[Federal Register Volume 61, Number 109 (Wednesday, June 5, 1996)]
[Notices]
[Pages 28604-28626]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 96-13878]



-----------------------------------------------------------------------


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 May 11, 1996, through May 23, 1996. The last 
biweekly notice was published on May 22, 1996 (61 FR 25696).

Notice Of Consideration Of Issuance Of Amendments To Facility 
Operating Licenses, Proposed No Significant Hazards onsideration 
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 July 5, 1996, the licensee may file a request for a hearing with 
respect to issuance of the amendment to the subject facility operating 
license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714 which is 
available at the Commission's Public Document Room, the Gelman 
Building, 2120 L Street, NW., Washington, DC and at the local public 
document room for the particular facility involved. If a request for a 
hearing or petition for leave to intervene is filed by the above date, 
the Commission or an Atomic Safety and Licensing Board, designated by 
the Commission or by the Chairman of the Atomic Safety and Licensing 
Board Panel, will rule on the request and/or petition; and the 
Secretary or the designated Atomic Safety and Licensing Board will 
issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also

[[Page 28605]]

provide references to those specific sources and documents of which the 
petitioner is aware and on which the petitioner intends to rely to 
establish those facts or expert opinion. Petitioner must provide 
sufficient information to show that a genuine dispute exists with the 
applicant on a material issue of law or fact. Contentions shall be 
limited to matters within the scope of the amendment under 
consideration. The contention must be one which, if proven, would 
entitle the petitioner to relief. A petitioner who fails to file such a 
supplement which satisfies these requirements with respect to at least 
one contention will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, 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.

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendment will 
relocate the administrative controls related to the quality assurance 
review and audit requirements of Section 6 from the Pilgrim Station 
Technical Specifications to the Boston Edison Quality Assurance Manual. 
This change is in accordance with the guidance contained in NRC 
Administrative Letter 95-06, ``Relocation of Technical Specification 
Administrative Controls Related to Quality Assurance.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The change will relocate the administrative controls related to 
the quality assurance review and audit requirements from the 
technical specifications to the quality assurance plan. These 
changes are administrative in nature and do not impact initiators of 
analyzed events, accident mitigation capabilities, or transient 
events. The quality assurance program is a logical candidate for 
such relocation due to the controls imposed by such regulations as 
Appendix B to 10 CFR [Part] 50, the existence of NRC approved 
quality assurance plans and commitments to industry quality 
assurance standards, and the established quality assurance program 
change control process in 10 CFR 50.54(a). Therefore, the changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The change will relocate the administrative controls related to 
the quality assurance review and audit requirements from the 
technical specifications to the quality assurance plan. The quality 
assurance program is a logical candidate for such relocation due to 
the controls imposed by such regulations as Appendix B to 10 CFR 
[Part] 50, the existence of NRC approved quality assurance plans and 
commitments to industry quality assurance standards, and the 
established quality assurance program change control process in 10 
CFR 50.54(a). The proposed changes do not involve a physical 
alteration of the plant or changes in methods governing plant 
operation. The changes will not impose or eliminate any new or 
different requirements. Therefore the changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The change will relocate the administrative controls related to 
the quality assurance review and audit requirements from the 
technical specifications to the quality assurance plan. These 
changes are administrative in nature. The quality assurance program 
is a logical candidate for such relocation due to the controls 
imposed by such regulations as Appendix B to 10 CFR [Part] 50, the 
existence of NRC approved quality assurance plans and commitments to 
industry quality assurance standards, and the established quality 
assurance program change control process in 10 CFR 50.54(a). The 
proposed change will not reduce a margin of safety because it has no 
impact on any safety analysis assumptions. Therefore, the operation 
of PNPS [Pilgrim Nuclear Power Station] in accordance with the 
proposed license amendment will not involve a significant reduction 
in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Jocelyn A. Mitchell, Acting

[[Page 28606]]

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendment will 
reflect the implementation of 10 CFR Part 50, Appendix J, Option B at 
the Pilgrim Nuclear Power Station.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed changes do not involve any physical or operational 
changes to structures, systems or components. The proposed changes 
provide a mechanism within the TS [Technical Specifications] for 
implementing a performance-based leakage rate test program which was 
promulgated by the revision to 10CFR50 to incorporate Option B into 
Appendix J. The TS Limiting Conditions for Operation (LCO) remain 
unaffected by these changes. Thus, the safety design basis for the 
accident mitigation functions of the primary containment is 
maintained. Therefore, these changes will not increase the 
probability or consequences of an accident previously evaluated.
    2. The operation of Pilgrim Station in accordance with the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    Revising surveillance requirement acceptance criteria and 
frequencies does not physically modify the plant and does not modify 
the operation of any existing equipment. Further, the TS LCOs remain 
unaffected by these changes.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in a 
margin of safety.
    The proposed changes do not involve a significant reduction in 
the margin of safety, nor do they affect a safety limit, an LCO, or 
the manner in which plant equipment is operated. The NRC letter 
dated November 2, 1995, recognizes that changes similar to the 
proposed changes are required to implement Option B of 10CFR50, 
Appendix J. In NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' which forms the basis for the Appendix J revision, the 
NRC concludes that adoption of performance-based test intervals for 
Appendix J testing 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:  Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Jocelyn A. Mitchell, Acting

Boston Edison Company, Docket No. 50-293, Pilgrim Nuclear Power 
Station, Plymouth County, Massachusetts

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendment would 
modify the definition of ``Core Alteration,'' and the Limiting 
Condition for Operation, Surveillance conditions and Bases section 
associated with Technical Specification (TS) 3.7.C, ``Secondary 
Containment.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Operation of PNPS [Pilgrim Nuclear Power Station] in accordance 
with the proposed license amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated because of the following:
    Proposed Change <greek-i>1: Definition of ``Alteration of the 
Reactor Core<gr-thn-eq>
    The definition, ``Alteration of the Reactor Core'', is being 
revised so that the term will apply only to those activities that 
create the potential for a reactivity excursion and, therefore, 
warrant special precautions or controls in the TS. The proposed 
definition includes normal control rod movement in the definition, 
but excludes control rod drive movement (such as rod removal from 
the core) when all four fuel bundles surrounding a control rod are 
removed. The proposed change does not increase the probability or 
consequences of an accident because the proposed definition, by 
identifying activities with the potential for causing a reactivity 
excursion, ensures that the additional precautions and controls in 
the TS are implemented at all appropriate times. In addition, the 
movement of components excluded by this definition is not assumed in 
the initiation of any analyzed event. Therefore, the proposed change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Proposed Change <greek-i>2: Secondary Containment
    The current specifications are revised to specify more clearly 
when secondary containment is required, what actions to take if 
secondary containment is inoperable, and time frames for completing 
the actions. These revisions enhance the existing specification and 
serve to make it more definitive by encompassing the conditions 
currently specified by TS and supplementing them to specify other 
conditions when secondary containment is required.
    Surveillances 4.7.C.1.a and b were only necessary during initial 
and Cycle 1 operations. Removing obsolete information from the 
existing specifications, re-numbering and re-arranging the wording 
is an administrative change.
    These changes are administrative in nature and do not impact 
initiators of analyzed events, accident mitigation capabilities, or 
transient events. Therefore, the changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. The proposed amendment does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The operation of PNPS in accordance with the proposed license 
amendment will not create the possibility of a new or different kind 
of accident from any accident previously evaluated because of the 
following:
    Proposed Change <greek-i>1: Definition of ``Alteration of the 
Reactor Core<gr-thn-eq>
    The definition change specifies more accurately which component 
movements constitute a ``Core Alteration''. This change does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or changes in methods governing 
normal plant operation. The proposed changes will allow movement of 
some components (camera, lights, etc.) during times when ``Core 
Alterations'' have been halted since these components will not 
affect core reactivity. Removal of a control rod involves unlatching 
and withdrawal/insertion from over-vessel handling equipment. These 
activities necessitate, by design, the removal of the adjacent four 
fuel assemblies. With this configuration (no fuel in the cell; 
handling the associated control rod), the proposed change will allow 
movement of a ``reactivity control component'' while not imposing 
requirements unique to ``Core Alterations'' (note: other 
requirements, such as those for handling loads over irradiated fuel, 
will remain applicable). The reactivity effects of this control rod 
movement are more than compensated for by the initial removal of the 
fuel assemblies. Therefore, this change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Proposed Change <greek-i>2: Secondary Containment
    The proposed change does not eliminate or relax any existing TS 
condition. Rather, it better defines when secondary containment is 
required, provides action statements for inoperability and removes 
obsolete

[[Page 28607]]

requirements (from first operating cycle). This change does not 
involve a physical change to structures, systems or components, and 
the safety design bases for the accident mitigating function of the 
secondary containment is maintained. Therefore, these changes will 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    3. The proposed amendment does not involve a significant 
reduction in a margin of safety.
    The operation of PNPS in accordance with the proposed license 
amendment will not involve a significant reduction in a margin of 
safety because of the following:
    Proposed Change <greek-i>1: Definition of ``Alteration of the 
Reactor Core<gr-thn-eq>
    The proposed definition more accurately identifies those 
activities with the potential for causing a reactivity excursion. 
The more accurate identification of ``Core Alterations'' will ensure 
that when there is a potential for reactivity excursions, 
appropriate precautions are applied. The components now excluded 
from the proposed definition are those that do not have the 
capability for adversely impacting core reactivity. The proposed 
change has no impact on safety analysis assumptions. Therefore, the 
change will not involve a significant reduction in a margin of 
safety.
    Proposed Change <greek-i>2: Secondary Containment
    The proposed additions of applicability conditions provide a 
more precise understanding of when secondary containment integrity 
is required and what actions to take if it becomes inoperable. The 
change does not eliminate any existing conditions. The deletion of 
surveillances applicable only for the first operating cycle and re-
numbering and re-arranging the remaining surveillance wording is an 
administrative change and has no impact on the operation of the 
plant or mitigation of accidents. Therefore, the operation of the 
facility in accordance with this proposed amendment would not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Plymouth Public Library, 11 
North Street, Plymouth, Massachusetts 02360.
    Attorney for licensee: W. S. Stowe, Esquire, Boston Edison Company, 
800 Boylston Street, 36th Floor, Boston, Massachusetts 02199.
    NRC Project Director: Jocelyn A. Mitchell, Acting

Carolina Power & Light Company, et al., Docket No. 50-325, 
Brunswick Steam Electric Plant, Unit 1, Brunswick County, North 
Carolina

    Date of amendment request: April 8, 1996
    Description of amendment request: The licensee has proposed to 
revise the Technical Specifications (TS) to include the following 
changes: 1. The Minimum Critical Power Ratio (MCPR) Safety Limit 
specified in TS 2.1.2 from 1.07 to 1.09 for Unit 1 Cycle 11 operation; 
TS 5.3.1 to reflect the new fuel type (GE13) that will be inserted 
during Unit 1 Refueling Outage 10; 2. The acceptable range of sodium 
pentaborate concentration for the standby liquid control system shown 
in TS Figure 3.1.5-1 to reflect changes to poison material 
concentration needed to achieve reactor shutdown based on the new GE13 
fuel type.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Proposed Change 1
    The proposed amendment will allow the loading and use of GE13 
fuel assemblies in the Brunswick Unit 1 reactor core. The use of 
GE13 fuel assemblies requires that the safety limit minimum critical 
power ratio value also be revised. The safety limit minimum critical 
power ratio is established to maintain fuel cladding integrity 
during operational transients. The GE13 fuel assembly design has 
been analyzed using methods that have been previously approved by 
the Nuclear Regulatory Commission and documented in General Electric 
Nuclear Energy's reload licensing methodology Topical Report (NEDE-
24011-P-A-11, ``General Electric Standard Application for Reactor 
Fuel (GESTAR II)'' dated November 1995).
    The proposed revision of the safety limit minimum critical power 
ratio does not alter any plant safety-related equipment, safety 
function, or plant operations that could change the probability of 
an accident. The change does not affect the design, materials, or 
construction standards applicable to the fuel bundles in a manner 
that could change the probability of an accident.
    A methodology that has been previously reviewed and accepted by 
the Nuclear Regulatory Commission was used to derive both the 
existing and updated safety limit minimum critical power ratio 
value. The same methodology and criteria have been applied to derive 
the existing safety limit minimum critical power ratio of 1.07 as 
that used to derive the updated safety limit minimum critical power 
ratio value of 1.09. The updated safety limit minimum critical power 
ratio assures that fuel cladding protection equivalent to that 
provided with the existing safety limit minimum critical power ratio 
value is maintained. This ensures that the consequences of 
previously evaluated accidents are not significantly increased.
    Proposed Change 2
    The standby liquid control system provides a means of reactivity 
control that is independent of the normal reactivity control system. 
The standby liquid control system must be capable of assuring that 
the reactor core can be placed in a subcritical condition at any 
time during reactor core life. Technical Specification Figure 3.1.5-
1 specifies the acceptable range of concentrations and volumes for 
sodium pentaborate solution used as a neutron absorber (i.e., for 
reactivity control). The portion of the sodium pentaborate 
concentration range shown in Technical Specification Figure 3.1.5-1 
applicable to the lower range of tank volumes is being revised to 
increase the required concentration of sodium pentaborate solution. 
This change is needed to account for the additional shutdown 
reactivity needed based on the planned use of GE13 fuel assemblies 
as reload fuel for the Unit 1 reactor core. Since the standby liquid 
control system is independent from the normal means of controlling 
reactor core reactivity and not used to control core reactivity 
during normal plant operations, the proposed revision to the sodium 
pentaborate concentration curve for the standby liquid control 
system does not alter any plant safety-related equipment, safety 
function, or plant operations that could change the probability of 
an accident.
    The current volume-concentration range of sodium pentaborate 
used in the standby liquid control system will achieve a sufficient 
concentration of boron in the reactor vessel to ensure reactor 
shutdown. Based on the increased reactivity of the new GE13 reload 
fuel assemblies, the required sodium pentaborate volume-
concentration range is being revised to ensure sufficient neutron 
absorbing solution is available to achieve reactor shutdown; 
therefore, the consequences of an accident previously evaluated are 
not significantly increased.
    2. The proposed amendment would not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    Proposed Change 1
    The GE13 fuel assembly has been designed and complies with the 
acceptance criteria contained in General Electric Nuclear Energy's 
standard application for reactor fuel (GESTAR-II), which provides 
the latest acceptance criteria for new General Electric fuel 
designs. The GE13 fuel assembly complies with GESTAR-II acceptance 
criteria that have been previously reviewed and accepted by the 
Nuclear Regulatory Commission. The similarity of the GE13 fuel 
design to the previously accepted GE11 fuel design, in conjunction 
with the increased critical power capability of the GE13 fuel 
design, ensure that no new mode or condition of plant operation is 
being authorized by the loading and use of the

