the need for an operable 
CREACS in Mode 5 or 6, have been deemed credible.
    Therefore, the proposed one-time change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed one-time change does not significantly reduce the 
margin of safety as defined in the Bases for the TS because (1) 
there is no credible event as analyzed in Salem UFSAR [updated final 
safety analysis report] Chapter 15 which can cause an unacceptable 
environment in the CRE since the fuel has been decaying for at least 
six months, (2) fuel movement inside the Fuel Handling Building 
(FHB) is restricted in accordance with plant TS unless FHB 
ventilation is operable, (3) dose to Salem control room personnel 
from a potential Hope Creek fuel handling accident, gaseous radwaste 
release or Loss of Coolant Accident will not exceed GDC 19 limits 
(4) the one event which might impact the Control Room, rupture of an 
ammonium hydroxide tanker, is precluded by administrative controls 
such that no ammonium hydroxide tanker deliveries will be allowed 
during the CREACS upgrade period, and (5) in the unlikely event that 
Control Room evacuation is required, there is no impact on operator 
ability to mitigate the consequences of an accident in the current 
plant configuration.
    Therefore, the proposed one-time TS change does not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW, Washington, DC 20005-3502
    NRC Project Director: John F. Stolz

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph 
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: March 29, 1996
    Description of amendment request: The proposed amendment would 
revise Technical Specification 3/4.4.6 ``Steam Generators'' and its 
associated Bases. Specifically, the steam generator repair limit would 
be modified to clarify that the appropriate method for determining 
serviceability for tubes with outside diameter stress corrosion 
cracking at the tube support plate is by a methodology that more 
reliably assesses structural integrity. This amendment request is in 
accordance with NRC's Generic Letter 95-05, ``Voltage-Based Repair 
Criteria for Westinghouse Steam Generator Tubes Affected by Outside 
Diameter Stress Corrosion Cracking.''
    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:
    Operation of Farley units in accordance with the proposed 
license amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Testing of model boiler specimens for free standing tubes at 
room temperature conditions shows burst pressures as high as 
approximately 5000 psi for indications of outer diameter stress 
corrosion cracking with voltage measurements as high as 26.5 volts. 
Burst testing performed on pulled tubes, including tubes pulled from 
Farley Unit 2, with up to 7.5 volt indications show burst pressures 
in excess of 5300 psi at room temperature. As stated earlier, tube 
burst criteria are inherently satisfied during normal operating 
conditions by the presence of the tube support plate. Furthermore, 
correcting for the effects of temperature on material properties and 
minimum strength levels (as the burst testing was done at room 
temperature), tube burst capability significantly exceeds the R.G. 
[Regulatory Guide] 1.121 criterion requiring the maintenance of a 
margin of 1.43 times the steam line break pressure differential on 
tube burst if through-wall cracks are present without regard to the 
presence of the tube support plate. Considering the existing data 
base, this criterion is satisfied with bobbin coil indications with 
signal amplitudes over twice the 2.0 volt voltage-based repair 
criteria, regardless of the indicated depth measurement. This 
structural limit is based on a lower 95% confidence level limit of 
the

[[Page 25712]]

