[Federal Register Volume 59, Number 12 (Wednesday, January 19, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: X94-10119]


[[Page Unknown]]

[Federal Register: January 19, 1994]


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NUCLEAR REGULATORY COMMISSION

 

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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 23, 1993, through January 6, 1994. 
The last biweekly notice was published on January 5, 1994 (59 FR 615).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Rules Review and 
Directives Branch, Division of Freedom of Information and Publications 
Services, Office of Administration, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555, and should cite the publication date and page 
number of this Federal Register notice. Written comments may also be 
delivered to Room P-223, Phillips Building, 7920 Norfolk Avenue, 
Bethesda, 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 20555. 
The filing of requests for a hearing and petitions for leave to 
intervene is discussed below.
    By February 18, 1994, 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 20555 and at the local 
public document room for the particular facility involved. If a request 
for a hearing or petition for leave to intervene is filed by the above 
date, the Commission or an Atomic Safety and Licensing Board, 
designated by the Commission or by the Chairman of the Atomic Safety 
and Licensing Board Panel, will rule on the request and/or petition; 
and the Secretary or the designated Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) the nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555, Attention: Docketing and Services 
Branch, or may be delivered to the Commission's Public Document Room, 
the Gelman Building, 2120 L Street, NW., Washington DC 20555, by the 
above date. Where petitions are filed during the last 10 days of the 
notice period, it is requested that the petitioner promptly so inform 
the Commission by a toll-free telephone call to Western Union at 1-
(800) 248-5100 (in Missouri 1-(800) 342-6700). The Western Union 
operator should be given Datagram Identification Number N1023 and the 
following message addressed to (Project Director): petitioner's name 
and telephone number, date petition was mailed, plant name, and 
publication date and page number of this Federal Register notice. A 
copy of the petition should also be sent to the Office of the General 
Counsel, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and 
to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's Public Document Room, the Gelman Building, 2120 L 
Street, NW., Washington, DC 20555, and at the local public document 
room for the particular facility involved.