[[Page 28608]]

GE13 fuel type. The proposed revision of the safety limit minimum 
critical power ratio from 1.07 to 1.09 does not modify any plant 
controls or equipment that will change the plant's responses to any 
accident or transient as given in any current analysis. Therefore, 
the proposed change to allow the loading and use of the GE13 fuel 
type and the revision of the safety limit minimum critical power 
ratio value from 1.07 to 1.09 will not create the possibility for a 
new or different kind of accident from any accident previously 
evaluated.
    Proposed Change 2
    As discussed above, the standby liquid control system provides a 
means of reactivity control that is independent of the normal 
reactivity control system and is capable of assuring that the 
reactor core can be placed in a subcritical condition at any time 
during reactor core life. The proposed revision to the sodium 
pentaborate concentration range does not modify the standby liquid 
control system or its controls, does not modify other plant systems 
and equipment, and does not permit a new or different mode of plant 
operation. As such, the proposed revision to the minimum pentaborate 
concentration value does 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 a margin of safety.
    Proposed Change 1
    As previously discussed, the GE13 fuel assembly design has been 
analyzed using methods that have been previously approved by the 
Nuclear Regulatory Commission and documented in General Electric 
Nuclear Energy's reload licensing methodology Topical Report (NEDE-
24011-P-A-11, ``General Electric Standard Application for Reactor 
Fuel (GESTAR II)'' dated November 1995). The safety limit minimum 
critical power ratio value is selected to maintain the fuel cladding 
integrity safety limit (i.e., that 99.9 percent of all fuel rods in 
the core are expected to avoid boiling transition during operational 
transients). Appropriate operating limit minimum critical power 
ratio values are established, based on the safety limit minimum 
critical power ratio value, to ensure that the fuel cladding 
integrity safety limit is maintained. The operating limit minimum 
critical power ratio values are incorporated in the Core Operating 
limits Report as required by Technical Specification 6.9.3.1. The 
new GE13 safety limit minimum critical power ratio value of 1.09 is 
based on the same fuel cladding integrity safety limit criteria [as] 
that for the GE11 safety limit minimum critical power ratio value of 
1.07 (i.e., that 99.9 percent of all fuel rods in the core are 
expected to avoid boiling transition during operational transients); 
therefore, the proposed change does not result in a significant 
reduction in the margin of safety.
    Proposed Change 2
    As previously stated, the purpose of the standby liquid control 
is to inject a neutron absorbing solution into the reactor in the 
event that a sufficient number of control rods cannot be inserted to 
maintain subcriticality. Sufficient solution is to be injected such 
that the reactor will be brought from maximum rated power conditions 
to subcritical over the entire reactor temperature range from 
maximum operating to cold shutdown conditions. General Electric 
methodology establishes a fuel type dependent standby liquid control 
system shutdown margin to account for calculational uncertainties. 
General Electric calculations show that an in-vessel concentration 
of 660 ppm will provide a standby liquid control system minimum 
shutdown margin in excess of the 3.2%[delta]k value required for the 
GE13 fuel. To achieve an in-vessel concentration of 660 ppm, the 
acceptable range of standby liquid control system tank 
concentrations is being revised for the lower range of tank volumes. 
Thus, the proposed revision of the standby liquid control system 
sodium pentaborate volume-concentration range ensures that there 
will not be a significant reduction in the amount of available 
shutdown margin and, therefore, not a significant reduction in the 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of North Carolina 
at Wilmington, William Madison Randall Library, 601 S. College Road, 
Wilmington, North Carolina 28403-3297.
    Attorney for licensee: 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

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

    Date of amendment request: February 27, 1996
    Description of amendment request: The proposed license amendment 
would modify the Action Statement of Technical Specification (TS) 
3.7.1.1.1. Currently, the TS action statement requires that with the 
self actuation function on one or more main steam line code safety 
valves associated with an operating loop inoperable, the licensee must 
restore the inoperable valve to operable status within 4 hours. 
Otherwise, the plant must be in hot standby within the next 6 hours and 
in hot shutdown within the following 30 hours. The proposed change will 
allow continued power operation at reduced power levels with main steam 
safety valves inoperable. The proposed change is consistent with the 
philosophy of the Westinghouse Standard Technical Specifications.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. [The proposed change does not involve] a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change to the Action Statement of LCO [Limiting 
Condition for Operation] 3.7.1.1.1 will allow indefinite operation 
at less than or equal to 75% power in the event that the self 
actuation function of no more than one safety valve per steam 
generator is inoperable, and allow indefinite operation at less than 
or equal to 50% power in the event that the self actuation function 
of no more than two safety valves per steam generator is inoperable. 
The requirement to reduce power will ensure that there is no 
increase in the consequences of a loss of load accident. The 
proposed change is consistent with the methodology in the 
Westinghouse Standard Technical Specifications. The methodology is 
conservative, since the PORVs [power operated relief valves] cannot 
affect the time of reactor trip on high pressurizer pressure. Thus, 
it is concluded that the change does not increase the consequences 
of any previously evaluated accident.
    The change only specifies a power reduction in the event that 
the self actuation function of steam generator safety valves is 
inoperable. It does not affect the probability of any accident. The 
change by itself does not affect the likelihood of an inoperable 
safety valve.
    2. [The proposed change does not create] the possibility of a 
new or different kind of accident from any previously evaluated.
    The change only specifies a power reduction in the event that 
the self actuation function of steam generator safety valves is 
inoperable. This does not create the potential for a new or 
different kind of accident. The lower power level assures that peak 
steam generator pressure and RCS [reactor coolant system] pressure 
will remain below 110% of design. This provides assurance that no 
equipment failure will occur due to overpressurization. Thus, the 
change does not create the possibility for a new or different kind 
of accident.
    3. [The proposed change does not involve] a significant 
reduction in a margin of safety.
    The allowable power levels have been selected, consistent with 
the Westinghouse Standard Technical Specifications, to assure that 
steam generator and RCS pressure will remain below 110% of design. 
Thus, there is no reduction in a margin of safety for overpressure 
protection.
    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 28609]]

amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: March 7, 1996
    Description of amendment request: The licensee will be replacing a 
locally operated (manual) containment sump suction isolation valve, RH-
V-808A, with a remote manually operated (motor operated) valve, RH-MOV-
808A during the upcoming Cycle 19 refueling outage. As a result, 
changes are being requested to the Haddam Neck Plant Technical 
Specifications to reflect this design change.
    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 technical specification change to Section 3/4.4.6.2 
and its bases are the replacement of the designation RH-V-808A with 
RH-MOV-808A. There are no changes to the requirements of this 
specification and this change is therefore an administrative change. 
The changes to Section 3/4.5.1 will make the requirements for RH-
MOV-808A identical to those of RH-MOV-22. RH-V-808A is being 
converted to a motor operated valve (MOV). This MOV will make the 
ability to establish a suction path from the containment to the 
Residual Heat Removal (RHR) System single failure proof from the 
control room. Both RH-MOV-22 and RH-MOV-808A will be opened to 
establish containment sump recirculation post-loss of coolant 
accident (LOCA). This will provide added assurance that core cooling 
will be maintained in the switch from injection to containment sump 
recirculation following a LOCA. The requirement for RH-MOV-808A to 
be closed and its hand wheel locked can not cause an accident. The 
credit for operation of RH-MOV-808A to ensure that the establishment 
of containment sump recirculation is single failure proof is 
equivalent to the current crediting of RH-V-808A with the only 
difference being that operation of the valve can now be performed 
from the control room. Also, since both RH-MOV-22 and RH-MOV-808A 
will be procedurally opened during establishment of containment sump 
recirculation, the elimination of the requirement to lock open the 
breaker for RH-MOV-22 will not affect the consequences of a LOCA. 
The proposed changes that reflect the conversion of RH-V-808A to a 
MOV and the proposed changes in how the valve is used do not 
increase the consequences of a LOCA.
    2. [The proposed change does not create] the possibility of a 
new or different kind of accident from any previously evaluated.
    The proposed changes will require RH-MOV-808A to be closed with 
the hand wheel locked. This provides assurance that the valve is in 
the required position. Also, RH-MOV-808A will be capable of remote 
manual operation during the monthly surveillance which provides 
assurance that the valve can be repositioned if necessary. The 
proposed opening of RH-MOV-808A at the same time as RH-MOV-22 is 
opened, provides greater assurance that a suction path is available 
to the RHR pumps as well as lowering the total effective piping 
resistance from the containment sump to the pump suction. Therefore, 
the proposed changes do not introduce the possibility of a new or 
different kind of accident.
    3. [The proposed change does not involve] a significant 
reduction in a margin of safety.
    The proposed changes make RH-MOV-808A identical to RH-MOV-22 
with the exception that RH-MOV-808A will not get a closure signal on 
Safety Injection Actuation. Both RH-MOV-22 and RH-MOV-808A are 
containment isolation valves in a closed system. For closed systems, 
the containment isolation requirement is that the valves be either: 
a) automatic, b) locked closed, or c) capable of remote manual 
operation. RH-MOV-808A and RH-MOV-22 are both capable of remote 
manual operation and therefore do not need automatic closure when 
they are opened as part of the technical specification required 
surveillance. Therefore, the proposed changes can 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: March 28, 1996
    Description of amendment request: The proposed license amendment 
will add an additional footnote to Limiting Condition for Operation 
(LCO) 3.4.2.1 and revise an existing footnote for LCO 3.4.2.2. 
Currently, the footnote for LCO 3.4.2.2 requires the pressurizer code 
safety valve as-found lift setting to be within +3 percent and -1 
percent of the setpoint. The proposed change will relax the negative 
as-found lift tolerance to -3 percent. The as-left lift tolerance will 
remain as plus or minus 1 percent. The same footnote will be added to 
LCO 3.4.2.1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. [The proposed change does not involve] a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change will relax the pressurizer safety valve 
negative as-found lift tolerance to -3 percent. The as-left lift 
tolerance will remain as plus or minus 1 percent. This proposed 
technical specification change will allow for the full use of the 
plus or minus 3 percent as-found acceptance criterion for valve 
testing consistent with 1989 ASME Section XI, Subsection IWV. The 
relaxing of the as-found lift tolerance can not cause an accident. 
The relaxing of the tolerance will allow the safety valve setpoint 
to be closer to the Power Operated Relief Valve (PORV) setpoint and 
could result in a slightly lower pressure for overheating events. 
The analysis that takes credit for the increase in pressure to the 
PORV setpoint is the Loss of Load analysis. The minimum departure 
from nucleate boiling ratio (DNBR) was reanalyzed without taking any 
credit for the transient increase in pressure. The minimum DNBR 
still remains above the acceptance criterion as well as above the 
limiting minimum DNBR predicted for all Updated Final Safety 
Analysis Report Chapter 15 accidents. Also, the relaxed tolerance in 
conjunction with a lower safety valve blowdown, yet still 
conservative, results in a slightly higher average pressure for a 
valve lift/reset cycle. This means that pressurizer overfill will 
not be predicted for the limiting transient, loss of feedwater. 
Thus, the proposed relaxation of as-found lift tolerance does not 
increase the probability or consequences of the design basis 
accidents previously evaluated.
    2. [The proposed change does not create] the possibility of a 
new or different kind of accident from any previously analyzed.
    The proposed relaxation of the lift tolerance still requires the 
safety valve lift setpoint to be above both the PORV setpoint and 
the pressurizer high pressure reactor trip setpoint. In addition, 
the as-left setpoint is not being changed. The relaxed tolerance in 
combination with a conservative safety valve blowdown still will 
preclude the prediction of water relief from the pressurizer. This 
means that the proposed change does not introduce the possibility of 
a new or different kind of accident.