data at operating temperatures. The 2.0 volt criterion provides a 
conservative margin of safety to the structural limit considering 
expected growth rates of outside diameter stress corrosion cracking 
at Farley. Alternate crack morphologies can correspond to a voltage 
so that a unique crack length is not defined by a burst pressure to 
voltage correlation. However, relative to expected leakage during 
normal operating conditions, no field leakage has been reported from 
tubes with indications with a voltage level of under 7.7 volts for a 
3/4 inch tube with a 10 volt correlation to 7/8 inch tubing (as 
compared to the 2.0 volt proposed voltage-based tube repair limit). 
Thus, the proposed amendment does not involve a significant increase 
in the probability or consequences of an accident.
    Relative to the expected leakage during accident condition 
loadings, the accidents that are affected by primary-to-secondary 
leakage and steam release to the environment are Loss of External 
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station 
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator 
Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a 
Control Rod Drive Mechanism Housing. Of these, the Major Secondary 
System Pipe Failure is the most limiting for Farley in considering 
the potential for off-site doses. The offsite dose analyses for the 
other events which model primary-to secondary leakage and steam 
releases from the secondary side to the environment assume that the 
secondary side remains intact. The steam generator tubes are not 
subjected to a sustained increase in differential pressure, as is 
the case following a steam line break event. This increase in 
differential pressure is responsible for the postulated increase in 
leakage and associated offsite doses following a steam line break 
event. In addition, the steam line break event results in a bypass 
of containment for steam generator leakage. Upon implementation of 
the voltage-based repair criteria, it must be verified that the 
expected distributions of cracking indications at the tube support 
plate intersections are such that primary-to-secondary leakage would 
result in site boundary dose within the current licensing basis. 
Data indicate that a threshold voltage of 2.8 volts could result in 
through-wall cracks long enough to leak at steam line break 
conditions. Application of the proposed repair criteria requires 
that the current distribution of a number of indications versus 
voltage be obtained during the refueling outages. The current 
voltage is then combined with the rate of change in voltage 
measurement and a voltage measurement uncertainty to establish an 
end of cycle voltage distribution and, thus, leak rate during steam 
line break pressure differential. The leak rate during a steam line 
break is further increased by a factor related to the probability of 
detection of the flaws. If it is found that the potential steam line 
break leakage for degraded intersections planned to be left in 
service coupled with the reduced allowable specific activity levels 
result in radiological consequences outside the current licensing 
basis, then additional tubes will be plugged or repaired to reduce 
steam line break leakage potential to within the acceptance limit. 
Thus, the consequences of the most limiting design basis accident 
are constrained to present licensing basis limits.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the proposed voltage-based tube support plate 
elevation steam generator tube repair criteria does not introduce 
any significant changes to the plant design basis. Use of the 
criteria does not provide a mechanism that could result in an 
accident outside of the region of the tube support plate elevations. 
Neither a single or multiple tube rupture event would be expected in 
a steam generator in which the repair criteria have been applied 
during all plant conditions. The bobbin probe signal amplitude 
repair criteria are established such that operational leakage or 
excessive leakage during a postulated steam line break condition is 
not anticipated. Southern Nuclear has previously implemented a 
maximum leakage limit of 150 gpd per steam generator. The R.G. 1.121 
criterion for establishing operational leakage limits that require 
plant shutdown are based upon leak-before-break considerations to 
detect a free span crack before potential tube rupture. The 150 gpd 
limit provides for leakage detection and plant shutdown in the event 
of the occurrence of an unexpected single crack resulting in leakage 
that is associated with the longest permissible crack length. R.G. 
1.121 acceptance criteria for establishing operating leakage limits 
are based on leak-before-break considerations such that plant 
shutdown is initiated if the leakage associated with the longest 
permissible crack is exceeded. The longest permissible crack is the 
length that provides a factor of safety of 1.43 against bursting at 
steam line break pressure differential. A voltage amplitude of 
approximately 9 volts for typical outside diameter stress corrosion 
cracking corresponds to meeting this tube burst requirement at the 
95% prediction interval on the burst correlation. Alternate crack 
morphologies can correspond to a voltage so that a unique crack 
length is not defined by the burst pressure versus voltage 
correlation. Consequently, a typical burst pressure versus through-
wall crack length correlation is used below to define the ``longest 
permissible crack'' for evaluating operating leakage limits.
    The single through-wall crack lengths that result in tube burst 
at 1.43 times steam line break pressure differential and steam line 
break conditions are about 0.54 inch and 0.84 inch, respectively. 
Normal leakage for these crack lengths would range from about 0.4 
gallons per minute to 4.5 gallons per minute, respectively, while 
lower 95% confidence level leak rates would range from about 0.06 
gallons per minute to 0.6 gallons per minute, respectively.
    An operating leak rate of 150 gpd per steam generator has been 
implemented. This leakage limit provides for detection of 0.4 inch 
long cracks at nominal leak rates and 0.6 inch long cracks at the 
lower 95% confidence level leak rates. Thus, the 150 gpd limit 
provides for plant shutdown prior to reaching critical crack lengths 
for steam line break conditions at leak rates less than a lower 95% 
confidence level and for three times normal operating pressure 
differential at less than nominal leak rates.
    Considering the above, the implementation of voltage-based 
plugging criteria will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    3) The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the voltage-based tube support plate elevation repair 
criteria is demonstrated to maintain steam generator tube integrity 
commensurate with the requirements of Generic Letter 95-05 and R.G. 
1.121. R.G. 1.121 describes a method acceptable to the NRC staff for 
meeting GDC [Generic Design Criteria] 2, 14, 15, 31, and 32 by 
reducing the probability of the consequences of steam generator tube 
rupture. This is accomplished by determining the limiting conditions 
of degradation of steam generator tubing, as established by 
inservice inspection, for which tubes with unacceptable cracking 
should be removed from service. Upon implementation of the criteria, 
even under the worst case conditions, the occurrence of outside 
diameter stress corrosion cracking at the tube support plate 
elevations is not expected to lead to a steam generator tube rupture 
event during normal or faulted plant conditions. The most limiting 
effect would be a possible increase in leakage during a steam line 
break event. Excessive leakage during a steam line break event, 
however, is precluded by verifying that, once the criteria are 
applied, the expected end of cycle distribution of crack indications 
at the tube support plate elevations would result in minimal, and 
acceptable primary to secondary leakage during the event and, hence, 
help to demonstrate radiological conditions are less than an 
appropriate fraction of the 10 CFR [Part] 100 guideline.
    The margin to burst for the tubes using the voltage-based repair 
criteria is comparable to that currently provided by existing 
Technical Specifications.
    In addressing the combined effects of LOCA [loss-of-coolant 
accident] + SSE [safe-shutdown earthquake] on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case as the tube support plates may become deformed as a result 
of lateral loads at the wedge supports at the periphery of the plate 
due to either the LOCA rarefaction wave and/or SSE loadings. Then, 
the resulting pressure differential on the deformed tubes may cause 
some of the tubes to collapse.
    There are two issues associated with steam generator tube 
collapse. First, the collapse of steam generator tubing reduces the 
RCS [reactor coolant system] flow area through the tubes. The 
reduction in flow area increases the resistance to flow of steam 
from the core during a LOCA which, in turn, may potentially increase 
Peak Clad Temperature (PCT). Second, there is a potential the 
partial through-wall cracks in tubes could progress to through-wall 
cracks during tube deformation or collapse or that short through-

[[Page 25713]]