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

    Date of amendment requests: October 26, 1993
    Description of amendment requests: The proposed amendment would 
change the maximum nominal enrichment of the fuel allowed to be used in 
the reactor core. Specifically, in Technical Specification 5.3.1, 
``Fuel Assemblies,'' the fuel enrichment would change from ``a maximum 
enrichment of 4.05 weight percent U-235'' to ``a maximum radially 
averaged enrichment of 4.30 weight percent U-235 at any axial 
location.''
    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:
    Standard 1 -- Involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The adequacy of a given core design must be demonstrated for each 
core prior to core reloading. The fuel enrichment is only one of the 
factors that must be considered in this determination. The fuel 
enrichment does not directly impact the results of the plant safety 
analysis.
    The Unit 1, 2, and 3 fuel and storage areas have been analyzed 
for a maximum radially averaged enrichment of any axial enrichment 
zone with a fuel assembly of 4.30 w/o U-235. The criticality 
analysis for Palo Verde's spent fuel pool is presented in Section 
D.3, Criticality Safety Analysis. The results of these analyses 
indicate that handling and storage of such fuel assemblies do not 
involve an unreviewed safety question. The results of these analyses 
are within the acceptance criteria defined in TS 5.6.1, 
``Criticality.''
    The applicable codes, standards and regulations of criticality 
safety for spent fuel and new fuel storage include the following:
    - General Design Criterion 62 - Prevention of Criticality in 
Fuel Storage and Handling.
    - NUREG -0800, USNRC Standard Review Plan, Section 9.1.2, Spent 
Fuel Storage and Section 9.1.1, New Fuel Storage.
    - ANSI/ANS-57.2-1983, ``Design Requirements for Light Water 
Reactor Spent Fuel Storage Facilities at Nuclear Power Plants,'' 
Section 6.4.2
    - ANSI/ANS-57.3-1983, ``Design Requirements for New Fuel Storage 
Facilities at Light Water Reactor Plants,'' Section 6.2.4
    - Qualification of Analytical Methodology Used In Spent Fuel 
Storage Rack Analyses, CE Benchmark, 590962-PHD-004 Revision 0 dated 
December 3, 1979.
    These regulations and guides require that for spent fuel racks 
the maximum calculated keff, including margin for uncertainty 
in calculational method and mechanical tolerances, be less than or 
equal to 0.95 with a 95% probability at a 95% confidence level.
    In order to assure the true reactivity will always be less than 
the calculated reactivity, the following conservative assumptions 
were made for spent fuel rack criticality analysis:
    - Pure, unborated water at 68 degrees Fahrenheit is used in all 
calculations,
    - An infinite array with no radial or axial leakage is modeled, 
and
    - Neutron absorption from spacer grids is neglected, i.e., 
replaced by water.
    For the new fuel vault, a dual criteria applies in which the 
maximum calculated keff, including uncertainties, is less than 
or equal to 0.95 when flooded and less than or equal to 0.98 under 
conditions of ``optimum moderation.''
    Because the new fuel vault is normally dry, and low density 
moderation of ``optimum moderation'' produces strong coupling 
between assemblies, the following conservative assumptions are used:
    - The storage array was enclosed on all six sides by a tight 
fitting two foot concrete reflector,
    - Unborated water is introduced uniformly throughout the storage 
array and the space between fuel pins,
    - Water density is varied uniformly from flooded to dry,
    - Neutron absorption from spacer grids is neglected, i.e., 
replaced by water.
    In the new fuel vault criticality analysis, leakage is 
explicitly modeled, because the assumption of an infinite array with 
no radial or axial leakage is unrealistic under conditions of low 
density moderation. Leakage suppresses criticality at low moderator 
density. Without 3-D modeling of the array, erroneously high values 
of keff are calculated. Thus, the assumption on array leakage 
is relaxed, but reflection from the walls, floor and ceiling is 
included.
    In addition to the above discussion of the new fuel vault 
criticality analysis, the following conservative assumptions are 
applied to both analyses:
    - No credit is taken for the presence of burnable poison rods. 
These rods displace fuel rod positions and are an integral part of 
selected fuel assemblies.
    - An upper bound for the fuel density was included in the 
nominal configuration of the fuel densities.
    - The upper statistical bound of the fuel assembly enrichment, 
as based on the fuel fabrication specification, is included in the 
statistical evaluation of uncertainties.
    The criticality analysis of the Palo Verde new and spent fuel 
racks shows that the maximum radially averaged fuel enrichment of 
any axial enrichment zone within a fuel assembly which meets the 
appropriate NRC limit with uncertainties is higher than 4.30 w/o U-
235.
    Although a higher enrichment fuel cycle may result in fuel 
burnup consisting of a slightly different mixture of nuclides and 
inventory, the effect is insignificant because the isotopic mixture 
and inventory of an irradiated assembly is relatively insensitive to 
the fuel assembly's initial enrichment. Therefore, the doses from 
postulated accidents are not significantly affected and continue to 
be acceptable.
    Standard 2 -- Create the possibility of a new or different kind 
of accident from any accident previously analyzed.
    Operation of Palo Verde with the proposed enrichment limit 
change will not create any new or different kinds of accidents from 
those previously evaluated.
    The adequacy of a given core design shall be demonstrated for 
each core prior to reloading. The fuel enrichment is only one of the 
factors that must be considered in this determination.
    Fuel handling and storage of fuel with radially averaged 
enrichment of any axial enrichment zone within a fuel assembly of 
4.30 w/o U-235 does not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    This amendment request will not involve a significant reduction 
in a margin of safety.
    The evaluation performed for each reload core assures that the 
core design meets appropriate safety limits, including a 
consideration of a significant reduction in the margin of safety. 
See response provided in Standard 1 for information pertaining to 
the demonstration of the adequacy of each core design.
    Criticality analyses for fuel assemblies with a maximum radially 
averaged enrichment (of any axial enrichment zone) of 4.30% U-235 
for the Palo Verde Fuel Pool configurations presented in this 
proposed Technical Specification amendment meet the criticality 
acceptance criterion for Keff listed in Technical Specification 
5.6.1.1. Technical Specification 5.6.1.1 states:
    The spent fuel storage racks are designed and shall be 
maintained with:
    a. A Keff equivalent to less than or equal to 0.95 when 
flooded with unborated water, which includes a conservative 
allowance of 2.6% delta k/k for uncertainties as described in 
Section 9.1 of the FSAR.
    . . .
    Based on the above evaluation, this proposed change does not 
constitute a significant hazards consideration.
    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    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: Theodore R. Quay

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

    Date of amendment requests: October 27, 1993
    Description of amendment requests: The proposed amendment would 
revise Technical Specification (TS) 6.9.1.8 to change the frequency for 
submitting the Radioactive Effluent Release Report from semiannual to 
annual, as allowed by the revised 10 CFR 50.36a requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    Standard 1 -- 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 are administrative in nature and do not involve 
any change to the configuration or method of operation of any plant 
equipment that is used to mitigate the consequences of an accident. 
Also, the proposed changes do not alter the conditions or 
assumptions in any of the FSAR accident analyses. Since the FSAR 
accident analyses remain bounding, the radiological consequences 
previously evaluated are not adversely affected by the proposed 
changes. Therefore, it can be concluded that the proposed changes do 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Standard 2 -- 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. 
The proposed changes are administrative in nature and do not involve 
any change to the configuration or method of operation of any plant 
equipment that is used to mitigate the consequences of an accident. 
Accordingly, no new failure modes have been defined for any plant 
system or component important to safety nor has any new limiting 
failure been identified as a result of the proposed changes. Also, 
there will be no change in the types or increase in the amount of 
effluents released offsite. Therefore, it can be concluded that the 
proposed changes do not create the possibility of a new or different 
kind of accident from any accident previously evaluated.
    Standard 3 -- Involve a significant reduction in a margin of 
safety.
    The proposed changes do not involve a significant reduction in a 
margin of safety. The proposed changes are administrative in nature 
and do not adversely impact the plant's ability to meet applicable 
regulatory requirements related to liquid and gaseous effluents, and 
solid waste releases. The proposed changes would also eliminate an 
unnecessary burden of governmental regulation without reducing 
protection for public health and safety. Therefore, it can be 
concluded that 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 
that review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Phoenix Public Library, 12 
East McDowell Road, Phoenix, Arizona 85004
    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: Theodore R. Quay