[[Page 28610]]

    3. [The proposed change does not involve] a significant 
reduction in a margin of safety.
    The proposed relaxation of the as-found lift tolerance for valve 
testing is consistent with 1989 ASME Section XI, Subsection IWV. The 
as-left lift tolerance will remain plus or minus 1 percent. In 
addition, the design basis analyses still meet their acceptance 
criteria with the -3 percent lift tolerance. Therefore, the proposed 
change can not cause 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: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: April 16, 1996
    Description of amendment request: The licensee is proposing to 
revise the Technical Specifications to permit the Haddam Neck Plant to 
remain in Mode 1, 2, 3, or 4 with the average water temperature of the 
ultimate heat sink (UHS) greater than 90 deg. additional action has 
been added which would require the plant to be placed in at least Hot 
Standby within 6 hours and in Cold Shutdown within the following 30 
hours upon identifying that the average water temperature of the UHS is 
greater than 95 deg.F. In addition, the licensee is proposing to 
include a new surveillance requirement for monitoring the average 
circulating water inlet temperature to be within its limits when the 
average water temperature of the UHS exceeds 89 deg.F.
    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 addition to the Action Statement of LCO 3.7.12 of 
an 8 hour period to monitor the average water temperature of the UHS 
does not involve an increase in the probability of an accident 
previously evaluated. The probability of an accident previously 
evaluated is not increased by a short-term increase in the average 
water temperature of the UHS. An evaluation of the service water 
loads associated with the loss-of-offsite power and a coincident 
worst case single failure of a diesel generator to start (resulting 
in the loss of two of the four service water pumps) determined that 
there is adequate margin to accomplish plant cooldown at a service 
water inlet temperature of 95 deg.F. The recirculation phase of a 
LOCA [loss-of-coolant accident] was evaluated to verify that 
adequate flow would be available to the RHR [residual heat removal] 
heat exchangers. The most limiting assumptions for the recirculation 
phase are offsite power is available and one RHR heat exchanger 
service water isolation valve fails to open. The injection phase of 
a LOCA was evaluated to verify that adequate flow would be available 
to the CAR [containment air recirculation] fan cooling coils. The 
most limiting assumption for the injection phase is a loss-of-
offsite power. The results of these evaluations determined that 
there is adequate service water flow to accomplish plant cooldown 
with average water temperature of the UHS up to 95 deg.F. CYAPCO 
[Connecticut Yankee Atomic Power Company] also proposes to include 
an additional surveillance requirement to monitor the average water 
temperature of the UHS at least once per hour if the average water 
temperature of the UHS exceeds 89 deg.F. This additional 
surveillance requirement ensures increased operator awareness as the 
average water temperature of the UHS approaches the 90 deg.F LCO 
limit. Based on the above, there is no significant increase in the 
consequences of any accident previously evaluated.
    2. [The proposed change does not] create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The proposed technical specification changes do not create the 
possibility of a new or different kind of accident from those 
previously evaluated. The addition of an 8 hour time period to 
monitor the average water temperature of the UHS increases from 6 to 
14 hours the amount of time that is allowed before the plant must 
proceed to Hot Standby should the average water temperature of the 
UHS increase above 90 deg.F. This extension of the time allowed for 
the plant to be in Hot Standby does not change the plant 
configuration. CYAPCO also proposes to include an additional 
surveillance requirement to monitor the average water temperature of 
the UHS at least once per hour if the average water temperature of 
the UHS exceeds 89 deg.F. This additional surveillance requirement 
ensures increased operator awareness as the average water 
temperature of the UHS approaches the 90 deg.F LCO limit.
    As such, the changes do not create the possibility of a new or 
different kind of accident from those previously evaluated.
    3. [The proposed change does not] involve a significant 
reduction in a margin of safety.
    The proposed technical specification changes do not involve a 
significant reduction in any margin of safety. The addition of an 8 
hour time period to monitor the average water temperature of the UHS 
increases from 6 to 14 hours the time required before the plant must 
proceed to Hot Standby should the average water temperature of the 
UHS temperature [exceed] 90 deg.F. An evaluation has been performed 
to demonstrate that the risk significance associated with the 
increased action time is very low. In addition, safe shutdown 
capability has been demonstrated for service water inlet 
temperatures as high as 95 deg.F. The addition of a surveillance 
requirement to monitor the average water temperature of the UHS at 
least once per hour if the average water temperature of the UHS 
exceeds 89 deg.F is an additional requirement, limitation, or 
restriction not currently within the technical specifications. 
Therefore, these changes do not involve a significant reduction in a 
margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270
    NRC Project Director: Phillip F. McKee

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

    Date of amendment request: April 22, 1996
    Description of amendment request: The proposed amendment will allow 
the use of the performance-based containment leakage testing 
requirements described in 10 CFR Part 50, Appendix J, Option B.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Involve a significant increase in the probability or 
consequences of an accident previously analyzed.
    The changes involved in this license amendment request revise 
the testing criteria for the containment penetrations. The revised 
criteria will be based on the guidance in Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.'' This guidance 
allows for the use of relaxed testing frequencies for containment 
penetrations that have performed satisfactorily on a historical 
basis. The Containment Leakage Rate Testing Program considers the 
type of service, the design of the penetration, and the safety 
impact of the penetration in determining the

[[Page 28611]]

testing interval of each penetration. The NRC Staff has reviewed the 
potential impact of performance-based testing frequencies for 
containment penetrations during the development of the Option B 
regulation. The NRC Staff review is documented in NUREG-1493, 
``Performance-Based Containment Leakage-Test Program.'' The review 
concluded that reducing the frequency of Type A tests (Integrated 
Leakage Rate Tests) from three per 10 years to one per 10 years 
leads to an imperceptible increase in risk. For Type B and C testing 
(Local Leakage Rate Tests), the change in testing frequency should 
not have significant impact since this leakage contributes less than 
0.1 percent of the overall risk based on the existing regulations. 
The use of Option B will allow the extension of testing intervals 
with a minimal impact on the radiological release rates since most 
penetration leakage is continually well below the specified limits. 
In the accident risk evaluation, the NRC Staff noted that the 
accident risk is relatively insensitive to the containment leakage 
rate because the accident risk is dominated by accident sequences 
that result in failure of or bypass of the containment. The use of a 
performance-based testing program will continue to provide assurance 
that the accident analysis assumptions remain bounding. Therefore, 
the proposed change does not involve a significant increase in the 
probability or consequences of an accident previously analyzed.
    Removal of the surveillance accuracy requirement in Section 
4.6.1.2.c will not affect the probability of an accident previously 
analyzed since a similar requirement is contained in ANSI/ANS-56.8-
1994, ``Containment System Leakage Testing Requirements.'' ANSI/ANS-
56.8-1994 will be used to develop the technical methods and 
techniques for the Containment Leakage Rate Test Program as stated 
in Regulatory Guide 1.163. The technical methods and techniques in 
ANSI/ANS-56.8-1994 have been determined to be acceptable to the NRC 
Staff.
    Changes to the Administrative Section describe the containment 
testing program only and cannot increase the probability or 
consequences of an accident previously analyzed.
    2. Create the possibility of a new or different kind of accident 
from any previously analyzed.
    The proposed license amendment does not change the operation or 
equipment of the plant. The change in the test frequency is 
dependent on the establishment of a Containment Leakage Test 
Program. This test program will ensure the performance history of 
each penetration is satisfactory prior to the changing of any test 
frequency. Since the performance history of the penetration will be 
known, there is no possibility of the implementation of the program 
creating a new or different kind of accident than previously 
analyzed. Since there is no change to the equipment or the operation 
of the plant, there is no possibility of creating a new or different 
kind of accident than previously analyzed. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously analyzed.
    Removal of the surveillance accuracy requirement in Section 
4.6.1.2.c will not create the possibility of a new or different kind 
of accident from those previously analyzed since a similar 
requirement is contained in ANSI/ANS-56.8-1994, ``Containment System 
Leakage Testing Requirements.'' ANSI/ANS-56.8-1994 will be used to 
develop the technical methods and techniques for the Containment 
Leakage Rate Test Program as stated in Regulatory Guide 1.163. The 
technical methods and techniques in ANSI/ANS-56.8-1994 have been 
determined to be acceptable to the NRC staff.
    Changes to the Administrative Section describe the containment 
testing program only and cannot create a different accident from any 
previously analyzed.
    3. Involve a significant reduction in a margin of safety.
    During the development of 10 CFR Part 50, Appendix J, Option B, 
the NRC staff determined the reduction in safety associated with the 
implementation of the performance-based testing program. The results 
of this review are documented in NUREG-1493. The review concluded 
that reducing the frequency of Type A tests (Integrated Leakage Rate 
Tests) from three per 10 years to one per 10 years leads to an 
imperceptible increase in risk. For Type B and C testing (Local 
Leakage Rate Tests), the increase in testing frequency should not 
have significant impact since this leakage contributes less than 0.1 
percent of the overall risk based on the existing regulations. The 
use of Option B will allow the extension of testing intervals with a 
minimal impact on the radiological release rates since most 
penetration leakage is continually well below the specified limits. 
In the accident risk evaluation, the NRC Staff noted that the 
accident risk is relatively insensitive to the containment leakage 
rate because the accident risk is dominated by accident sequences 
that result in failure of or bypass of the containment. The use of a 
performance based testing program will continue to provide assurance 
that the accident analysis assumptions remain bounding. Therefore, 
this change does not involve a significant reduction in the margin 
of safety.
    Removal of the surveillance accuracy requirement in Section 
4.6.1.2.c will not involve a significant reduction in the margin of 
safety since a similar requirement is contained in ANSI/ANS-56.8-
1994, ``Containment System Leakage Testing Requirements.'' ANSI/ANS-
56.8-1994 will be used to develop the technical methods and 
techniques for the Containment Leakage Rate Test Program as stated 
in Regulatory Guide 1.163. The technical methods and techniques in 
ANSI/ANS-56.8-1994 have been determined to be acceptable to the NRC 
Staff.
    Changes to the Administrative Section describe the containment 
testing program only and do not reduce the margin of safety.
    Moreover, the Commission has provided guidance concerning the 
application of standards in 10 CFR 50.92 by providing certain 
examples (51 FR 7751, March 6, 1986) of amendments that are 
considered not likely to involve an SHC [significant hazards 
consideration]. Although the proposed change is not enveloped by a 
specific example, it has been shown that the proposed change is not 
an SHC.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Russell Library, 123 Broad 
Street, Middletown, CT 06457.
    Attorney for licensee: Lillian M. Cuoco, Esq., Senior Nuclear 
Counsel, Northeast Utilities Service Company, P.O. Box 270, Hartford, 
CT 06141-0270.
    NRC Project Director: Phillip F. McKee

Consumers Power Company, Docket No. 50-255, Palisades Plant, Van 
Buren County, Michigan

    Date of amendment request: February 6, 1996
    Description of amendment request: The proposed amendment would 
delete the requirement to perform additional operability testing of 
safety system train components when a required component in the 
redundant train becomes 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. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed changes remove the requirement for testing which is 
in addition to the normal surveillance interval. The affected 
equipment is subject to periodic surveillance testing required by 
the Technical Specifications. Removing the requirement for 
additional testing cannot alter any plant operating conditions, 
operating practices, equipment settings, or equipment capabilities. 
Therefore, changing an AOT [allowable outage time] or a surveillance 
interval cannot increase the probability or consequences of an 
accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    The proposed changes remove the requirement for testing which is 
in addition to the normal surveillance interval. The affected 
equipment is subject to periodic surveillance testing required by 
the Technical Specifications. Removing the requirement for 
additional testing cannot alter any plant operating conditions, 
operating practices, equipment settings, or equipment capabilities. 
Therefore, changing an AOT or a surveillance interval cannot create 
the possibility of a new or different

[[Page 28612]]

kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    The proposed changes remove the requirement for testing which is 
in addition to the normal surveillance interval, in effect extending 
the surveillance interval. An excessive surveillance interval 
extension could reduce the margin of safety by reducing assurance 
that required equipment will function as designed; an overly 
restrictive surveillance interval could also reduce the margin of 
safety by imposing unnecessary testing wear, equipment 
manipulations, and system transients on the plant.
    The existing requirements to perform cross-train testing were 
based on the operating experience available when they were added to 
the TS. Typically this was done during the initial plant licensing 
in 1971. The recently published Standard Technical Specifications 
(NUREG 1432) do not include cross-train testing requirements for the 
Engineered Safety Features components. It has been judged by the NRC 
and by the industry, that cross-train testing is unnecessary, and 
that testing at normal surveillance intervals is adequate to assure 
equipment operability. This recent judgment is based on a much 
larger accumulation of operating experience than was available at 
the time Palisades was licensed. There are no special features of 
the Palisades plant which would invalidate these more recent 
judgments of optimal testing requirements. Therefore, operation of 
the facility in accordance with the proposed changes will not 
involve a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Van Wylen Library, Hope 
College, Holland, Michigan 49423
    Attorney for licensee: Judd L. Bacon, Esquire, Consumers Power 
Company, 212 West Michigan Avenue, Jackson, Michigan 49201
    NRC Project Director: Mark Reinhart