wall indications would leak at significantly higher leak rates than 
included in the leak rate assessments.
    Consequently, a detailed leak-before-break analysis was 
performed and it was concluded that the leak-before-break 
methodology (as permitted by GDC 4) is applicable to the Farley 
reactor coolant system primary loops and, thus, the probability of 
breaks in the primary loop piping is sufficiently low that they need 
not be considered in the structural design basis of the plant. 
Excluding breaks in the RCS primary loops, the LOCA loads from the 
large branch line breaks were analyzed at Farley and were found to 
be of insufficient magnitude to result in steam generator tube 
collapse or significant deformation.
    Regardless of whether or not leak-before-break is applied to the 
primary loop piping at Farley, any flow area reduction is expected 
to be minimal (much less than 1%) and PCT margin is available to 
account for this potential effect. Based on analyses' results, no 
tubes near wedge locations are expected to collapse or deform to the 
degree that secondary to primary in-leakage would be increased over 
current expected levels. For all other steam generator tubes, the 
possibility of secondary-to-primary leakage in the event of a LOCA + 
SSE event is not significant. In actuality, the amount of secondary-
to-primary leakage in the event of a LOCA + SSE is expected to be 
less than that originally allowed, i.e., 500 gpd per steam 
generator. Furthermore, secondary-to-primary in-leakage would be 
less than primary-to-secondary leakage for the same pressure 
differential since the cracks would tend to tighten under a 
secondary-to-primary pressure differential. Also, the presence of 
the tube support plate is expected to reduce the amount of in-
leakage.
    Addressing the R.G. 1.83 considerations, implementation of the 
tube repair criteria is supplemented by 100% inspection requirements 
at the tube support plate elevations having outside diameter stress 
corrosion cracking indications, reduced operating leakage limits, 
eddy current inspection guidelines to provide consistency in voltage 
normalization, and rotating probe inspection requirements for the 
larger indications left in service to characterize the principle 
degradation mechanism as outside diameter stress corrosion cracking.
    As noted previously, implementation of the tube support plate 
elevation repair criteria will decrease the number of tubes that 
must be taken out of service with tube plugs or repaired. The 
installation of steam generator tube plugs or tube sleeves would 
reduce the RCS flow margin, thus implementation of the voltage-based 
repair criteria will maintain the margin of flow that would 
otherwise be reduced through increased tube plugging or sleeving.
    Considering the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the Final Safety Analysis Report or any 
bases of the plant Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Southern Nuclear Operating Company, Inc., Docket No. 50-364, Joseph 
M. Farley Nuclear Plant, Unit 2, Houston County, Alabama

    Date of amendment request: April 22, 1996
    Description of amendment request: The proposed amendment would 
implement a new F* criterion based on maintaining existing safety 
margins for steam generator tube structural integrity concurrent with 
allowance for NDE (nondestructive examination) eddy current 
uncertainty.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated. The proposed change retains the existing margin in the F* 
distance used to meet regulatory guidance of draft Regulatory Guide 
1.121 and only changes the amount of assumed NDE eddy current 
uncertainty based on the type of eddy current technology utilized in 
the inspection. Therefore, there is no significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any previously evaluated. WCAP 
11306, Revision 2, ``Tubesheet Region Plugging Criterion for the 
Alabama Power Company Farley Nuclear Station Unit 2 Steam 
Generators,'' provides adequate basis for the F* distance proposed 
of 1.54 plus allowance for eddy current uncertainty measurement. 
Since the value of 1.54 inches was used in the analysis no new or 
different kind of accident from any accident previously evaluated 
will be created.
    3. The proposed change does not involve a significant reduction 
in a margin safety. Since the value of 1.54 inches already is used 
in the steam generator tube pull out analysis, there is no 
significant change to a margin safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: M. Stanford Blanton, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: Herbert N. Berkow

Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1, 
Callaway County, MissouriDate of application request: February 23, 
1996, as supplemented by letter dated April 24, 1996.

    Description of amendment request: The amendment would add a 
footnote in the license for Callaway Plant, Unit No. 1 to indicate that 
Union Electric Company has entered into a merger agreement with CIPSCO 
Incorporated which provides for Union Electric Company to become a 
wholly-owned operating company of Ameren Corporation, a registered 
public utility holding company under the Public Utility Holding Company 
Act of 1935, as amended. After the merger, Union Electric Company would 
continue to own and operate the Callaway Plant as an operating company 
subsidiary of Ameren Corporation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not affect accident initiators or 
assumptions. The radiological consequences of any accident 
previously evaluated remain unchanged. The change is an 
administrative change to reflect Union Electric's status as an 
operating company subsidiary of Ameren.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not reduce the margin of safety assumed 
in any accident analysis or affect any safety limits. The change is 
administrative and reflects Union Electric's status as an operating 
company subsidiary of Ameren.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not reduce the margin of safety assumed 
in any accident

[[Page 25714]]