Baltimore Gas and Electric Company, Docket Nos. 50-317 and 50-318, 
Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert 
County, Maryland

    Date of amendments request: December 8, 1993
    Description of amendments request: As an active participant in the 
industry-NRC improved Standard Technical Specifications (STS) 
implementation effort, Baltimore Gas and Electric Company (BG&E) 
volunteered to develop criteria for determining what requirements are 
appropriate for inclusion in the Design Features section of the 
Technical Specifications (TSs), and to submit a lead plant license 
amendment applying those criteria. This proposed license amendment 
justifies the adoption of the ``Design Features'' section of the 
Combustion Engineering STS for the Calvert Cliffs Nuclear Power Plant 
and then applies the developed criteria to the STS. The proposed 
criteria for determining what requirements should be placed in the 
Design Features section of the TSs are:
    1. The amount, kind, and source of special nuclear material 
required;
    2. The place of the use of the special nuclear material; and
    3. Those features of the facility such as materials of construction 
and geometric arrangements, which, if altered or modified, would have 
an immediate and significant effect on safety and are not covered in 
the safety limits, limiting conditions for operation or surveillance 
requirements of the Technical Specifications.
    The Design Features section is Section 5.0 for the Calvert Cliffs 
TS and Section 4.0 for the STS. Specifically, the following changes are 
proposed to the Calvert Cliffs TSs and the STS for the Design Features 
section.
    Section 5.1 Site
    Sections 5.1.1 and 5.1.2 contain maps of the site boundary and low 
population zones, respectively. These maps also show the major 
structures, effluent release points, meteorological tower location, and 
the minimum exclusion area radius. The STS requires inclusion of the 
site and exclusion area boundaries and the low population zone and 
allows descriptions instead of maps. These maps or descriptions do not 
meet any of the criteria. However, Criteria 2 requires that the place 
of use of the special nuclear material be described. BG&E proposes to 
delete the existing Sections 5.1.1 and 5.1.2 and to create Section 5.1, 
entitled ``Site Location,'' and to include a text description of the 
location of the site. The current information and maps in the sections 
will be relocated to the Updated Final Safety Analysis Report (UFSAR). 
This change is also proposed for Section 4.1 of the STS.
    Section 5.2 Containment
    This section does not meet Criteria 1, 2 or 3 in that modification 
of the containment would not create an immediate and significant effect 
on safety. Furthermore, containment integrity is covered in the 
Limiting Conditions for Operation and Surveillance Requirements 
sections of the TS. This Section is not included in the STS. BG&E 
proposes to eliminate it from the Calvert Cliffs Design Features.
    Section 5.3 Reactor Core
    Calvert Cliffs TS Section 5.3 and corresponding STS Specification 
4.2.1 and meets Criteria 1, e.g., state the amount, kind, and source of 
special nuclear material. For consistency with the STS, BG&E proposes 
to adopt the STS titles and wording.
    Section 5.3.3 Control Element Assemblies
    Calvert Cliffs Section 5.3.3, ``Control Element Assemblies,'' and 
the corresponding STS Section 4.2.2 does not meet Criteria 1, 2 or 3. 
The safety significant aspects of control rods, e.g., the reactivity 
worth of control rods and their required insertion times, are included 
in other portions of the TSs. Therefore, control rods do not fall under 
the Criteria 3 as features not described in other sections of the TSs. 
BG&E proposes that this section be eliminated from the Calvert Cliffs 
Design Features and the STS.
    Section 5.4 Reactor Coolant System
    This section does not meet any of the criteria and is not included 
in the STS. It does not meet Criteria 3 in that the requirements on 
degradation, pressure, and temperature are contained in other portions 
of the TS and changes in the total water and steam volume would not 
have an immediate and significant impact on safety. BG&E proposes that 
it be eliminated from the Calvert Cliffs Technical Specifications.
    Section 5.5 Meteorological Tower Location
    This section does not meet any of the criteria and is not included 
in the STS. BG&E proposes to eliminate this section from the Calvert 
Cliffs Design Features section.
    Section 5.6 Fuel Storage
    Calvert Cliffs Section 5.6.1, ``Criticality - Spent Fuel,'' 
specifies the minimum center-to-center distance, a Keff limit, and 
the maximum enrichment for fuel in the spent fuel storage racks. These 
requirements are also contained in STS Section 4.3.1.1, Items a, b, and 
c. Section 5.6.2, ``Criticality - New Fuel,'' specifies the minimum 
center-to-center distance, a keff limit, and a maximum enrichment 
for fuel in the new fuel racks. These requirements are contained in STS 
Section 4.3.1.2, Items a, b, c, and d. These sections meet Criteria 3, 
e.g., geometries which, if altered, would have an immediate and 
significant impact on safety. These requirements do not appear in other 
sections of the TS. The STS language contains the same restrictions as 
the Calvert Cliffs Design Features while introducing no new 
requirements. BG&E proposes adopting the STS language and we will add 
information on uncertainties of the referenced sections of the UFSAR.
    Section 5.6.3 Drainage
    This section in the Calvert Cliffs Design Features section and the 
corresponding STS Section 4.3.2 does not meet any of the criteria. It 
does not meet Criteria 3 in that it does not describe geometry or 
materials of construction and because the requirements are contained in 
another portion of the TS. BG&E proposes that this section be 
eliminated from the Calvert Cliffs Design Features section and from the 
STS.
    Section 5.6.4 Capacity
    Section 5.6.4, ``Capacity,'' states the maximum spent fuel storage 
capacity. This meets Criteria 1 in that it limits the amount of special 
nuclear material that may be stored on site. The Calvert Cliffs TS 
language varies slightly from the STS language in that it makes clear 
that the storage capacity limit applies to the combined storage pool 
for Units 1 and 2. Therefore, we propose to retain the Calvert Cliffs 
TS language.
    Section 5.7 Component Cyclic or Transient Limits
    This section is not included in the STS and does not meet any of 
the criteria. We propose that it be eliminated from the Calvert Cliffs 
Design Features section.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change to the Design Features section adopts 
language from the Design Features section of the Standard Technical 
Specifications (STS) for Combustion Engineering Plants (NUREG-1432, 
September, 1992), based on the Commission's Final Policy Statement 
on Technical Specifications Improvements for Nuclear Power Reactors 
(July 16, 1993). Some requirements in the current Technical 
Specifications have been eliminated or relocated to the UFSAR based 
on the STS as evidence that the NRC no longer considers those 
requirements to meet the criteria for Design Features in 10 CFR 
50.36(c)(4). In some cases, we have proposed elimination of some 
requirements in the STS and the Calvert Cliffs Technical 
Specifications based on a determination that they have no legal or 
regulatory basis.
    We propose eliminating the ``Site'' sections, present in the 
Calvert Cliffs and the STS Design Features sections, which contain 
maps or descriptions of the site boundary and low population zone. 