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

    Date of amendment request: December 14, 1995, as supplemented by 
letter dated May 16, 1996.
    Description of amendment request: The proposed amendments would 
change the Technical Specifications (TS) to improve the TS Action 
Statements and Surveillance Requirements for diesel generators in 
accordance with the recommendations and guidance in Generic Letter 93-
05, Generic Letter 94-01, NUREG-1366, and NUREG-1431. The proposed 
amendments would also incorporate technical and administrative 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:
    Criterion 1
    Operation of the facilities in accordance with the requested 
amendments will not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
Improvements to the LCOs [limiting condition for operation] and 
surveillance requirements for the emergency diesel generators do not 
affect their capability to provide emergency power to plant vital 
instruments and safety related equipment. In fact, these 
improvements make the diesel generators more reliable since they 
significantly reduce the amount of wear and stress due to excessive 
and unnecessary testing. The proposed monthly testing of the diesel 
generator continues to ensure that the system is ready for service 
when needed. The fast starts and fast loadings continue to ensure 
that the timing and loading requirements for engineered safety 
features actuation are met. The proposed changes do not affect any 
of the design basis accident analyses previously evaluated. 
Therefore, these proposed changes do not involve any increase in the 
probability or consequences of any accident previously evaluated. 
The proposed changes are fully consistent with the recommendations 
and guidance contained in GL [Generic Letter] 93-05, GL 94-01, 
NUREG-1366, NUREG-1431, and are compatible with plant operating 
experience.
    Criterion 2
    Operation of the facilities in accordance with the requested 
amendments will not create the possibility of a new or different 
kind of accident from any accident previously evaluated. The 
proposed changes in fact improve the reliability of the diesel 
generators by eliminating unnecessary wear and stress. Improved 
reliability decreases the failure probability which also decreases 
the probability of an accident not previously evaluated. None of the 
requested amendments increase the common mode failure probability 
thus would not increase the chance of both EDG's [emergency diesel 
generators] for a particular nuclear unit being out of service 
simultaneously. The proposed changes are fully consistent with the 
recommendations and guidance contained in GL 93-05, GL 94-01, NUREG-
1366, NUREG-1431, and are compatible with plant operating 
experience.
    Criterion 3
    Operation of the facilities in accordance with the requested 
amendments will not involve a significant reduction in a margin of 
safety. The proposed monthly testing of the diesel generators 
continues to ensure that the system is ready for service when 
needed. The fast starts and fast loadings continue to ensure that 
the timing and loading requirements for engineered safety features 
actuation are met. The proposed changes improve the reliability of 
the diesel generators. Implementation of the Maintenance Rule also 
ensures continued reliability of the diesel generators. No margin of 
safety is decreased as a result of these TS changes.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Atkins Library, University of 
North Carolina, Charlotte (UNCC Station), North Carolina 28223
    Attorney for licensee: Mr. Albert Carr, Duke Power Company, 422 
South Church Street, Charlotte, North Carolina 28242
    NRC Project Director: Herbert N. Berkow

Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, 
Unit No. 1, Pope County, Arkansas

    Date of amendment request: April 29, 1996
    Description of amendment request: The proposed amendment relocates 
several cycle specific operating parameters from the technical 
specifications to the Core Operating Limits Report per Generic Letter 
88-16. The parameters being relocated by this change include the 
variable low reactor coolant system pressure trip (VLPT) and the 
variable low reactor coolant system pressure-temperature protective 
limits.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Criterion 1. Does Not Involve a Significant Increase in the 
Probability or Consequences of an Accident Previously Evaluated.
    The removal of the cycle-dependent variable low RCS pressure-
temperature protective limits and the VLPT setpoint from technical 
speciications and placing them into the COLR has no impact on plant 
safety and is considered to be administrative in nature. The 
proposed change does not affect the safety analyses, physical 
design, or operation of the plant. Technical specifications will 
continue to require operation within the core protective and 
operational limits for each reload cycle as calculated by the 
approved reload design methodologies. The appropriate actions 
required if limits are violated will remain in the technical 
specifications. The reload report presents the results of cycle-
specific evaluations of accident analyses and transients addressed 
in the ANO-1 Safety Analysis Report. The cycle-specific 10CFR50.59 
evaluation of the reload

[[Page 28613]]

report demonstrates that changes in fuel cycle design and the 
corresponding COLR do not involve a significant increase in the 
probability or 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.
    The proposed change to relocate the variable low RCS pressure-
temperature protective limits and the VLPT setpoint from the 
technical specifications to the COLR is administrative in nature. No 
change to the design configuration or method of operation of the 
plant is made by this proposed change, and therefore, no new 
transient initiator has been created. Technical specifications will 
continue to require operation within the required core protective 
and operating limits and appropriate actions will be taken if the 
limits are exceeded. Because plant operation will continue to be 
limited by the cycle-specific COLR limits that are established using 
NRC-approved methodologies, these relocations will have no impact on 
plant safety.
    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.
    Existing technical specification operability and surveillance 
requirements are not reduced by the proposed change to relocate the 
variable low RCS pressure-temperature protective limits and the VLPT 
setpoint to the COLR. The proposed changes are administrative in 
nature and do not relate to or modify the safety margins defined in 
and maintained by the technical specifications. The cycle-specific 
COLR limits for future reload fuel cycles will continue to be 
developed based on NRC approved methodologies. Each future reload 
undergoes a 10CFR50.59 evaluation to assure that operation of the 
plant within the cycle-specific limits will not involve a 
significant reduction in a margin of safety.
    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: May 6, 1996
    Description of amendment request: The amendment would reflect that 
the name of Mississippi Power & Light Company (MP&L) has been changed 
to Entergy Mississippi, Inc. The amendment revises Operating License 
NPF-29 and Antitrust Conditions for the Grand Gulf Nuclear Station, 
Unit 1 (GGNS) to (1) add the phrase ``(now renamed Entergy Mississippi, 
Inc.)'' after the name of Mississippi Power & Light Company (MP&L), (2) 
replace the name of Mississippi Power & Light Company (MP&L) by the 
name Entergy Mississippi, Inc., and (3) replace a footnote by the 
statement: ``Amendment ---- resulted in a name change for Mississippi 
Power & Light Company (MP&L) to Entergy Mississippi, Inc.''.The 
proposed amendment involves only a change in company name. It does not 
involve any changes to the Technical Specifications for GGNS, or to any 
requirements or limiting conditions for operation on any equipment or 
any systems in the plant.
    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:
    Entergy Operations, Inc. proposes to change the current Grand 
Gulf Nuclear Station Facility Operating License and Antitrust 
Conditions. The specific proposed change is to reflect that the name 
of one of the companies owning Grand Gulf Nuclear Station has 
legally changed from Mississippi Power & Light Company to Entergy 
Mississippi, Inc.
    The Commission has provided standards for determining whether a 
no significant hazards consideration exists as stated in 10 CFR 
50.92(c). 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 and 
determined that no significant hazards consideration results from 
this change. 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 proposed change documents changing the legal name of the 
company. The proposed change will not affect any other obligations. 
The company will still own all of the same assets, serve the same 
customers, and all existing obligations and commitments will 
continue unaffected.
    [The proposed change does not affect any of the existing 
requirements or commitments on equipment or systems that are 
designed for the safe operation of the plant. It does not affect the 
design or operation of the plant.]
    Therefore, the proposed change does 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 administrative changes to the Operating License [and 
Antitrust Condition] requirements [to change the name of Mississippi 
Power & Light] do not involve any change in the design or operation 
of the plant. The company will still own all of the same assets, 
serve the same customers, and all existing obligations and 
commitments will continue unaffected.
    [The proposed changes do not affect equipment or systems that 
could caused an accident at the plant.]
    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 proposed change [in name] is administrative in nature, as 
described above; therefore, this change does not reduce the level of 
safety imposed by any current requirements. [The proposed changes do 
not affect any equipment or systems at the plant.] The company will 
still own all of the same assets, serve the same customers, and all 
existing obligations and commitments will continue unaffected.
    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. herefore, 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

[[Page 28614]]

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

    Date of amendment request: May 8, 1996
    Description of amendment request: The amendment request would 
replace the current frequency requirements in Surveillance Requirement 
(SR) 3.6.1.3.5, on the leakage rate testing for each containment purge 
valve with resilient seals, in the Technical Specifications for Grand 
Gulf Nuclear Station, Unit 1 (GGNS). The proposed changes would place 
these purge valves on a performance-based leakage testing frequency, 
instead of the current once every 184 days and once within 92 days 
after opening the valve.The proposed changes do not change the limiting 
conditions for operation, the required actions for inoperability, or 
the other surveillance requirements on these primary containment 
isolation valves.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    In accordance with 10 CFR 50.92, Entergy Operations, Inc. has 
evaluated the proposed change to the Operating License of GGNS and 
has determined that the operation of the facility in accordance with 
the proposed amendment would not involve any significant hazards 
considerations. In accordance with 10 CFR 50.91(a), Entergy 
Operations, Inc. is providing the following analysis of the proposed 
amendment against the three [following] standards of 10 CFR 
50.92(c):
    1) The proposed change does not significantly increase the 
probability or consequences of an accident previously evaluated.
    This change deletes the augmented testing requirement for these 
containment isolation valves and allows the surveillance intervals 
to be set in accordance with the Appendix J testing program. 
[Appendix J to 10 CFR Part 50 defines primary containment leakage 
testing requirements for water-cooled power reactors as GGNS and 
these requirements include frequency of testing for the primary 
containment isolation valves.] This change does not affect the 
system function or design. The purge valves are not an initiator of 
any previously analyzed accident. Leakage rates do not affect the 
probability of the occurrence of any accident. Operating history has 
demonstrated that these valves do not degrade and cause leakage as 
previously anticipated. Because these valves have been demonstrated 
to be reliable, these valves can be expected to perform the 
containment isolation function as assumed in the accident analyses.
    Therefore, there is no significant increase in the consequences 
of any previously evaluated accident.
    2) The proposed change would not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Extending the test intervals has no influence on, nor does it 
contribute in any way to, the possibility of a new or different kind 
of accident or malfunction from those previously analyzed. No change 
has been made to the design, function or method of performing 
leakage testing [or to the design and function of these valves]. 
Leakage acceptance criteria have 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.
    [Therefore, the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.]
    3) The proposed change does not involve a significant reduction 
in a margin of safety
    The only margin of safety that has the potential of being 
impacted by the proposed changes involves the offsite dose 
consequences of postulated accidents which are directly related to 
the containment leakage rate. The proposed change does not alter the 
method of performing the tests nor does it change the leakage 
acceptance criteria. Sufficient data has been collected to 
demonstrate that the resilient seals do not degrade at an 
accelerated rate.
    [Also, the proposed change would test these valves in accordance 
with the Appendix J testing program at the plant. Appendix J to 10 
CFR Part 50 defines primary containment leakage testing requirements 
for water-cooled power reactors as GGNS and these requirements 
include frequency of testing for the primary containment isolation 
valves.]
    Because of this demonstrated reliability, this change will 
provide sufficient surveillance to determine an increase in the 
unfiltered leakage prior to the leakage exceeding that assumed in 
the accident analysis.
    Therefore, the proposed change does not result in a significant 
reduction in a margin of safety.
    Based on the above evaluation, Entergy Operation, Inc. has 
concluded that operation in accordance with the proposed amendment 
involves no significant hazards considerations.
    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., et al., Docket No. 50-416, Grand Gulf 
Nuclear Station, Unit 1, Claiborne County, Mississippi

    Date of amendment request: May 9, 1996
    Description of amendment request: The amendment request would (1) 
increase the safety limit minimum critical power ratio (MCPR) for two 
loop operation and single loop operation to 1.10 and 1.11, 
respectively, and (2) add a General Electric topical report to the list 
of documents describing the analytical methods used to determine the 
core operating limits. The proposed changes are to Section 2.1.1, 
Reactor Core Safety Limits, and Section 5.6.5, Core Operating Limits 
Report (COLR), respectively, of the Technical Specifications (TSs).
    The licensee also proposed changes to the Bases of the TSs 
associated with the above proposed 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:
    Entergy Operations, Inc. proposes to change the current Grand 
Gulf Nuclear Station [GGNS] Technical Specifications. The specific 
change is to modify the Minimum Critical Power Ratio (MCPR) safety 
limits reported in Technical Specification 2.1.1.2, the list of 
references in Technical Specification 5.6.5, and associated Bases 
changes. The proposed change is necessary in order to switch reload 
fuel vendors. [General Electric GE11 fuel is being added to the core 
in place of Siemens Power Corporation (SPC) fuel.]
    The Commission has provided standards for determining whether no 
significant hazards considerations exists as stated in 10 CFR 50.92 
(c). A proposed amendment to an operating license involves no 
significant hazards 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; (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 and 
determined that no significant hazards considerations result from 
this change. 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.