analysis or affect any safety limits. The change is administrative 
and reflects Union Electric's status as an operating company 
subsidiary of Ameren.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts 
& Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: April 30, 1996
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
3.1.b.1, its associated bases, and Figure TS 3.1-4 by extending the low 
temperature overpressure protection (LTOP) requirements through the end 
of operating cycle 33 or 33.41 effective full power years. The only 
technical change being proposed is the substitution of end of life 
fluence for the end of operating cycle 21 fluence.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The LTOP setpoint and revised P/T [pressure/temperature] limits 
reflected in proposed Figure TS 3.1-4 ensure that the Appendix G 
pressure/temperature limits are not exceeded, and therefore, help 
ensure that RCS integrity is maintained. The changes do not modify 
the reactor coolant system pressure boundary, nor make any physical 
changes to the facility design, material, construction standards, or 
setpoints. The LTOP valve setpoint remains set at 500 psi. The LTOP 
enabling temperature based on Figure TS 3.1-2 is 338 deg.F and is 
more conservative than a value of 303 deg. Figure TS 3.1-4. The LTOP 
enabling temperature based on Figure TS 3.1-2 remains unchanged by 
this PA [proposed amendment]. The probability of a LTOP event 
occurring is independent of the pressure-temperature limits for the 
RCS pressure boundary. Therefore, the probability of a LTOP event 
occurring remains unchanged.
    The calculation of pressure temperature limits in accordance 
with approved regulatory methods provides assurance that reactor 
pressure vessel fracture toughness requirements are met and the 
integrity of the RCS [reactor coolant system] pressure boundary is 
maintained. Similar methodology was used in calculations to support 
approved amendment 120 to the Kewaunee Technical Specifications 
dated April 26, 1995. The material property basis, including 
chemistry factor and initial reference temperature for the 
unirradiated material (RTNDT), used for this PA is the same as 
that used in the current TS. The only technical change being made in 
this PA is the use of end of life fluence.
    The use of predicted fluence values through the end of operating 
cycle 33 is appropriately considered within the calculations in 
accordance with standard industry methodology previously docketed 
under WCAP 13227 and WCAP 14279. The neutron exposure projections 
utilized for calculation of the reference temperature were 
multiplied by a factor of 1.11 to adjust for biases observed between 
cycle specific calculations and the results of neutron dosimetry for 
the four surveillance capsules removed from the KNPP reactor. The 
factor of 1.11 was derived by taking the average of the measured to 
calculation (M/C) flux ratios obtained from the dosimetry results of 
capsules V, R, P, and S removed from the KNPP reactor vessel. The 
resulting effect of using predicted fluence values through the end 
of cycle 33 instead of cycle 21 is to require the plant to evaluate 
LTOP transients to more limiting requirements. The proposed PT 
limits are shifted to a lower pressure and higher temperature, which 
is more conservative.
    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 
LTOP relief valve or other safety related system to accidents 
described in Chapter 14 of the USAR. In addition, the changes do not 
alter any assumption previously made in the radiological 
consequences evaluations nor affect the mitigation of the 
radiological consequences of an accident described in the USAR. 
Therefore, the consequences of an accident previously evaluated in 
the USAR will not be increased.
    Thus, the operation of KNPP Unit 1 in accordance with the PA 
does not involve a significant increase in the probability or 
consequences of any accident previously evaluated.
    2. Create the possibility of a new or different type of accident 
from an accident previously evaluated.
    The Appendix G pressure temperature limitations were prepared 
using methods derived from the ASME Boiler and Pressure Vessel Code 
and the criteria set forth in NRC Regulatory Standard Review Plan 
5.3.2. The changes do not cause the initiation of any accident nor 
create any new credible limiting 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 
USAR.
    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 USAR. Thus, the PA does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The use of Paragraph (c)(2)(ii)(A) of 10 CFR 50.61, initial 
reference temperature of -50 deg.F, and the fluence values through 
EOC [end of cycle] 33 does not modify the reactor coolant system 
pressure boundary, nor make any physical changes to the LTOP 
setpoint or system design. Proposed Figure TS 3.1-4 was prepared in 
accordance with regulatory requirements and requires evaluation of 
LTOP events to more limiting requirements of neutron exposure 
projections of 33.41 EFPY instead of 18.40 EFPY.
    Therefore, the PA does not create the possibility of a new or 
different type of accident from any accident previously evaluated.
    3. Involve a significant reduction in the margin of safety.
    The Appendix G pressure temperature limitations were prepared 
using methods derived from the ASME Boiler and Pressure Vessel Code 
and the criteria set forth in NRC Regulatory Standard Review Plan 
5.3.2. These documents along with the calculational limitations 
specified in 10 CFR 50.61 are an acceptable method for implementing 
the requirements of 10 CFR 50 Appendices G and H. Inherent 
conservatism in the P/T limits resulting from these documents 
include:
    a. An assumed defect in the reactor vessel wall with a depth 
equal to 1/4 of the thickness of the vessel wall (1/4T) and a length 
equal to 1-1/2 times the thickness of the vessel wall.
    b. Assumed reference flaw oriented in both longitudinal and 
circumferential directions and limiting material property. At KNPP, 
the only weld in the core region is oriented in the circumferential 
direction.
    c. A factor of safety of 2 is applied to the membrane stress 
intensity factor.
    d. The limiting toughness is based upon a reference value 
(KIR) which is a lower bound on the dynamic crack initiation or 
arrest toughness.
    e. A 2-sigma margin term is applied in determining the adjusted 
reference temperature (ART) that is used to calculate the limiting 
toughness.
    Similar methodology was used in calculations to support approved 
amendment 120 dated April 26, 1995. Beyond the conservatism 
described above, WPSC

[[Page 25715]]