There are no legal or regulatory requirements for including this 
information in the Technical Specifications and this information 
will be relocated to the UFSAR. We propose adding a ``Site 
Location'' section which contains a description of the Calvert 
Cliffs location as required by the Atomic Energy Act of 1954 [,as 
amended]. The ``Containment,'' ``Reactor Coolant System,'' 
``Meteorological Tower Location,'' and ``Component Cyclic or 
Transients Limits'' sections do not meet the legal or regulatory 
requirements for inclusion in the Design Features section, are not 
included in the STS, and have been deleted. The information in these 
sections is or will be contained in the UFSAR and will be controlled 
under 10 CFR 50.59. The ``Control Element Assemblies'' and ``Fuel 
Storage - Drainage'' sections are included in the STS and Calvert 
Cliffs Technical Specifications but eliminated in this change as 
these sections do not meet any of the legal or regulatory 
requirements for the Design Features section. This information is 
contained in the UFSAR and will be controlled under 10 CFR 50.59. We 
propose adopting the STS wording for the ``Reactor Core'' and ``Fuel 
Storage'' sections with no changes in the present limits or 
controls. Some information in the Calvert Cliffs Technical 
Specifications is not contained in the STS ``Reactor Core'' section. 
This information is contained in the UFSAR and will be controlled 
under 10 CFR 50.59.
    All information eliminated from the Design Features section of 
the Technical Specifications is or will be located in the UFSAR and 
will be controlled under 10 CFR 50.59. The design and operation of 
the plant have not changed. Therefore, the proposed change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Would not create the possibility of a new or different type 
of accident from any accident previously evaluated.
    The proposed change does not represent a change in the 
configuration or operation of the plant. All information eliminated 
from the Technical Specifications will continue to be controlled 
under 10 CFR 50.59. All legal and regulatory requirements for the 
Design Features section continue to be met. Therefore, the proposed 
change does not create the possibility of a new or different type of 
accident from any accident previously evaluated.
    3. Would not involve a significant reduction in a margin of 
safety.
    The proposed change does not represent a change in the 
configuration or operation of the plant. All information eliminated 
from the Technical Specifications will continue to be controlled 
under 10 CFR 50.59. All legal and regulatory requirements for the 
Design Features section continue to be met. 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 
amendments request involves no significant hazards consideration.
    Local Public Document Room location: Calvert County Library, Prince 
Frederick, Maryland 20678.
    Attorney for licensee: Jay E. Silbert, Esquire, Shaw, Pittman, 
Potts and Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 10, 1993
    Description of amendment request: The proposed amendment would 
extend existing plant surveillance intervals to 24 months from 18 
months. This is the third of three submittals and it changes specific 
setpoints to accommodate a 24 month fuel cycle and provides a 
justification for extending the surveillance interval for those 
components and systems that are not related to instrument setpoint 
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:
    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.
    This submittal results in changes to various equipment 
surveillance intervals and related instrument calibration 
frequencies and setpoints.
    The impact of lengthening the current 18 month interval to 24 
months was evaluated and identified no significant system or 
component degradation as a consequence of lengthening the interval 
to 24 months; therefore, systems and components will continue to 
perform their design function. In some cases, the 24 month interval 
required setpoint changes to ensure instrument drift associated with 
the extended interval would not result in exceeding an instrument's 
acceptable setpoint tolerance. In other cases, justification of an 
extended interval was not developed because the surveillance could 
be performed on-line. In these cases, the ``once/cycle'' 
surveillance requirement is changed to the currently allowed 18 
months.
    The proposed changes were developed using the guidance provided 
in Generic Letter 91-04 and Note 1 of Table 4.2.A through 4.2.G of 
Pilgrim's technical specifications. The proposed changes do not 
degrade the performance or increase the challenges to the associated 
safety systems assumed to function in the accident analysis.
    The impact of lengthening the current calibration/functional 
test interval from 1 to 3 months for certain components was also 
evaluated. The evaluation used the guidance in Generic Letter 91-04 
and Note 1 of Table 4.2.A through 4.2.G of Pilgrim's technical 
specifications. No significant system or component degradation was 
identified as a consequence of lengthening the interval to 3 months.
    The proposed changes do not affect the availability of equipment 
or systems required to mitigate the consequences of an accident, and 
do not affect the availability of redundant systems or equipment. 
The plant will continue to operate within the limits specified in 
the Core Operating Limits Report (COLR) and will continue to take 
the same actions if setpoint limits are exceeded.
    Therefore, both the proposed setpoint and non-setpoint changes 
do not significantly 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 analyzed.
    The proposed changes with one exception, do not add or remove 
active components and, therefore, do not introduce failure 
mechanisms of a different type than those previously evaluated. In 
one case, the EDG breaker time delay relays will be replaced with 
more accurate relays to ensure the specified time sequence for 
starting and accepting the emergency load remains within 
specification for the extended cycle. The replacement relays will be 
similar in size, weight, voltage and temperature operating range as 
those being replaced; therefore, the possibility of a new or 
different kind of accident is not created. In addition, the 
surveillance test requirements and the way surveillance tests are 
performed will remain unchanged. Since the intended operation and 
function of the analyzed systems do not change as a result of the 
setpoint and non-setpoint analyses, no new initiators are introduced 
capable of initiating an accident that would render these systems 
unable to provide their required protection. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The operation of Pilgrim Station in accordance with the 
proposed amendment will not involve a significant reduction in the 
margin of safety.
    Although the proposed Technical Specification changes will 
result in an increase in the interval between surveillance tests, 
the existing margins of safety are maintained through our proposed 
setpoint revisions. The proposed setpoint changes either increase 
the plant safety margin or maintain the existing margin and do not 
significantly impact the availability, performance, or intended 
function of the affected systems. In the case of non-setpoint 
changes, evaluation of the affected systems indicates lengthening 
the interval to 24 months does not have significant impact on 
performance. Therefore, the assumptions in Pilgrim's accident 
analyses are not impacted, and the proposed Technical Specification 
changes do 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: Walter R. Butler