[[Page 28615]]

    The Minimum Critical Power Ratio (MCPR) safety limit is defined 
in the Bases to Technical Specification 2.1.1 as that limit which 
``ensures that during normal operation and during Anticipated 
Operational Occurrences (AOOs), at least 99.9% of the fuel rods in 
the core do not experience transition boiling.'' The MCPR safety 
limit is re-evaluated for each reload and, for GGNS [Operating] 
Cycle 9, the analyses have concluded that a two-loop MCPR safety 
limit of 1.10 based on the application of the generic GE MCPR 
methodology is necessary to ensure that this acceptance criterion is 
satisfied. For single-loop operation, a MCPR safety limit of 1.11 
based on the generic GE MCPR methodology was determined to be 
necessary. Core MCPR operating limits are developed to support the 
Technical Specification 3.2 requirements and ensure these safety 
limits are maintained in the event of the worst-case transient. 
Since the MCPR safety limit will be maintained at all times, 
operation under the proposed changes will ensure at least 99.9% of 
the fuel rods in the core do not experience transition boiling. 
Therefore, The Minimum Critical Power Ratio (MCPR) safety limit 
change does not affect the probability or consequences of an 
accident.
    The implementation of GE's GESTAR-II approved methodology has no 
effect on the probability or consequences of any accidents 
previously evaluated. One exception to GESTAR is that the mis-
oriented and mis-located bundle events will continue to be analyzed 
as accidents subject to the acceptance criteria in the current 
licensing basis. The design of the GE11 fuel bundles is such that 
the bundles are not likely to be mis-oriented or mis-located and the 
normal administrative controls will be in effect for assuring proper 
orientation and location. Therefore, the probability of a fuel 
loading error is not increased. This analysis ensures that 
postulated dose releases will not exceed a small fraction (10 
percent) of 10CFR100 limits.
    Therefore, the consequences of accidents previously evaluated 
are unchanged.
    II. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The GE11 fuel to be used in [Operating] Cycle 9 is of a design 
compatible with fuel present in the core and used in the previous 
cycle. Therefore, the GE11 fuel will not create the possibility of a 
new or different kind of accident. The proposed changes do not 
involve any new modes of operation, any changes to setpoints, or any 
plant modifications. They introduce revised MCPR safety limits that 
have been proved to be acceptable for Cycle 9 operation. Compliance 
with the applicable criterion for incipient boiling transition 
continues to be ensured. The proposed MCPR safety limits do not 
result in the creation of any new precursors to an accident.
    Therefore, the proposed changes do not create the possibility of 
a new or different type of accident from any accident previously 
evaluated.
    III. The proposed change does not involve a significant 
reduction in a margin of safety.
    The MCPR safety limits have been evaluated to ensure that during 
normal operation and during AOOs [abnormal operating occurrences], 
at least 99.9% of the fuel rods in the core do not experience 
transition boiling. Therefore, the implementation of the proposed 
changes in the MCPR safety limit ensure there is no reduction in the 
margin of safety.
    As with the current SPC methodology, GGNS will implement only 
the NRC-approved revisions to GE's GESTAR methodology. This GE 
methodology is similar to those SPC reports currently listed in TS 
5.6.5 and it will be applied in a similar, conservative fashion. One 
exception to GESTAR is that the mis-oriented and mis-located bundle 
events will continue to be analyzed as accidents subject to the 
acceptance criteria in the current licensing basis. This analysis 
ensures that postulated dose releases will not exceed a small 
fraction (10 percent) of 10CFR100 limits. On this basis, the 
implementation of this GE methodology does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: 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: February 6, 1996
    Description of amendment request: The proposed change will amend 
the Allowable Values of parameters in Table 3.3-4 of Waterford Steam 
Electric Station, Unit 3, (Waterford 3) Technical Specifications (TSs) 
to make it consistent with the identical parameters in Table 2.2-1 of 
TSs for Waterford 3. The proposed change will add Mode 4 to the 
surveillance requirements of Table 4.3-2, Item 5.c (Safety Injection 
System Automatic Actuation Logic) that was inadvertently removed. 
Finally, the proposed change removes a reference to TS 3.3.3.2 in 
Surveillance Requirements TS 4.10.2.2 and 4.10.4.2 since Incore 
Detectors has been removed from the TSs.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed changes described herein are administrative changes 
necessary to correct administrative errors. The proposed changes 
will have no affect on any design basis accidents nor will these 
changes affect any material condition of the plant. Therefore, the 
proposed changes will not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    The proposed changes are purely administrative. There are no new 
system or design changes associated with this proposal. Therefore, 
the proposed change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The proposed change will have no impact on any protective 
boundary, safety limit, or margin to safety. The proposed change 
corrects inconsistencies in the TS and is purely administrative in 
nature. Therefore, the proposed change will not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: 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

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

    Date of amendment request: May 7, 1996 (TSCR 247)
    Description of amendment request: The proposed change to the 
technical specifications would adopt the provisions of the Standard 
Technical Specifications (STS), NUREG-1433, Rev. 1, which clarify 
surveillance requirement applicability and allow a maximum period of 24 
hours to complete a surveillance requirement upon discovery that the 
surveillance has been missed.
    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 
amendment would not

[[Page 28616]]

involve a significant increase in the probability of occurrence or 
consequence of an accident previously evaluated. The proposed 
changes only affect administrative requirements regarding the 
applicability and performance of surveillances. This change 
clarifies surveillance requirement applicability and allows a 
maximum 24 hour delay period for the performance of a surveillance 
when it is discovered that the surveillance has not been performed 
within the required frequency, consistent with the STS. There is 
minimal safety significance associated with a delay of 24 hours in 
completing the required surveillance, particularly due to the fact 
that the most probable result of any particular surveillance 
performed is the successful verification of conformance with the 
requirements.
    2. Operation of the facility in accordance with the proposed 
amendment would not create the possibility of a new or different 
kind of accident from any previously evaluated. The proposed changes 
only affect administrative requirements regarding the applicability 
of surveillance requirements and the performance of surveillances to 
allow a maximum 24 hour delay period when it is discovered that a 
surveillance has been missed. No changes to plant equipment or 
operation are affected.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in the margin of 
safety since the change contained in the proposed amendment does not 
change any existing safety margins.
    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: Ocean County Library, 
Reference Department, 101 Washington Street, Toms River, NJ 08753
    Attorney for licensee: Ernest L. Blake, Jr., Esquire. Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: John F. Stolz

GPU Nuclear Corporation, Docket No. 50-320, Three Mile Island 
Nuclear Station, Unit No. 2 (TMI-2), Dauphin County, Pennsylvania

    Date of amendment request: February 16, 1995
    Description of amendment request: The proposed amendment would 
revise TMI-2 Operating License No. DPR-73 by modifying sections 4.02, 
4.04, and 4.1.1.3 of the unit technical specifications. The revisions 
to sections 4.02 and 4.04 would add flexibility to the scheduling of 
surveillance activities and would allow for a 24 hour period to perform 
missed surveillances before declaration of a limiting condition for 
operation, respectively. These changes would make the TMI-2 technical 
specifications consistent with the Standard Technical Specifications 
for B&W Plants (NUREG-1430). The revision to section 4.1.1.3 would 
allow extension of the time interval for surveillance of the 
containment airlock doors from quarterly to annually. The proposed 
changes to the TMI-2 technical specifications section 4.1.1.3 would 
allow a decrease in worker exposure to radiation while maintaining an 
adequate level of environmental protection at the facility.
    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:
    10 CFR 50.92 provides the criteria which the Commission uses to 
perform a no significant hazards consideration. 10 CFR 50.92 states 
that an amendment to a facility license involves no significant 
hazards 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.
    The proposed changes to the technical specifications sections 
4.02 and 4.04 are administrative and do not involve any physical 
changes to the facility. No changes are made to operating limits or 
parameters, nor to any surveillance activities. The changes to 
section 4.1.1.3 extends the interval between surveillance of the 
containment airlocks; it does not change the operability 
requirements, test methodology or acceptance criteria. Based on 
this, GPU Nuclear has concluded that the proposed changes to 
sections 4.02 and 4.04 do not:
    1. Involve a significant increase in the probability of 
occurrence or the consequences of an accident previously evaluated. 
The changes do not modify any operating parameters or the release of 
radioactive materials. The clarification of maximum time extensions 
for surveillance is consistent with the NRC's Standard Technical 
Specifications for Babcock and Wilcox Plants (NUREG-1430).
    2. Create the possibility of a new or different kind of accident 
since these change are administrative and no plant configuration or 
operational changes are involved.
    3. Involve a change in the margin of safety. These changes are 
administrative in nature, compatible with standard technical 
specifications, and do not affect any safety settings or operational 
limits.
    GPU Nuclear has also concluded that the proposed changes to 
section 4.1.1.3 do not:
    1. Involve a significant increase in the probability of 
occurrence of or consequences of an accident previously evaluated. 
The change to this section does not change operating parameters, 
equipment operability requirements, surveillance test methodology, 
or acceptance criteria.
    2. Create the possibility of a new or different kind of accident 
since the change does not affect plant equipment, plant 
configuration, or plant operating parameters.
    3. Involve a change in the margin of safety since the change 
does not affect any operational limits.
    Based on the above analysis the licensee concluded that the 
proposed changes involve no significant safety hazards 
considerations as defined by 10 CFR 50.92.
    The NRC staff has reviewed the analysis of the licensee and, based 
on this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room Location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105
    Attorney for licensee: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts & Trowbridge, 2300 N Street, NW, Washington, DC 20037
    NRC Project Director: Seymour H. Weiss

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498 and 50-499, South Texas Project, Units 1 and 2, 
Matagorda County, Texas

    Date of amendment request: May 1, 1996
    Description of amendment request: The proposed amendments would 
change the Technical Specifications to implement 10 CFR Part 50, 
Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leakage-Test Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    [South Texas Project] STP has evaluated the proposed Technical 
Specification Amendment and determined that it does not represent a 
significant hazards consideration. Based on the criteria for 
defining a significant hazards consideration established in 10 CFR 
50.92, operation of STP in accordance with the proposed amendment 
will not:
    1) Involve a significant increase in the probability or 
consequences of an accident previously evaluated because of the 
following:
    10 CFR [Part] 50, Appendix J has been amended to include 
provisions regarding

[[Page 28617]]