[Wisconsin Public Service Corporation] has incorporated the 
following additional margin in preparing this PA:
    a. The neutron exposure projections were multiplied by a factor 
of 1.11 to adjust for biases observed between cycle specific 
calculations and the results of neutron dosimetry for the four 
surveillance capsules removed from the KNPP reactor. The factor of 
1.11 was derived by taking the average of the measured to 
calculation (M/C) flux ratios obtained from the dosimetry results of 
capsules V, R, P, and S removed from the KNPP reactor vessel.
    b. The calculated material-specific chemistry factor value is 
191.27 and is based on KNPP surveillance capsule data from capsules 
V, R, and P. Utilization of KNPP's most recent surveillance capsule 
data from capsule S results in chemistry factor value of 190.6. 
Consistent with calculation C10689, Revision 1 the value used for 
chemistry factor in this PA remains 191.27, which is conservative.
    c. The LTOP enabling temperature based on Figure TS 3.1-2 is 
338 deg.F and is more conservative than a value of 303 deg.F which 
is supported by proposed Figure TS 3.1-4. The LTOP enabling 
temperature based on Figure TS 3.1-2 remains unchanged by this PA.
    d. The reactor coolant pump starting restrictions of TS 
3.1.a.1.c remain in place.
    An alternative methodology to the safety margins required by 
Appendix G to 10 CFR Part 50 has been developed by the ASME Working 
Group on Operating Plant Criteria. This methodology is contained in 
ASME Code Case N-514. The Code Case N-514 provides criteria to 
determine pressure limits during LTOP events that avoid certain 
unnecessary operational restrictions, provide adequate margins 
against failure of the reactor pressure vessel, and reduce the 
potential for unnecessary activation of the relief valve used for 
LTOP. Specifically, the ASME Code Case N-514 allows determination of 
the setpoint for LTOP events such that the maximum pressure in the 
vessel would not exceed 110% of the P/T limits of the existing ASME 
Appendix G; and redefines the enabling temperature as a coolant 
temperature less than 200 deg.F or a reactor vessel metal 
temperature less than RTNDT + 50 deg.F greater. Code Case N-
514, ``Low Temperature Overpressure Protection,'' has been approved 
by the ASME Code Committee but not yet approved for use in 
Regulatory Guide 1.147. The content of this code case has been 
incorporated into Appendix G of Section XI of the ASME Code and 
published in the 1993 Addenda to Section XI. It is expected that 
when the NRC revises 10 CFR 50.55a, it will endorse the 1993 Addenda 
and Appendix G of Section XI into the regulations. As stated above, 
this PA utilizes Appendix G limits and an enabling temperature 
corresponding to a reactor vessel metal temperature less than 
RTNDT + 90 deg.F, which is more conservative than the 
alternative methodology contained in Code Case N-514.
    The revised calculations meet the NRC acceptance criteria for 
the LTOP setpoint and system design as described in NRC Safety 
Evaluation Report (SER) dated September 6, 1995 which concluded that 
``the spectrum of postulated pressure transients would be 
mitigated...such that the temperature pressure limits of Appendix G 
to 10 CFR 50 are maintained.''
    Utilization of methodology set forth in the ASME Boiler and 
Pressure Vessel Code, NRC Regulatory Standard Review Plan 5.3.2, 10 
CFR 50.61, and 10 CFR 50 Appendices G and H with the above 
additional margins ensures that proper limits and safety factors are 
maintained. Thus, the PA does not involve a significant reduction in 
the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendment would 
revise Kewaunee Nuclear Power Plant (KNPP) Technical Specification (TS) 
4.2.b, ``Steam Generator Tubes,'' its associated bases, and Figure TS 
4.2-1 by redefining the pressure boundary for Westinghouse mechanical 
hybrid expansion joint (HEJ) steam generator (SG) tube sleeves. The 
proposed amendment supersedes in its entirety a previously submitted 
proposed amendment dated October 6, 1995, which was published in the 
Federal Register on November 8, 1995 (60 FR 56372).
    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:
    This proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist.
    1. Operation of the KNPP in accordance with the proposed license 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    Mechanical testing shows inherent structural integrity of the 
HEJ [hybrid expansion joint] upper joint such that the tube rupture 
capability recommendations of RG [Regulatory Guide] 1.121 are met, 
even for instances of 100-percent throughwall, 360 degree 
degradation in the HRLT [hardroll lower transition] region. 
Structural test results are documented in WCAPs-14157, -14157 
Addendum 1, -14446 and -14641. Based on this test data, the 
structural recommendations of RG 1.121 are satisfied when there is a 
difference of at least 0.003 inch, between the maximum hardroll 
diameter of the sleeve, and the diameter at the elevation of the PTI 
[parent tube indication] center line; i.e. there is an interference 
lip of 0.003 inch or more. The proposed pressure boundary will allow 
PTIs located such that there is a minimum diameter change of 0.003 
inch (not including an allowance for measurement uncertainty) 
between the maximum point of the sleeve hardroll, and the diameter 
at the elevation of the PTI peak amplitude to remain in service. 
Based on the high degree of structural integrity of the HEJ upper 
joint, it can be concluded that application of the revised pressure 
boundary criteria will not result in an increased probability of an 
accident previously evaluated.
    Each sleeved tube with a PTI located in the HRLT such that there 
is a change in diameter of 0.003 inch to 0.013 inch, will be 
assigned a conservatively bounding primary-to-secondary SLB [steam 
line break] leakage value of 0.025 gpm per indication. Indications 
located such that there is a change in diameter of greater than 
0.013 inch will not contribute to the SLB leakage. The total number 
of indications remaining in service will be limited such that the 
primary-to-secondary leakage during a postulated SLB will not exceed 
a small fraction of the 10 CFR Part 100 guidelines. For KNPP this 
has been calculated to be 34.0 gpm for the faulted loop. Therefore, 
it can be concluded that application of the revised pressure 
boundary criteria will not increase the consequences of an accident 
previously evaluated.
    2. The proposed license amendment request does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the revised pressure boundary will not 
introduce a change to the design basis or operation of the plant. 
Mechanical testing of degraded sleeve joints supports the 
conclusions that the joint retains structural integrity (tube burst) 
capability consistent with RG 1.121, and leakage integrity with 
regards to a small fraction of the 10 CFR Part 100 guidelines. As 
with the initial installation of the sleeves, implementation of the 
relocated pressure boundary does not interact with other portions of 
the reactor coolant system. Any hypothetical accident as a result of 
potential PTIs is bounded by the existing tube rupture accident 
analysis. Neither the sleeve design nor implementation of the 
redefined pressure boundary affects any other component or location 
of the tube outside of the immediate area repaired. Therefore 
application of the revised pressure boundary criteria will not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    3. The proposed license amendment does not involve a significant 
reduction in the margin of safety.