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

    Date of amendment request: March 26, 1993
    Description of amendment request: The proposed amendment is a part 
of Commonwealth Edison Company's (CECo's) Technical Specification 
Upgrade Program (TSUP) to improve the quality of the current Technical 
Specifications (TS) for Dresden and Quad Cities. The proposed amendment 
would for both Dresden and Quad Cities, upgrade the requirements of 
Section 3.9/4.9, ``Auxiliary Electrical Systems,'' to include operating 
and shutdown Limiting Condition(s) for Operation (LCO) and Surveillance 
Requirement(s) (SR) that are consistent with the Standard Technical 
Specifications (STS) and later operating plant provisions. Within the 
upgrade to Section 3.9/4.9, Emergency Diesel Generator (EDG) 
reliability provisions are added to implement the recommendations of 
Generic Letter(s) 84-15 and 91-09; Information Notice(s) 84-69 and 91-
62; and Regulatory Guide 1.9, draft Revision 3. Other Generic Letters 
considered in the proposed TS include 83-26, 83-30, and 87-09.
    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:
    Involve a significant increase in the probability or 
consequences of an accident previously evaluated because:
    In general, the proposed changes represent the conversion of 
current requirement to a more generic format, or the addition of 
requirements which are based on the current safety analyses. 
Implementation of these changes will provide increased reliability 
of equipment assumed to operate in the current safety analyses, or 
provide continued assurance that specified parameters remain within 
their acceptance limits, and as such, will not significantly 
increase the probability or consequences of a previously evaluated 
accident.
    Some of the proposed changes represent minor curtailments of the 
current requirements which are based on generic guidance or 
previously approved provisions for other stations. These proposed 
changes are consistent with the current safety analyses and have 
been previously determined to represent sufficient requirements for 
the assurance and reliability of equipment assumed to operate in the 
safety analyses, or provide continued assurance that specified 
parameters remain within their acceptance limits. As such, these 
changes will not significantly increase the probability or 
consequences of a previously evaluated accident.
    A.C. Sources - Operating: The proposed modifications for Section 
3/4.9.A, ``A.C. Sources - Operating'', administratively incorporate 
the requirements of STS, where applicable to Dresden and Quad Cities 
Stations. Most deviations from the STS requirements are based upon 
generic guidance and other approved requirements at other sites. 
Dresden and Quad Cities Station are retaining the current seven day 
allowed outage time from their current specifications for loss of an 
EDG. However, an additional verification of EDG operability has been 
proposed for approximately midway through the seven day AOT. The 
additional verification and additional details in the surveillances 
will significantly improve the overall safety of both Dresden and 
Quad Cities Station.
    Both Dresden and Quad Cities Station's EDG history have shown 
them to be very reliable. This can be demonstrated by the excellent 
pass/fail rate observed during the monthly surveillance tests. As 
stated previously, the addition of several new STS enhancements to 
fuel storage and transfer requirements and other miscellaneous EDG 
surveillances recommended by ASTM codes will improve EDG 
reliability. Therefore, because the EDG's for Dresden and Quad 
Cities Station do not act as accident initiators, the probability of 
an accident previously evaluated for the sites is not increased by 
the incorporation of the proposed requirements.
    Other changes based upon STS guidance are more restrictive and 
limit operation of the site with respect to all A.C. power sources. 
A.C. power sources do not act as initiators of accidents. Therefore, 
the probability of an accident previously evaluated for the sites is 
also not increased by the incorporation of the STS requirements.
    The consequences of any previously evaluated accidents are not 
increased as more restrictions and limitations are added to the 
current versions of both Dresden and Quad Cities specifications. The 
retention of the seven day allowed outage times for the EDG's and 
the offsite power sources does not increase the consequences of any 
previously analyzed accident for both sites as these are the current 
requirements. Therefore, the consequences of any previously 
evaluated accident is not increased as a result of the proposed 
changes.
    A.C. Sources - Shutdown: The proposed modification for Section 
3/4.9.B, ``A.C. Sources - Shutdown'', administratively incorporate 
the requirements of STS where applicable to Dresden and Quad Cities 
Station. The STS requirements add additional provisions not in the 
current Technical Specifications for EDG fuel storage capability 
that will reduce the consequences of a previously analyzed accident. 
The probability of any previously evaluated accident is reduced 
during shutdown by the additional STS restrictions proposed for fuel 
handling type of activities. Therefore, the proposed changes do not 
involve a significant increase in the probability or consequences of 
any previously evaluated accident for Dresden and Quad Cities 
Station.
    D.C. Sources - Operating: The proposed modifications for Section 
3/4.9.C, ``D.C. Sources - Operating'', incorporate the requirements 
of STS where applicable for Dresden and Quad Cities Stations. 
Dresden and Quad Cities are proposing to retain the current 
provisions specified in the current version of Quad Cities Technical 
Specifications that allow the 125 and 250 VDC systems to be out-of-
service for a period up to 72 hours. This 72-hour AOT may be 
extended, when applied to the 125 VDC systems, for up to a maximum 
period of 7 days with both units operating if the alternate 125 VDC 
battery is operable. These changes introduce a difference when 
compared to the STS requirements. However, the STS requirements as 
applied to Dresden and Quad Cities battery systems would prove to be 
overly burdensome requiring dual unit shutdowns to perform most 
maintenance or testing activities. The additional batteries in the 
design of the stations, with their surveillances, charger 
requirements, and breaker verifications, compensate for any STS 
deviations. The proposed requirements are comparable to the existing 
requirements and AOTs for Quad Cities Station; therefore, the 
consequences of any previously evaluated accident are not increased. 
The proposed changes add additional surveillance requirements to the 
D.C. systems at Dresden and Quad Cities to enhance their reliability 
and operational readiness. This also ensures the consequences of any 
previously evaluated accident are not increased. Because the D.C. 
system is not assumed as an accident initiator, the probability of 
any previously evaluated accident is not increased.
    D.C. Sources - Shutdown: The proposed modifications for Section 