performance based leakage testing requirements (Option B). Option B 
allows plants with satisfactory Integrated Leak Rate Testing (ILRT) 
performance history to extend the Type A testing interval from three 
tests in ten years to one test in ten years. For Type B and Type C 
tests, Option B allows extended testing interval[s] based on the 
leak rate test history of each component. To be consistent with the 
requirements of 10 CFR [Part] 50, Appendix J, Option B, STP proposes 
to include appropriate changes to the Technical Specifications that 
incorporate the necessary revisions associated with 10 CFR [Part] 
50, Appendix J, Option B.
    The proposed amendment represents the conversion of current 
Technical Specification requirements to maintain consistency with 
those requirements specified by 10 CFR [Part] 50, Appendix J, Option 
B. The proposed changes are consistent with the current safety 
analyses. Implementation of these changes will provide continued 
assurance that specified parameters associated with containment 
integrity will remain within acceptance limits, and will not 
significantly increase the probability or consequences of a 
previously evaluated accident.
    Some proposed changes represent minor relaxations in current 
Technical Specification requirements, but are based on the 
requirements specified by Option B of 10 CFR [Part] 50, Appendix J. 
Changes are consistent with the current safety analyses and 
determined to represent sufficient requirements for the assurance 
and reliability of equipment assumed to operate in the safety 
analyses, and provide continued assurance that specified parameters 
associated with containment integrity remain within their acceptance 
limits. These changes will not significantly increase the 
probability or consequences of a previously evaluated accident.
    The systems affecting containment integrity related to this 
proposed amendment request are not assumed in any safety analyses to 
initiate any accident sequence. The probability of any accident 
previously evaluated is not increased by this proposed amendment. 
The proposed changes to Technical Specification LCOs or SRs maintain 
an equivalent level of reliability and availability for all affected 
systems. The proposed amendment does not increase the consequences 
of any accident previously evaluated.
    There is no change to the consequences of an accident previously 
evaluated because maintaining leakage within the analyzed limit 
assumed for any associated accident analyses does not adversely 
affect either the on-site or off-site dose consequences resulting 
from an accident. There is no adverse impact on the probability of 
accident initiators. There is no significant increase in the 
probability of any previously analyzed accident. A plant specific 
risk-based analysis of Appendix J performed for STP indicates the 
containment penetration leakage dose rate contribution to the total 
dose rate in person-rem is insignificant.
    2) Create the possibility of a new or different kind of accident 
from any accident previously evaluated because:
    10 CFR [Part] 50, Appendix J, Option B specifies, in part, that 
a Type A test which measures both the containment system overall 
integrated leakage rate at containment pressure and system 
alignments assumed during a large break LOCA [loss-of-coolant 
accident], and demonstrates the capability of primary containment to 
withstand an internal pressure load, may be conducted at an interval 
based on the performance of the overall containment system. The 
acceptable leakage rates are specified in the plant's Technical 
Specifications. For Type B and Type C tests, intervals are proposed 
based on the performance history of each component. Acceptance 
criteria for each component is based upon demonstration that the sum 
leakage rates at design basis pressure conditions for applicable 
penetrations, is within the limit specified in the Technical 
Specifications.
    The proposed amendment represents the conversion of current 
Technical Specification requirements to maintain consistency with 
those requirements specified in 10 CFR [Part] 50, Appendix J, Option 
B. The proposed changes are consistent with the current safety 
analyses. Some minor relaxations in current Technical Specification 
requirements, associated with containment integrity are based on 
generic guidance provided in Option B, NEI 94-01 and ANSI/ANS 56.8, 
1994. These changes do not involve revisions to the design of the 
station. Some of the changes may involve revision in the testing of 
components; however, these are in accordance with the STP current 
safety analyses and provide for appropriate testing or surveillance 
that are consistent with 10 CFR [Part] 50, Appendix J, Option B. The 
proposed changes will not introduce new failure mechanisms beyond 
those already considered in the current safety analyses.
    The proposed amendment has been reviewed for acceptability 
considering similarity of system or component design affecting 
containment integrity. No new modes of operation are introduced by 
the proposed changes. Surveillance requirements are changed to 
reflect corresponding changes associated with Option B of 10 CFR 
[Part] 50, Appendix J and improvements in technique or interval of 
leak rate testing performance. The proposed changes maintain, at 
minimum, the present level of operability of any system that affects 
containment integrity. The proposed changes do not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    The associated systems that affect leak rate integrity related 
to the proposed amendment, are not assumed in any safety analysis to 
initiate any accident sequence. The proposed surveillance 
requirements for any affected systems are consistent with the 
current requirements specified within the Technical Specifications 
and are consistent with the requirements of Option B of 10 CFR 
[Part] 50, Appendix J. The proposed surveillance requirements 
maintain an equivalent level of reliability and availability of all 
affected systems and therefore, does not increase the consequences 
of any previously evaluated accident.
    3) Involve a significant reduction in the margin of safety 
because:
    The provisions specified in Option B of 10 CFR [Part] 50 
Appendix J allow changes to Type A, Type B, and Type C test 
intervals based upon the performance of past leak rate tests. The 
effect of extending containment leakage rate testing intervals has a 
corresponding increase in the likelihood of containment leakage. The 
degree to which intervals can be extended is a direct function of 
the potential effect to existing safety margins and the public 
health and safety that can occur due to an increased likelihood of 
containment leakage. 10 CFR [Part] 50 Appendix J, Option B allows 
longer intervals between leakage tests based on performance trends 
but does not increase the leakage acceptance criteria. La [maximum 
allowable leakage rate] is still limited to 0.3 wt%/day. By 
referencing the Containment Leakage Rate Testing Program in LCO 
3.6.1.2 ACTION, the point at which ACTION is required is increased 
from .75 La to 1.0 La. This makes the specification consistent with 
the intent of having margin between an AS-LEFT leakage of less than 
or equal to .75 La and maintaining operability with less than or 
equal to 1.0 La.
    Changing Appendix J test intervals from those currently provided 
in the Technical Specification to those provided in 10 CFR [Part] 
50, Appendix J, Option B, slightly increases the risk associated 
with Type A, Type B, and Type C specified accident sequences. 
Historical data suggests that increasing the Type C test interval 
can slightly increase the associated risk; however, this is 
compensated by the corresponding risk reduction benefits associated 
with reduction in component cycling, stress, and wear associated 
with increased test intervals. When considering the total integrated 
risk which includes all analyzed accident sequences, the risk 
associated with increasing test intervals is negligible. A plant 
specific risk-based analysis of Appendix J performed for STP 
indicates the containment penetration leakage dose rate contribution 
to total dose rate in person-rem is insignificant.
    STP proposes to revise the Technical Specifications to be 
consistent with those provisions specified in Option B of 10 CFR, 
Appendix J. The proposed changes are consistent with the STP current 
safety analyses. These proposed changes do not involve revisions to 
the design of the station. The proposed changes will maintain the 
same level of reliability of equipment associated with containment 
integrity assumed to operate in the safety analysis, and provide 
continued assurance that specified parameters affecting plant leak 
rate integrity will remain within acceptance limits. The proposed 
changes provide continued assurance of leakage integrity of 
containment without adversely affecting the public health and safety 
and will not significantly reduce existing safety margins. Plant 
specific risk-based analysis indicates sufficient technical 
justification exists to further extend the limits beyond those 
allowed by Option B.
    The proposed amendment to the Technical Specifications 
implements present requirements, or the requirements in accordance 
with the guidelines set forth in Option B of 10 CFR [Part] 50, 
Appendix J. NUREG-1493, ``Performance-Based Containment Leak-Test 
Program,'' served as the technical basis for Option B. STP

[[Page 28618]]

performed a plant specific risk-based analysis of containment 
penetration leakage dose utilizing the same methodology used in 
NUREG-1493. The analysis indicates the containment penetration 
leakage dose rate contribution to the total dose rate in person-rem 
is insignificant. This plant specific analysis serves to validate 
the applicability of the proposed changes for STP. The proposed 
changes have been approved by the NRC, are applicable to STP, 
maintain necessary levels of system or component reliability 
affecting containment integrity, and do not involve a significant 
reduction in the margin of safety.
    The performance-based approach to leakage rate testing concludes 
the impact on public health and safety due to revised testing 
intervals is negligible. The proposed amendment will not reduce 
availability of systems associated with containment integrity when 
required to mitigate accident conditions; therefore, the proposed 
changes do not involve a significant reduction in the margin of 
safety.
    Guidance has been provided in ``Final Procedures and Standards 
on No Significant Hazards Considerations,'' Final Rule, 51 FR 7744, 
for the application of standards to license change requests for 
determination of the existence of significant hazards 
considerations. This document provides examples of amendments which 
are and are not considered likely to involve significant hazards 
considerations.
    This proposed amendment does not involve a significant 
relaxation of the criteria used to establish safety limits, a 
significant relaxation of the bases for limiting safety system 
settings or a significant relaxation of the bases for LCOs. 
Therefore, based on the guidance provided in the Federal Register 
and criteria established in 10 CFR 50.92(c), the proposed change 
does not constitute a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Local Public Document Room location: Wharton County Junior College, 
J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 77488
    Attorney for licensee: Jack R. Newman, Esq., Morgan, Lewis & 
Bockius, 1800 M Street, N.W., Washington, DC 20036-5869
    NRC Project Director: William D. Beckner

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

    Date of amendment requests: December 14, 1995
    Description of amendment requests: The proposed amendments would 
revise the Administrative Control (Chapter 6) Section and other 
affected Sections of the Prairie Island Technical Specifications to 
generally conform with NUREG-1431, Standard Technical Specifications, 
Westinghouse Plants, Revision 1, dated April 7, 1995.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendment will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    Operation of the Prairie Island plant in accordance with the 
proposed changes does not involve a significant increase in the 
probability or consequences of an accident previously evaluated. 
None of the proposed changes involve a physical modification to the 
plant, a new mode of operation or a change to the Updated Safety 
Analysis Report transient analyses. These proposed amendments 
generally conform to the guidance of NUREG-1431, Revision 1, Section 
5.0 which was previously reviewed, accepted and issued by the NRC.
    Some Section 5.0 Specifications in NUREG-1431 were not 
incorporated in this License Amendment Request. These Specifications 
were not proposed because they 1) specify requirements not currently 
in the Prairie Island Technical Specifications or otherwise 
committed to, 2) are addressed elsewhere in the current Technical 
Specifications, or 3) the current Technical Specifications level of 
commitment is maintained. In all these instances, the NRC has 
previously reviewed and approved the proposed level of commitment 
through the issuance of the current Prairie Island Technical 
specifications.
    The proposed changes, in themselves, do not reduce the level of 
qualification or training such that personnel requirements would be 
decreased.
    In total these changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes do not create the possibility of a new or 
different kind of accident from any accident previously evaluated 
because the proposed changes, in themselves, do not introduce a new 
mode of plant operation, surveillance requirement or involve a 
physical modification to the plant. These proposed amendments 
generally conform to the guidance of NUREG-1431, Revision 1, Section 
5.0 which was previously reviewed, accepted and issued by the NRC.
    Some Section 5.0 Specifications in NUREG-1431 were not 
incorporated in this License Amendment Request. These Specifications 
were not proposed because they 1) specify requirements not currently 
in the Prairie Island Technical Specifications or otherwise 
committed to, or 2) are addressed elsewhere in the current Technical 
Specifications. Other features are not fully implemented but rather, 
the current Technical Specification level of commitment is 
maintained. In all these instances, the NRC has previously reviewed 
and approved the proposed level of commitment through the issuance 
of the current Prairie Island Technical Specifications.
    In general, the proposed changes are administrative in nature. 
The changes propose to revise, delete or relocate Specifications 
within the Technical Specifications or from the Technical 
Specifications to the Updated Safety Analysis Report, plant 
procedures or the Operational Quality Assurance Plan through which 
adequate control is maintained. The proposed changes do not alter 
the design, function, or operation of any plant components and 
therefore, no new accident scenarios are created.
    Therefore, the possibility of a new or different kind of 
accident from any accident previously evaluated would not be created 
[by] these amendments.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety because the Current Technical Specifications 
requirements for safe operation of the Prairie Island plant are 
maintained or increased. The proposed changes are administrative in 
nature and do not involve a physical modification to the plant, a 
new mode of operation or a change to the Updated Safety Analysis 
Report transient analyses. The proposed changes do not alter the 
scope of equipment currently required to be operable or subject to 
surveillance testing nor does the proposed change affect any 
instrument setpoints or equipment safety functions.
    Therefore, a significant reduction in the margin of safety would 
not be involved with these amendments.
    Based on the evaluation describe above, and pursuant to 10 CFR 
Part 50, Section 50.91, Northern States Power Company has determined 
that operation [of] the Prairie Nuclear Generating Plant in 
accordance with the proposed license amendment request does not 
involve any significant hazards considerations as defined by Nuclear 
Regulatory Commission regulations in 10 CFR Part 50, Section 50.92.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Minneapolis Public Library, 
Technology and Science Department, 300 Nicollet Mall, Minneapolis, 
Minnesota 55401
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts, and 
Trowbridge, 2300 N Street, NW, Washington, DC 20037

[[Page 28619]]

    NRC Project Director: Mark Reinhart (Acting Director)

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

    Date of amendment requests: February 15, 1996
    Description of amendment requests: The proposed amendments would 
revise the combined Technical Specifications (TS) for the Diablo Canyon 
Nuclear Power Plant, Unit Nos. 1 and 2 to revise Technical 
Specification 3.5.2, ``ECCS Subsystems - Tavg Greater Than or Equal to 
350 deg.F,'' to change the allowed outage time for any one safety 
injection pump from 72 hours to 7 days. The specific TS change proposes 
to add a new footnote that increases the allowed outage time (AOT) for 
one safety injection (SI) pump from 72 hours to 7 days for performance 
of non-routine, emergent maintenance and requires review by the Plant 
Staff Review Committee (PSRC), and requires Plant Manager approval 
prior to exceeding 72 hours.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed allowed outage time (AOT) extension does not change 
the operating practices of Diablo Canyon Power Plant (DCPP). 
Although the proposed change increases the allowed time in which the 
safety injection (SI) system may be out of service for maintenance 
or testing, this extended AOT will only be used in emergent 
circumstances.
    Increasing the AOT for the SI pumps does not involve physical 
alteration of any plant equipment and does not affect analysis 
assumptions regarding functioning of required equipment designed to 
mitigate the consequences of accidents. Further, the severity of 
postulated accidents and resulting radiological effluent releases 
will not be affected by the increased AOT.
    Finally, the probabilistic risk assessment determined that the 
increase in the core damage probability is not considered 
significant.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed increase to the SI pump AOTs does not change the 
method by which DCPP operates. Further, the proposed change would 
not result in any physical alteration to any plant system, and there 
would not be a change in the method by which any safety related 
system performs its function.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    There is no safety analysis impact since the extension of the SI 
pump AOT interval will have no effect on any safety limit, 
protection system setpoint, or limiting condition of operation. 
There is no hardware change that would impact existing safety 
analysis acceptance criteria.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Project Director: William H. Bateman

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

    ]Date of application request: April 17, 1996
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3/4.3 to support a future 
modification to replace existing digital portions of the main steam and 
feedwater isolation system (MSFIS) with digital processor equipment and 
would authorize revision of the FSAR to include a description of the 
MSFIS modification. The MSFIS modification is a change to the facility, 
as described in the safety analysis report, that involves an unreviewed 
safety question. The modification involves an unreviewed safety 
questions because: (1) the MSFIS design will use software which could 
result in a common mode failure, (2) the original NRC review of the 
MSFIS did not evaluate 2 out of 3 coincidence circuitry, which could 
introduce new system failure modes, and (3) the MSFIS modification 
utilizes manual handswitches that could introduce new system failure 
modes. The NRC will review the modification in accordance with 10 CFR 
50.59(a)(2) in conjunction with the review of the proposed TS 
amendment.
    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 addition of the MSFIS actuation logic and relays to the TS 
has no adverse impact on the probability of occurrences or the 
consequences of an accident. The proposed amendment does not change 
or alter the design assumptions for the systems or components used 
to mitigate the consequences of an accident and the methodologies 
used in the accident analysis remain unchanged. The operating limits 
will not be changed.
    No design basis accidents will be affected by this design change 
since the logic which currently exists will continue to be 
performed. Thus, the radiological consequences will not change.
    The system response time is enveloped by the current 5 second 
valve stroke time. The MSFIS response time will be less than 500 
msec.
    A common mode software failure could exist if both separation 
groups have their PLCs [programmable logic controllers] (3 per train 
- six total) malfunction at the same time. However, a diverse means 
of isolating the feedwater lines exists given the ability of the 
Main Feed Control Valves to close on a Feedwater Isolation Signal. 
The MSIVs [main steam isolation valves] do not have a diverse means 
of isolating their respective steam lines if a common mode software 
failure occurs. As a result, this modification provides a means to 
manually fast close the valves at the MSFIS cabinets. The operators 
will be alerted of the failure conditions of any PLC logic channel 
via MCB [main control board] annunciators and indicators. This 
failure mode has a low probability of occurrence based upon the 
inherent quality of the design provided by the V&V [verification & 
validation] process. Therefore, the accident consequences are not 
increased for this failure mode.
    The test panel in the MSFIS cabinets has been laid out to 
provide the same functions as the existing test panel, except that 
PLC status indication and coincidence logic test functions are 
provided. The Emergency Override Panel, located below the Test 
Panel, provides the operator with the ability to bypass the FWIS 
[feedwater isolation signal] signal and manually fast close each 
MSIV as required by the Emergency Operating Procedures. The MSIV 
manual FC [fast close] switch operation is necessary for a diverse 
means of operation for software common mode failures. The FWIS 
bypass switch will allow main feedwater flow to be re-established to 
each Steam Generator.