[[Page 25716]]

    The safety factors used in establishment of the HEJ sleeved tube 
pressure boundary are consistent with the safety factors in the ASME 
Boiler and Pressure Vessel Code used in SG [steam generator] design. 
Based on the sleeve-to-tube geometry, it is unrealistic to consider 
that application of the revised pressure boundary could result in 
single tube leak rates exceeding the normal makeup capacity during 
normal operating conditions. The pressure boundary developed in 
WCAPs-14446 and -14641 have been developed using the methodology of 
RG 1.121. The performance characteristics of the postulated degraded 
parent tubes of HEJ sleeve/tube joints have been verified by testing 
to retain structural integrity and preclude significant leakage 
during normal and postulated accident conditions. Testing indicates 
that postulated circumferentially separated tubes which the pressure 
boundary [addresses] would not experience axial displacement during 
either normal operation or SLB conditions. The existing offsite dose 
evaluation performed for KNPP in support of the voltage based repair 
criteria for axial ODSCC [outside diameter stress corrosion 
cracking] at TSP [tube support plate] intersections established a 
faulted loop primary to secondary leak rate of 34.0 gpm. Following 
implementation of the criteria, postulated leakage from all sources 
must not exceed 34.0 gpm in the faulted loop. Maintenance of this 
limit will ensure that offsite doses would not exceed the currently 
accepted limit of a small fraction of the 10 CFR Part 100 
guidelines. The pressure boundary definition uses a conservatively 
established ``per indication'' leak rate for estimation of SLB 
leakage. This leak rate is applied to all indications left in 
service within the HRLT, regardless of indications length and 
throughwall extent. Application of the revised pressure boundary 
criteria will not result in a significant reduction in the margin of 
safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin, 
Cofrin Library, 2420 Nicolet Drive, Green Bay, Wisconsin 54311-7001.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: July 29, 1994, as superseded by letter 
dated September 15, 1995, and supplements dated March 8, 1996, and 
April 18, 1996
    Description of amendment request: The proposed amendment revises TS 
3/4.8.1 and its associated Bases to improve overall emergency diesel 
generator reliability and availability.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
Emergency diesel generator operability and reliability will continue 
to be assured while minimizing the number of required emergency 
diesel generator starts. Also, emergency diesel generator 
reliability will be enhanced by minimizing severe test conditions 
which can lead to premature failures.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    These proposed changes do not involve a change in the 
operational limits or physical design of the emergency power system. 
The performance capability of the emergency diesel generator will 
not be affected. Emergency diesel generator reliability and 
availability will be improved by the implementation of the proposed 
changes. There is no actual impact on any accident analysis.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    These proposed change do not involve a change in the operational 
limits or physical design of the emergency power system. The 
performance capability of the emergency diesel generator will not be 
affected. Emergency diesel generator reliability and availability 
will be improved by the implementation of the proposed changes. No 
margin of safety is reduced.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

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

    Date of amendment request: May 1, 1996
    Description of amendment request: This license amendment request 
proposes to revise Section 6.0 of the technical specifications to 
reflect position title changes within the Wolf Creek Nuclear Operating 
Corporation (WCNOC) organization.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change does not involve a significant increase in 
the probability of consequences of an accident previously evaluated. 
These changes involve administrative changes to the WCNOC 
organization and to the position qualification of plant personnel.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated. 
This change is administrative in nature and does not involve a 
change to the installed plant systems or the overall operating 
philosophy of Wolf Creek Generating Station.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change does not involve a significant reduction in 
a margin of safety. This change does not involve any changes in 
overall organizational commitments. A position title change alone 
does not reduce the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

[[Page 25717]]

Previously Published Notices Of Consideration Of Issuance Of 
Amendments To Facility Operating Licenses, Proposed No Significant 
Hazards Consideration Determination, And Opportunity For A Hearing

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

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

    Date of amendment request: April 25, 1996
    Brief description of amendment request: The amendment relocates the 
technical specification (TS) Traversing In-Core Probe System Limiting 
Condition for Operation 3/4.3.7.7 and its Bases 3/4.3.7.7 to the 
Technical Requirements Manual, and modifies Note (f) of TS Table 
4.3.1.1-1.
    Date of publication of individual notice in Federal Register: May 
8, 1996 (61 FR 20840)
    Expiration date of individual notice: June 7, 1996
    Local Public Document Room location:  Pottstown Public Library, 500 
High Street, Pottstown, Pennsylvania 19464.

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

    Date of application for amendment: April 26, 1996
    Brief description of amendment request: The proposed amendment 
would correct minor technical and administrative errors in the Improved 
Technical Specifications prior to its implementation.
    Date of individual notice in Federal Register: May 9, 1996 (61 FR 
21213)
    Expiration date of individual notice: June 10, 1996
    Local Public Document Room location: Perry Public Library, 3753 
Main Street, Perry, Ohio

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendments: February 1, 1996
    Brief description of amendments: These amendments revised (1) 
Technical Specifications (TS) 3/4.1.1.1, 6.9.1.9, and 6.9.1.10 to 
relocate the shutdown margin (reactor trip breakers open) to the Core 
Operating Limits Report; (2) TS 3/4.3.2 (Tables 3.3-3 and 3.3-4) to 
specify an additional restriction for the allowed low-pressurizer-
pressure trip setpoint when reducing reactor coolant (RCS) system 
pressure in Mode 3; (3) TS Section 2.2.1 (Table 2.2-1) to make it 
consistent with the footnote in TS Tables 3.3-3 and 3.3-4; and (4) TS 
Sections 3/4.5.2 and 3/4.5.3 to require two emergency core cooling 
system subsystems to be operable in Mode 3 whenever the RCS cold-leg 
temperature is equal to or above 485 deg.F. The Table of Contents and 
the Bases are also revised to reflect these changes.
    Date of issuance: April 30, 1996
    Effective date: April 30, 1996, to be implemented within 45 days of 
issuance
    Amendment Nos.:  Unit 1 - 106; Unit 2 - 98; Unit 3 - 78
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13522) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 30, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004

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

    Date of application for amendments: December 27, 1995
    Brief description of amendments: These amendments modify Tables 
3.3-11 and 4.3-7 of Beaver Valley Power Station, Unit Nos. 1 and 2 
(BVPS-1 and BVPS-2) Technical Specification 3.3.3.8 (Accident 
Monitoring Instrumentation) such that only one valve position 
indication system for the power-operated relief valves and safety 
valves is required to be operable. Minor editorial changes to BVPS-1 TS 
3.3.3.8 and its associated Action Statements are also being made. These 
changes make the requirements of TS 3.3.3.8 consistent with the NRC's 
Improved Standard Technical Specifications (NUREG-1431, Revision 1) and 
with the guidance of Regulatory Guide 1.97, NUREG-0578, and NUREG-0737.
    Date of issuance: May 1, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment Nos.: 199 and 81

[[Page 25718]]

    Facility Operating License Nos. DPR-66 and NPF-73: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3499) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 1, 1996No significant 
hazards consideration comments received: No.
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, Pennsylvania 15001.