3/4.9.D, ``D.C. Sources - Shutdown'', administratively incorporate 
the requirements of STS where applicable to Dresden and Quad Cities 
Station. The proposed changes add additional surveillance 
requirements and more explicitly clarify the LCO's. The additional 
provision for fuel handling type of activities reduces the 
probability of previously evaluated accidents from occurring. The 
additional surveillance activities also improve D.C. reliability and 
thus, reduce the probability of D.C. system unavailability and 
hence, reduce the consequences of previously evaluated accidents. 
Because the D.C. system is not considered as an accident initiator, 
the probability of any previously evaluated accident is not 
increased.
    Distribution - Operating: The proposed modifications for Section 
3/4.9.E, ``Distribution - Operating'', administratively incorporate 
the requirements of STS where applicable to Dresden and Quad Cities 
Station. The proposed changes add additional surveillance 
requirements and LCO's. The proposed requirements/actions for the 
D.C. distribution system are retained to be consistent to the 
proposed AOTs for the D.C. system. The STS requirements as applied 
to the Dresden/Quad Cities D.C. distribution system would prove to 
be overly burdensome, requiring plant shutdowns to perform 
maintenance or testing activities. The additional distribution 
system surveillance and LCO's compensate for the STS deviations. The 
proposed requirements are comparable to the existing requirements 
and AOTs for Quad Cities Station; therefore, the consequences of any 
previously evaluated accident are not increased. The additional 
surveillances and STS-type requirements ensures the reliability and 
operational readiness of the Distribution System and ensures the 
consequences of any previously evaluated accident are not increased. 
Because the Distribution System is not assumed as an accident 
initiator, the probability of any previously evaluated accident is 
not increased.
    Distribution - Shutdown: The proposed modification for Section 
3/4.9.F, ``Distribution - Shutdown'', administratively incorporate 
the requirements of STS where applicable for Dresden and Quad Cities 
Station. The proposed changes add additional requirements that 
ensure the consequences and the probability of any previously 
evaluated accident are not increased.
    RPS Power Monitoring: The proposed modifications for Section 3/
4.9.G, ``RPS Power Monitoring''. administratively incorporate the 
requirements of STS for Dresden and Quad Cities Station. The 
proposed changes add additional requirements for Dresden and clarify 
the existing requirements at Quad Cities. Therefore, the 
consequences and the probability of any previously evaluated 
accident are not increased.
    Create the possibility of a new or different kind of accident 
from any previously evaluated because:
    In general, the proposed changes represent the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analyses. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. These changes do not involve revisions to the design 
of the station. Some of the changes may involve revision in the 
operation of the station; however, these provide additional 
restrictions which are in accordance with the current safety 
analyses, or are to provide for additional testing or surveillances 
which will not introduce new failure mechanisms beyond those already 
considered in the current safety analyses. The retention of the 
current AOTs for EDGs, offsite power sources, and DC systems 
maintain the existing assumptions from the current accident 
analyses; therefore, these changes will not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    The proposed changes for Dresden and Quad Cities Station's 
Technical Specification Section 3/4.9 are based on STS guidelines or 
later operating BWR plants' NRC accepted changes. These proposed 
changes have been reviewed for acceptability at the Dresden and Quad 
Cities Nuclear Power Stations considering similarity of system or 
component design versus the STS of later operating BWRs. No new 
modes of operation are introduced by the proposed changes, 
considering the acceptable operational modes in present 
specifications, the STS, or later operating BWRs. Surveillance 
requirements are changed to reflect improvements in technique, 
frequency of performance or operating experience at later plants. 
Proposed changes to action statements in many places add 
requirements that are not in the present technical specifications or 
adopt requirements that have been used successfully at other 
operating BWRs with designs similar to Dresden and Quad Cities. The 
proposed changes maintain at least the present level of operability. 
Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from any previously evaluated.
    Involve a significant reduction in the margin of safety because:
    In general, the proposed changes represent the conversion of 
current requirements to a more generic format, or the addition of 
requirements which are based on the current safety analyses. Others 
represent minor curtailments of the current requirements which are 
based on generic guidance or previously approved provisions for 
other stations. Some of the later individual items may introduce 
minor reductions in the margin of safety when compared to the 
current requirements. However, other individual changes are the 
adoption of new requirements which will provide significant 
enhancement of the reliability of the equipment assumed to operate 
in the safety analyses, or provide enhanced assurance that specified 
parameters remain with their acceptance limits. These enhancements 
compensate for the individual minor reductions, such that taken 
together, the proposed changes will not significantly reduce the 
margin of safety.
    The proposed changes to the Technical Specification Section 3/
4.9 implement present requirements, or the intent of present 
requirements in accordance with the guidelines set forth in the STS. 
The proposed changes are intended to improve readability, usability, 
and the understanding of technical specification requirements while 
maintaining acceptable levels of safe operation. The proposed 
changes have been evaluated and found to be acceptable for use at 
Dresden and Quad Cities based on system design, safety analyses 
requirements and operational performance. The retention of the 
current AOTs for EDGs, offsite power sources, and DC systems 
maintain the existing assumptions from the current accident 
analyses. Since the proposed changes are based on NRC accepted 
provisions at other operating plants that are applicable at Dresden 
and Quad Cities and maintain necessary levels of system, component 
or parameter readability, the proposed changes do not involve a 
significant reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: for Dresden, the Morris Public 
Library, 604 Liberty Street, Morris, Illinois 60450, and for Quad 
Cities, the Dixon Public Library, 221 Hennepin Avenue, Dixon, Illinois 
61021
    Attorney for licensee: Michael I. Miller, Esquire; Sidley and 
Austin, One First National Plaza, Chicago, Illinois 60690
    NRC Project Director: James E. Dyer