[[Page 28620]]

    The replacement system is functionally the same as the current 
system since it performs the same logic, receives the same inputs, 
and produces the same outputs. However, the system is more reliable 
and possesses triple redundant logic. Therefore, the probability of 
malfunction will not be increased.
    The electrical load of the A-B PLC equipment and existing 48 VDC 
[volts direct current] actuation relays is less than that of the 
existing equipment so the system will not require any additional 
cooling over the existing equipment. Proper grounding is provided 
for the PLC 5 VDC and actuation relay 48 VDC power supplies, which 
are electrically isolated from each other.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The addition of the MSFIS actuation logic and relays to the TS 
will not create a new type of accident or malfunction than any 
previously evaluated in the Safety Analysis Report. The safety 
functions of the system are not changed in any manner, nor is the 
reliability of any structure, system or component reduced. All 
design and performance criteria continue to be met. Since the safety 
functions and reliability are not adversely affected, the proposed 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    The operator's ability to adequately respond to an accident is 
not hindered by the man-machine interface added as a result of this 
modification since the operator interface is similar to the current 
system and the MCB controls will not change. The operators will be 
alerted to system malfunctions through annunciation. The current 
system has a status output for each MSIV and FIV [feedwater 
isolation valve] valve on the Engineered Safety Feature Status 
Panel, which will be maintained. In addition, an isolated plant 
annunciator interface will provide a MSFIS Channel Failure plant 
annunciator window for both trains. Training will be provided to the 
technicians, engineers, and operators on the new features of the 
system prior to installation. Therefore, this modification does not 
increase the consequential effects due to the man-machine interface.
    The system is compatible with the normal and accident 
environments and will be seismically qualified in accordance with 
the SNUPPS [standardized nuclear unit power plant system] seismic 
spectra profile. The equipment will be qualified for Electromagnetic 
Interference concerns in accordance with EPRI [Electric Power 
Research Institute] document TR-102323-EPRI Guideline and will meet 
the EPRI EMI [electromagnetic interference] limiting practices.
    The system has the same failure mode upon loss of power as the 
current system and behaves similarly upon power restoration. A loss 
of power will not result in a MSFIS actuation.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The addition of the MSFIS actuation logic and relays to the TS 
will not affect or change a safety limit or affect plant operations. 
This change will not reduce the margin of safety assumed in the 
accident analysis nor reduce any margin of safety as defined in the 
basis for any TS.
    The system response time for any given valve will not exceed the 
required valve stroke time. Since the MSFIS does not contain any 
analog channels, no channel trip accuracies are impacted.
    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

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

    Date of amendment request: October 25, 1995
    Description of amendment request: The proposed changes would 
provide an allowed outage time of 14 days for the pressurizer power-
operated relief valve (PORV) nitrogen accumulators, as well as provide 
separate action statements for the PORV depending on the reason for the 
PORV inoperability during plant operation in power Modes 1, 2, or 3.
    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.
    Specifically, operation of North Anna Power Station in 
accordance with the proposed Technical Specifications changes will 
not:
    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The PORVs are assumed to mitigate the consequences of a steam 
generator tube rupture as described in the North Anna UFSAR [Updated 
Final Safety Analysis Report] as well as to limit the undesired 
opening of the pressurizer safety valves for a primary overpressure 
event. The proposed action statements ensure that the steam 
generator tube rupture accident analysis requirements are met. The 
proposed Technical Specification changes require the backup nitrogen 
supply be available for the PORVs to be consideredoperable and add 
action statements and surveillance requirements for the nitrogen 
supply commensurate with its significance. The proposed action 
statements enhance the availability of the automatic actuation of 
the PORVs by not requiring the block valves to be closed when the 
backup nitrogen supplies are inoperable. The proposed surveillance 
requirements enhance the reliability of the backup nitrogen supply 
to the PORVs by verifying that there is sufficient nitrogen pressure 
in the accumulators for the PORVs to perform their design function. 
The proposed Technical Specification changes do not change any 
accident analyses, therefore, the probability of any accident and 
its resulting consequences are not increased.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed Technical Specification changes do not involve any 
physical modification to the plant or result in a change in a method 
of operation. The backup nitrogen supply continues to be required 
for PORV operability. The proposed Technical Specification changes 
provide operational flexibility and ensure the availability of the 
PORVs using the normal supply of instrument air while the backup 
nitrogen supply is being restored. This also prevents undesirable 
challenges to the pressurizer safety valves. The new surveillance 
requirements verify that there is sufficient nitrogen pressure in 
the accumulators for the PORVs to perform their design functions.
    3. Involve a significant reduction in a margin of safety.
    The proposed Technical Specification changes do not affect any 
safety limits or limiting safety system settings. The availability 
of the PORVs will be maintained as required in Generic Letter 90-06. 
The proposed Technical Specifications will continue to ensure that 
the PORVs will be capable of performing their intended functions.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: The Alderman Library, Special 
Collections Department, University of Virginia, Charlottesville, 
Virginia 22903-2498
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219
    NRC Project Director: Eugene V. Imbro

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

    Date of amendment request: April 24, 1996

[[Page 28621]]

    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 15.7, ``Radiological 
Effluent Technical Specifications (RETS).'' Portions of the RETS would 
be moved to licensee-controlled documents consistent with Nuclear 
Regulatory Commission guidance on TS improvements. Changes to other 
sections of the TSs are also proposed consistent with the removal of 
portions of the RETS.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:
    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed amendment simplifies the RETS and implements the 
recommendations of GL 89-01 and of GL 95-10. The proposed change 
relocates the operational requirements of RETS but keeps the 
programmatic controls for these requirements in the Technical 
Specifications. Therefore, the proposed changes are administrative 
in nature and do not affect plant operations. Hence, the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident previously evaluated because no 
safety-related equipment, safety function, or plant operation will 
be altered as a result of this proposed change. Also, the changes 
are unrelated to the initiation and mitigation of accidents and 
equipment malfunctions addressed in the Final Safety Analysis 
Report.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any accident previously 
evaluated?
    As stated above, the proposed action is the relocation of the 
RETS procedural details to various manuals while retaining the 
administrative controls in RETS. The relocation is consistent with 
the intent of the guidance of GL 89-01 and of GL 95-10. It is 
administrative and has no impact on plant operation or safety. No 
safety-related equipment, safety function, or plant operation will 
be altered as a result of this proposed change. No changes to plant 
components or structures are introduced which could create new 
accidents or malfunctions not previously evaluated.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated because no new accident scenario is created and no 
previously evaluated accident scenario is changed by the relocation 
of the procedural details of RETS from one controlled document to 
another.
    3. Does the proposed amendment involve a significant reduction 
in the margin of safety?
    The proposed change does not include a change to any plant 
structure, system, component, or operation. The proposed changes do 
not alter the basic regulatory requirements and do not affect any 
safety analyses. The proposed change is administrative. The 
procedural details of the current RETS are relocated while the 
programmatic controls consistent with regulatory requirements, 
including controls on revisions to the manuals receiving the RETS 
procedural details, the Environmental Manual (EM), Radiological 
Effluent Control Program Manual (RECM), Offsite Dose Calculation 
Manual (ODCM), and Process Control Program (PCP), remain in RETS.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    NRC Project Director: Gail H. Marcus

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

    Date of amendment request: April 29, 1996
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Section 15.3.14, ``Fire Protection 
System,'' and Section 15.4.15, ``Fire Protection System.'' These 
specifications would be relocated to other licensee-controlled 
documents in accordance with Nuclear Regulatory Commission generic 
guidance. Additional administrative changes consistent with the 
relocation are also proposed.
    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 this facility under the proposed Technical 
Specifications change will not increase the probability or 
consequences of an accident previously evaluated.
    This change request proposes to remove certain fire protection 
program requirements from the Point Beach Technical Specifications 
and incorporate them into the Final Safety Analysis Report (FSAR) 
and the Fire Protection Evaluation Report (FPER). No requirements 
are eliminated, modified, or de-emphasized by this change. The 
proposed amendment ensures that any future changes to the fire 
protection program will be subject to an appropriate evaluation in 
accordance with NRC regulations to ensure that there are no 
unreviewed safety questions.
    Therefore, these proposed changes are administrative in nature. 
There are no proposed changes to the physical plant or the processes 
which ensure the plant's capability to mitigate fires and achieve 
safe shutdown. Therefore, there is no potential effect on the 
probability or consequences of previously evaluated accidents.
    2. Operation of this facility under the proposed Technical 
Specifications change will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    New or different accidents can only be created by new or 
different accident initiators or sequences. Because there are no 
proposed changes to the physical plant or the processes which ensure 
the plant's fire protection capability, new or different kinds of 
accident initiators will not be introduced by this change. The 
proposed changes are administrative in nature.
    3. Operation of this facility under the proposed Technical 
Specifications change will not create a significant reduction in a 
margin of safety.
    The margins of safety for Point Beach are based on the design 
and operation of the reactor and containment and the safety systems 
that provide their protection. Because there are no proposed changes 
to the physical plant or the processes which ensure the plant's fire 
protection capability, there will be no effect on the reactor, 
reactor containment, or the safety systems which provide their 
protection. Therefore, the proposed changes will not create a 
reduction in a margin of safety. The proposed changes are 
administrative in nature.
    Additionally, the proposed revision to Point Beach's operating 
license will not allow Wisconsin Electric to make changes to the 
approved fire protection program without prior approval of the 
Nuclear Regulatory Commission should these proposed changes 
adversely affect the ability to achieve and maintain safe shutdown 
in the event of a fire. In accordance with NRC Generic Letter 86-10, 
any proposed change to the approved fire protection program requires 
the performance of a 10 CFR 50.59 evaluation and a fire hazards 
analysis. Should these evaluations indicate that the ability to 
reach and maintain safe shutdown has been adversely affected, prior 
NRC review and approval will be obtained prior to effecting the 
changes. Thus, a significant reduction in a margin of safety cannot 
occur.
    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: Joseph P. Mann Library, 1516

[[Page 28622]]