Duquesne Light Company, et al., Docket No. 50-334, Beaver Valley 
Power Station, Unit No. 1, Shippingport, Pennsylvania

    Date of application for amendment: February 12, 1996
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 4.6.2.2.d to delete the reference to the specific 
test acceptance criteria for the Containment Recirculation Spray Pumps 
and replaces the specific test acceptance criteria with reference to 
the requirements of the Inservice Testing (IST) Program. In addition, 
the 18-month test frequency is replaced with the test frequency 
requirements specified in the IST Program. The amendment also revises 
the Bases for TS 4.6.2.2.d to describe this revision to TS 4.6.2.2.d.
    Date of issuance: May 7, 1996
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No: 200
    Facility Operating License No. DPR-66. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: March 13, 1996 (61 FR 
10393) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 7, 1996 No significant 
hazards consideration comments received: No
    Local Public Document Room location: B. F. Jones Memorial Library, 
663 Franklin Avenue, Aliquippa, PA 15001

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

    Date of application for amendments: March 21, 1996 as supplemented 
April 8, 15, and 18, 1996.
    Description of amendment request: The proposed amendment provides 
for interim repair criteria for volumetric intergranular attack (IGA) 
indications in the once-through-steam generators (OTSG). The interim 
repair criteria is based on bobbin coil voltage response and motorized 
rotating pancake coil probe dimensional measurements. The amendment 
would be applicable for IGA indications within the region below the 
first tube support plate and the secondary face of the lower tubesheet 
(first span) of the OTSG and for one cycle only until Refuel 11.
    Date of issuance: April 30, 1996
    Effective date: April 30, 1996Amendment Nos. 154
    Facility Operating License No. DPR-72: Amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (61 FR 13888). That notice 
provided an opportunity to submit comments on the Commission's proposed 
no significant hazards consideration determination. No comments have 
been received. The notice also provided for an opportunity to request a 
hearing by April 29, 1996, but indicated that if the Commission makes a 
final no significant hazards consideration determination any such 
hearing would take place after issuance of amendment. The Commission's 
related evaluation of this amendment is contained in a Safety 
Evaluation dated April 30, 1996
    Local Public Document Room location: Coastal Region Library, 8619 
W. Crystal Street, Crystal River, Florida 32629

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

    Date of application for amendment: March 12, 1996 (AEP:NRC:1248)
    Brief description of amendment: The amendment removes the technical 
specifications related to shutdown and control rod position indication 
while in shutdown modes 3, 4, and 5.
    Date of issuance: May 2, 1996
    Effective date: May 2, 1996, with full implementation within 45 
days
    Amendment No.: 194
    Facility Operating License No. DPR-74. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13527) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 2, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Maud Preston Palenske Memorial 
Library, 500 Market Street, St. Joseph, Michigan 49085

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

    Date of amendment request: May 5, 1995 and July 14, 1995, 
supplemented by letter dated March 5, 1996
    Brief description of amendment: The amendment revised the Technical 
Specifications to 1) verify that the redundant diesel generator is 
operable upon the loss of one diesel generator, and implement 
provisions to verify that the operable diesel generator does not have a 
common cause failure; 2) incorporate provisions to allow a modified 
start for the diesel generators; and 3) remove the requirement that the 
reactor power level be reduced to 25% of rated power upon loss of both 
diesel generator units or both incoming power sources (start-up and 
emergency transformers). In addition, the period of time allowed for 
continued reactor operation with both diesels inoperable was reduced 
from 24 to two hours.
    Date of issuance: April 29, 1996
    Effective date: April 29, 1996
    Amendment No.: 175
    Facility Operating License No. DPR-46: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 27, 1995 (60 
FR 49939) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 29, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Auburn Memorial Library, 1810 
Courthouse Avenue, Auburn, NE 68305.

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

    Date of amendment request: September 22, 1995
    Description of amendment request: The amendment changes the ACTION 
specified in Table 3.3-3, Engineered Safety Features Actuation System 
Instrumentation, from ACTION 18 to ACTION 15 for Functional Unit 8.b, 
Automatic Switchover to Containment Sump - RWST Level Low-Low.
    Date of issuance: May 7, 1996,
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 47
    Facility Operating License No. NPF-86. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62493) The Commission's related

[[Page 25719]]

evaluation of the amendment is contained in a Safety Evaluation dated 
May 7, 1996. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Exeter Public Library, 
Founders Park, Exeter, NH 03833.