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: November 11, 1993
    Description of amendment request: The proposed amendments would 
consolidate the Quality Verification Department with the Nuclear 
Generation Department and realign the Nuclear Safety Review Board such 
that it reports to the Senior Vice-President of the Nuclear Generation 
Department.
    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 amendments do not involve a significant increase in the 
probability or consequences of an accident previously evaluated.]
    The proposed revisions to consolidate the Quality Verification 
Department with the Nuclear Generation Department and realign the 
NSRB [Nuclear Safety Review Board] such that it reports to the 
Senior Nuclear Officer, change the reference from Semiannual to 
Annual, change the reference from group to division, delete titles 
of persons designated to approve modifications, clarify the 
responsibilities of the Safety Assurance Manager, and delete the 
requirement to perform an annual independent Fire Protection Audit 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated because the changes 
do not have any impact upon the design or operation of any plant 
systems or components.
    [2. The amendments do not create the possibility of a new or 
different kind of accident from any accident previously evaluated.]
    The proposed revisions will not create the possibility of a new 
or different kind of accident from any previously evaluated because 
the changes are administrative in nature and operation of Catawba, 
McGuire, and Oconee Nuclear Stations in accordance with these TS 
[technical specifications] will not create any failure modes not 
bounded by previously evaluated accidents.
    [3. The amendments do not involve a significant reduction in a 
margin of safety.]
    The proposed revisions will not involve a reduction in a margin 
of safety because they are administrative in nature.
    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: Loren R. Plisco, Acting