Sixteenth Street, Two Rivers, Wisconsin 54241
    Attorney for licensee: Gerald Charnoff, Esq., Shaw, Pittman, Potts, 
and Trowbridge, 2300 N Street, NW., Washington, DC 20037
    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: May 16, 1996. This supersedes the 
October 24, 1995, request published in the Federal Register on November 
27, 1995 (60 FR 58409).
    Description of amendment request: This license amendment request 
proposes to revise Surveillance Requirement 4.7.6.e.4 to reflect a 
proposed design change to the output rating, from 15kW to 5kW, of the 
charcoal filter adsorber unit heater in the pressurization system 
portion of the control room emergency ventilation system (CREVS). 
Surveillance Requirements 4.7.6.c.2, 4.7.6.d, and 4.9.13.b and c, are 
also being revised to reflect a proposed change to the acceptance 
criteria for the testing of carbon samples from the CREVS charcoal 
adsorbers and the auxiliary/fuel building emergency exhaust system 
charcoal adsorbers. Surveillance Requirement 4.7.7.a for the auxiliary 
building portion of the auxiliary/fuel building emergency exhaust 
system is also affected by this proposed change. However, since 
Surveillance Requirement 4.7.7.a refers to Surveillance Requirements 
4.9.13.b and c, no changes to 4.7.7.a are required.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The design function of the filter adsorber unit heater in the 
pressurization system portion of CREVS is to reduce the relative 
humidity of the air entering the charcoal filter beds to 70% 
relative humidity. Although the original design specified a heater 
with a rating of 15 kW, review of the design basis calculation for 
this system indicates that only about 3.13 kW is actually required 
(including applicable margins to allow for voltage variations). The 
proposed change to the CREVS heaters output rating from 15 kW to 5 
kW will not affect the method of operation of the system, and the 
new heater capacity will still exceed filter operational 
requirements and safety margin. Neither the heater change nor the 
charcoal testing protocol changes will affect system operation or 
performance, nor do they affect the probability of any event 
initiators. These changes do not affect any Engineered Safety 
Features actuation setpoints or accident mitigation capabilities. 
Therefore, the proposed changes will not significantly increase the 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated in the USAR [Updated Safety Analysis 
Report].
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The requested change to the CREVS heaters' output rating and the 
changes to the charcoal sample testing protocol will not affect the 
method of operation of the systems, and the new heater capacity will 
still exceed filter operational requirements and safety margin by a 
significant amount. The proposed changes only affect the heater size 
in the system and the testing criteria for the charcoal samples. No 
new or different accident scenarios, transient precursors, failure 
mechanisms, or limiting single failures will be introduced as a 
result of these changes. Therefore, the possibility of a new or 
different kind of accident other than those already evaluated will 
not be created by this change.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The requested change to the CREVS heaters' output rating will 
reduce the heater output of the system, but the new heater output 
will still exceed filter operational requirements and safety margin 
by a significant amount. In addition, the reduction in heat load 
output from the heater will increase the design margin between the 
cooling capacity of the system air conditioning units and the 
building heat load. The new charcoal adsorber sample laboratory 
testing protocol is more stringent than the current testing practice 
and more accurately demonstrates the required performance of the 
adsorbers following a design basis LOCA [loss-of-coolant accident]. 
Therefore, these changes will not reduce the margin of safety of the 
HVAC [heating, ventilation, and air conditioning] systems' 
operation.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, N.W., Washington, D.C. 20037
    NRC Project Director: William H. Bateman

Notice Of Issuance Of Amendments To Facility Operating Licenses

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

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

    Date of application for amendments: January 12, 1996, as 
supplemented March 4, April 3 and April 10, 1996.
    Brief description of amendments: The amendments revise the 
Technical Specification so that the containment integrated leak rate 
Type A testing will now be performed consistent with the revised 10 CFR 
Part 50, Appendix J, Option B, by referring to Regulatory Guide 1.163, 
``Performance-Based Containment Leak-Test Program.'' No

[[Page 28623]]

changes to implement Option B for the Type B and Type C tests were 
requested by the licensee at this time.
    Date of issuance: May 13, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 144 and 138
    Facility Operating License Nos. NPF-35 and NPF-52: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: January 21, 1996 (61 FR 
3498); and April 10, 1996 (61 FR 15988) The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
May 13, 1996.No significant hazards consideration comments received: No
    Local Public Document Room location: York County Library, 138 East 
Black Street, Rock Hill, South Carolina 29730

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

    Date of application for amendments: March 5, 1996
    Brief description of amendments: These amendments delete the 
requirement to perform a pressurizer heater surveillance test and 
change the requirement for containment visual inspection to prevent 
sump clogging. These changes are in accordance with selected line items 
from NRC Generic Letter 93-05, ``Line-Item Technical Specification 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation.''
    Date of issuance: May 13, 1996
    Effective date: May 13, 1996
    Amendment Nos. 184 and 178Facility Operating Licenses Nos. DPR-31 
and DPR-41: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 10, 1996 (61 
FR15989) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 13, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Florida International 
University, University Park, Miami, Florida 33199.

Houston Lighting & Power Company, City Public Service Board of San 
Antonio, Central Power and Light Company, City of Austin, Texas, 
Docket Nos. 50-498, South Texas Project, Unit 1, Matagorda County, 
Texas

    Date of amendment request: January 22, 1996, as supplemented by 
letter dated April 18, 1996.
    Brief description of amendment: The amendment modified the steam 
generator tube plugging criteria in Technical Specification 3/4.4.5, 
Steam Generators, and the associated Bases, to allow the implementation 
of alternate steam generator tube plugging criteria for the tube-to-
tubesheet joints (known in the industry as F*) for Unit 1.
    Date of issuance: May 14, 1996Effective date: May 14, 1996
    Amendment No.: 82
    Facility Operating License No. NPF-76: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7553) The additional information contained in the supplemental 
letter dated April 18, 1996, was clarifying in nature and thus, within 
the scope of the initial notice and did not affect the staff's proposed 
no significant hazards consideration determination.The Commission's 
related evaluation of the amendments is contained in a Safety 
Evaluation dated May 14, 1996. No significant hazards consideration 
comments received: No
    Local Public Document Room location:  Wharton County Junior 
College, J. M. Hodges Learning Center, 911 Boling Highway, Wharton, TX 
77488

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

    Date of application for amendment: February 9, 1996, as 
supplementedMarch 15, 1996, and April 22, 1996.
    Brief description of amendment: The amendment revised the 
Administrative Controls Section 5.6.6 of the Ginna Technical 
Specifications to incorporate a reference to the methodology for 
determining pressure/temperature and low-temperature overpressure 
protection limits.
    Date of issuance: May 23, 1996
    Effective date: May 23, 1996
    Amendment No.: 64
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7557) The March 15, 1996, and April 22, 1996, letters provided 
clarifying information that did not change the initial proposed no 
significant hazards consideration determination. The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated May 23, 1996.No significant hazards consideration comments 
received: No
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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

    Date of application for amendment: February 9, 1996
    Brief description of amendment: This amendment changes the 
setpoints for the steam generator water level-high feedwater isolation 
function.Date of issuance: May 20, 1996
    Effective date: May 20, 1996
    Amendment No.: 63
    Facility Operating License No. DPR-18: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 28, 1996 (61 
FR 7558) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 20, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Rochester Public Library, 115 
South Avenue, Rochester, New York 14610.

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

    Date of application for amendment: November 21, 1995, as 
supplemented on March 13, 1996.
    Brief description of amendment: The amendment adds GPU Nuclear 
Corporation as a licensee for the SNEF along with SNEC and transfers 
all management-related responsibilities for the SNEF from SNEC to GPU 
Nuclear Corporation.
    Date of issuance: May 10, 1996
    Effective date: May 10, 1996
    Amendment No.: 13Amended Facility License No. DPR-4: Amendment 
changed the Technical Specifications.
    Date of initial notice in Federal Register: January 31, 1996 (61 FR 
3502). The Commission also published a notice of consideration of 
transfer of control of license pursuant to 10 CFR 50.80 on March 19, 
1996 (61 FR 11231). The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 10, 1996.o 
significant hazards consideration comments received: No
    Local Public Document Room Location: Saxton Community Library, 911 
Church Street, Saxton, Pennsylvania 16678

[[Page 28624]]

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

    Date of application for amendment: December 8, 1995
    Brief description of amendment: The amendment revises the Technical 
Specifications (TS) to: 1) add a new surveillance requirement to 
4.1.2.2, 2) delete 3.1.2.3 and 3.1.2.4, revise 3.4.9.3 to assure that 
only one charging pump is capable of injecting water into the primary 
coolant whenthe reactor is in a shutdown mode, 4) add a new 
surveillance requirement to 4.4.9.3, 5) revise the Emergency Core 
Cooling Water System pump testing acceptance criteria, and 6) revise 
the BASES supporting the above changes.
    Date of issuance: May 10, 1996
    Effective date: 30 days after issuance
    Amendment No.: 134
    Facility Operating License No. NPF-12: Amendment revises the TS.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1635) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 10, 1996.No significant hazards 
consideration comments received: No
    Local Public Document Room location:  Fairfield County Library, 300 
Washington Street, Winnsboro, SC 29180

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 
50-364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston 
County, Alabama.

    Date of amendments request: December 19, 1995, as supplemented by 
letters dated January 5, 1996 and May 3, 1996.
    Brief description of amendments: The amendments replace the 
requirements associated with the control room emergency ventilation 
system contained in Technical Specification Section 3/4.7.7 with 
requirements related to the operation of the control room emergency 
filtration/pressurization system and the control room air conditioning 
system. In addition, a one-time extension to the allowable outage time 
for the control room recirculation filtration system is included to 
facilitate implementation of design modifications.
    Date of issuance: May 21, 1996
    Effective date: As of the date of issuance to be implemented within 
30 days
    Amendment Nos.: 119 and 111
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: January 22, 1996 (61 FR 
1637) The January 5, 1996 and May 3, 1996 letters provided clarifying 
information that did not change the scope of the December 19, 1995, 
application and initial proposed no significant hazards consideration 
determination. The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 21, 1996.No significant 
hazards consideration comments received: No
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

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

    Date of amendment request: April 23, 1996
    Brief description of amendment: The amendment would allow steam 
generator tubes to remain in service with bands of axial degradation in 
the tube sheet region, for the remainder of Cycle 11, provided 
sufficient undegraded tubing remains to satisfy the L<SUP>*-type 
criteria restrictions established by the licensee.
    Date of issuance: May 20, 1996
    Effective date: May 20, 1996
    Amendment No.: 110
    Facility Operating License No. NPF-8. The amendment revised the 
Technical Specifications. Public comments requested as to proposed no 
significant hazards consideration: Yes (61 FR 19092). The 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 May 30, 1996, but indicated that if the Commission makes a 
final no significant hazards consideration determination, any such 
hearing would take place after issuance of the amendment.
    The Commission's related evaluation of the amendment, finding of 
exigent circumstances, and final determination of no significant 
hazards consideration are contained in a Safety Evaluation dated May 
20, 1996.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302

Notice Of Issuance Of Amendments To Facility Operating Licenses And 
Final Determination Of No Significant Hazards Consideration And 
Opportunity For A Hearing (Exigent Public Announcement Or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an

[[Page 28625]]

opportunity for public comment. If comments have been requested, it is 
so stated. In either event, the State has been consulted by telephone 
whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document 
Room, the Gelman Building, 2120 L Street, NW., Washington, DC, and at 
the local public document room for the particular facility involved.
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By July 5, 1996, the licensee 
may file a request for a hearing with respect to issuance of the 
amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should 
consult a current copy of 10 CFR 2.714 which is available at the 
Commission's Public Document Room, the Gelman Building, 2120 L Street, 
NW., Washington, DC and at the local public document room for the 
particular facility involved. If a request for a hearing or petition 
for leave to intervene is filed by the above date, the Commission or an 
Atomic Safety and Licensing Board, designated by the Commission or by 
the Chairman of the Atomic Safety and Licensing Board Panel, will rule 
on the request and/or petition; and the Secretary or the designated 
Atomic Safety and Licensing Board will issue a notice of a hearing or 
an appropriate order. As required by 10 CFR 2.714, a petition for leave 
to intervene shall set forth with particularity the interest of the 
petitioner in the proceeding, and how that interest may be affected by 
the results of the proceeding. The petition should specifically explain 
the reasons why intervention should be permitted with particular 
reference to the following factors: (1) the nature of the petitioner's 
right under the Act to be made a party to the proceeding; (2) the 
nature and extent of the petitioner's property, financial, or other 
interest in the proceeding; and (3) the possible effect of any order 
which may be entered in the proceeding on the petitioner's interest. 
The petition should also identify the specific aspect(s) of the subject 
matter of the proceeding as to which petitioner wishes to intervene. 
Any person who has filed a petition for leave to intervene or who has 
been admitted as a party may amend the petition without requesting 
leave of the Board up to 15 days prior to the first prehearing 
conference scheduled in the proceeding, but such an amended petition 
must satisfy the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-001, 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 20555-001, and to the 
attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

[[Page 28626]]

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

    Date of application for amendments: May 15, 1996
    Brief description of amendments: The amendment revised Surveillance 
Requirement (SR) 4.5.2.d.2 in Technical Specification 3/4 5.2 to state 
that the trisodium phosphate (TSP) contained in the storage baskets in 
containment is in the form of anhydrous TSP, rather than dodecahydrate 
TSP, as currently specified.
    Date of issuance: May 15, 1996
    Effective date: May 15, 1996
    Amendment Nos.: Unit 1 - 107; Unit 2 - 99; Unit 3 - 79
    Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The 
amendment revised the Technical Specifications.Public comments 
requested as to proposed no significant hazards consideration: No.The 
Commission's related evaluation of the amendments, finding of emergency 
circumstances, and final determination of no significant hazards 
consideration are contained in a Safety Evaluation dated May 15, 1996.
    Local Public Document Room location: Phoenix Public Library, 1221 
N. Central Avenue, Phoenix, Arizona 85004
    Attorney for licensee: Nancy C. Loftin, Esq., Corporate Secretary 
and Counsel, Arizona Public Service Company, P.O. Box 53999, Mail 
Station 9068, Phoenix, Arizona 85072-3999
    NRC Project Director: William H. Bateman
    Dated at Rockville, Maryland, this 29th 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. 96-13878 Filed 6-4-96; 8:45 am]
BILLING CODE 7590-01-9