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

    Date of application for amendment: May 26, 1995, as supplemented 
October 20, 1995, and May 3, 1996.
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 3.8.1.2, ``Electrical Power Systems, Shutdown,'' TS 
3.8.2.2, ``Electrical Power Systems, A.C. Distribution - Shutdown,'' 
and TS 3.8.2.4, ``Electrical Power Systems, D.C. Distribution - 
Shutdown,'' to provide operational flexibility as well as consistency 
between action statements and to eliminate certain surveillance 
requirements that are not applicable in Mode 5 or 6.
    The proposed changes relating to TS 3.8.1.1, ``Electrical Power 
Systems, A.C. Sources, Operating,'' are not included in this amendment 
since this portion of the TS change is still under review by the staff 
and will be addressed at a later date.
    Date of issuance: May 6, 1996
    Effective date: As of the date of issuance, to be implemented 
within 30 days.
    Amendment No.: 197
    Facility Operating License No. DPR-65. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 6, 1995 (60 FR 
62493) The October 20, 1995, letter formally withdrew the need for 
exigent handling of the May 26, 1995, request and requested an 
additional change to TS 3.8.2.4. The May 3, 1996, letter withdrew a 
portion of the initial request which did not affect the initial 
proposed no significant hazards consideration. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 6, 1996. No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Learning Resources Center, 
Three Rivers Community-Technical College, 574 New London Turnpike, 
Norwich, CT 06360, and Waterford Library, ATTN: Vince Juliano, 49 Rope 
Ferry Road, Waterford, CT 06385.

Power Authority of the State of New York, Docket No. 50-286, Indian 
Point Nuclear Generating Unit No. 3, Westchester County, New York

    Date of application for amendment:  March 14, 1996
    Brief description of amendment: The amendment allows a one-time 
extension of the intervals for the pressurizer safety valve setpoint 
and snubber functional testing that is due in May 1996.
    Date of issuance: May 3, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 165
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 3, 1996, (61 FR 
14835) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 3, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location:  White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendments: January 4, 1996
    Brief description of amendments: The amendments change Technical 
Specification 3/4.8.2.5, ``28-Volt D.C. Distribution - Operating.'' The 
amendment for Unit 1 makes Unit 1 requirements similar to Unit 2 by 
defining the specific battery chargers that are required for each train 
and by restricting the use of the backup battery charger to 7 days. The 
amendments for both units also require that the 28-Volt DC bus be 
energized for that bus to be OPERABLE.
    Date of issuance: April 29, 1996
    Effective date: Both units, as of date of issuance, to be 
implemented within 60 days.Amendment Nos. 182 and 163
    Facility Operating License Nos. DPR-70 and DPR-75. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5818) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated April 29, 1996No significant 
hazards consideration comments received: No
    Local Public Document Room location: Salem Free Public Library, 112 
West Broadway, Salem, New Jersey 08079

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

    Date of application for amendment: September 6, 1995, as 
supplemented by letters dated January 30, 1996, March 27, 1996, and 
April 2, 1996.
    Brief description of amendment: The amendment revises TS 5.3.1 to 
reflect a change in the maximum initial enrichment for reload fuel, 
subject to the integral fuel burnable absorber (IFBA) requirements, and 
a change in the maximum fuel enrichment not requiring IFBAs. The 
amendment also changes the maximum reference kinfinity in TS 
5.6.1.1 for fuel storage in Region 1 of the spent fuel pool and revises 
TS Figure 3.9-1 to reflect a change to the maximum initial enrichment 
for fuel stored in Region 2 of the spent fuel pool.
    Date of issuance: April 30, 1996
    Effective date: April 30, 1996, to be implemented within 30 days 
from the date of issuance.
    Amendment No.: 109
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 8, 1995 (60 FR 
56372). The January 30, 1996, March 27, 1996, and April 2, 1996, 
supplemental letters provided additional clarifying information and did 
not change the original no significant hazards consideration 
determination. The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 30, 1996.No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendment: February 9, 1996
    Brief description of amendment: The amendment revised Technical 
Specification 5.3.1 to allow the use of ZIRLO clad fuel rods and ZIRLO 
filler rods.
    Date of issuance: April 30, 1996
    Effective date: April 30, 1996, to be implemented within 30 days of 
issuance.
    Amendment No.: 110
    Facility Operating License No. NPF-30: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7558) The Commission's related

[[Page 25720]]

evaluation of the amendment is contained in a Safety Evaluation dated 
April 30, 1996.No significant hazards consideration comments received: 
No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

Virginia Electric and Power Company, Docket Nos. 50-280 and 50-281, 
Surry Power Station, Unit Nos. 1 and 2, Surry County, Virginia.

    Date of application for amendments: January 30, 1996
    Brief description of amendments: These amendments modify the 
Technical Specifications requirements for the sampling of the reactor 
coolant for dissolved oxygen chlorides and fluorides.
    Date of issuance: 209 and 209
    Effective date: April 29, 1996
    Amendment Nos. 209 and 209
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 27, 1996 (61 FR 
13533) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated April 29, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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

    Date of application for amendment: January 19, 1996, as 
supplemented by letter dated March 19, 1996.
    Brief description of amendment: The amendment modifies the 
Technical Specifications for leak tests of containment isolation 
valves. The amendment replaces the current specified surveillance 
intervals for containment leak testing with new surveillance 
requirements to conduct containment leak testing according to a 
performance-based containment leak test program.
    Date of issuance: May 8, 1996
    Effective date: May 8, 1996, to be implemented within 30 days of 
issuance.
    Amendment No.: 144
    Facility Operating License No. NPF-21: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 14, 1996 (61 
FR 5820) The March 19, 1996, supplemental letter provided additional 
clarifying information and did not change the initial no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
May 8, 1996.No significant hazards consideration comments received: No.
    Local Public Document Room location: Richland Public Library, 955 
Northgate Street, Richland, Washington 99352
    Dated at Rockville, Maryland, this 15th day of May 1996.
    For the Nuclear Regulatory Commission
Steven A. Varga,
Director, Division of Reactor Projects - I/II, Office of Nuclear 
Reactor Regulation
[Doc. 96-12691 Filed 5-21-96; 8:45 am]
BILLING CODE 7590-01-F