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

    Date of amendment request: December 6, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to add provisions to allow 
repair of steam generator tubes by the sleeving process.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The proposed change will allow the use of specific steam 
generator tubing sleeves to repair Waterford 3 steam generator tubes 
which exhibit degradation and can be sleeved (in the tube sheet 
crevice area and egg crate supports). The technical specification 
change is proposed to reference the following reports [following NRC 
approval], which qualify the use of steam generator tube sleeves as 
an alternative to tube plugging,
    Combustion Engineering Report CEN-605-P, ``Waterford 3 Steam 
Generator Tube Repair Using Leak Tight Sleeves'', Revision 00-P, 
dated December, 1992.
    Westinghouse Report WCAP-13698, ``Laser Welded Sleeves For 3/4 
Inch Diameter Tube Feedring-Type and Westinghouse Preheater Steam 
Generators'', Revision 1, dated May, 1993.
    Babcock & Wilcox Report 51-1223750-00, ``BWNS Kinetic Sleeve 
Design For CE SGs with 0.048'' Wall Tubes'', Revision 00, dated June 
29, 1993.
    These reports demonstrate that the repair of degraded steam 
generator tubes using tube sleeves will result in tube bundle 
integrity consistent with the original design basis.
    Sleeve design, materials, and joints were designed to the 
applicable ASME Boiler and Pressure Vessel Codes. Extensive analyses 
and testing were performed to demonstrate the adequacy of the tube 
sleeves. The analyses were performed using design and operating 
transient parameters which enveloped loads imposed during normal 
operating, upset and accident conditions. Mechanical testing was 
performed to demonstrate leak resistance and joint strength, 
including fatigue resistance.
    Corrosion testing was also performed to assess the corrosion 
resistance of the sleeve and joint. Based upon the results of the 
analytical and test programs described in detail in the above 
mentioned reports, these tube sleeves meet or exceed all the 
established design and operating criteria.
    Utilization of tube sleeves not only reduces the risk of primary 
to secondary leakage in the steam generator, but can also provide 
for more margin in the safety analysis. Sleeving a tube results in a 
primary flow reduction which has no significant effect on the steam 
generator performance with respect to heat transfer or system flow 
resistance and pressure drop. The cumulative impact of multiple 
sleeved tubes has been evaluated to ensure the effects remain within 
the design bases. The installation of tube sleeves can be 
accomplished within the tube plugging analysis.
    Based on the extensive analysis and test program performed and 
the ability to monitor and remove degraded sleeves from service, 
this change does not significantly increase the probability or 
consequences of an accident previously evaluated.
    A sleeved tube performs the same function in the same passive 
manner as the unsleeved tube. Tube sleeves are designed, qualified, 
and maintained under the stress and pressure limits of ASME Section 
III and Regulatory Guide 1.121. Eddy current testing is performed 
following installation of each sleeve in order to verify the proper 
installation of the tube sleeve and to obtain baseline eddy current 
data. This baseline data is used to monitor any subsequent 
degradation.
    Therefore, the use of tube sleeves does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Steam generator tube integrity is maintained under the same 
limits for sleeved tubes as for unsleeved tubes in accordance with 
ASME Section III and Regulatory Guide 1.1.21. The degradation limit 
at which a tube is considered inoperable remains unchanged and is 
detectable for sleeves as well as tubes. The technical 
specifications continue to require monitoring and restriction of 
primary to secondary system leakage through the steam generators, 
such that there remains reasonable assurance that a significant 
increase in leakage, due to failure of a sleeved (or unsleeved) 
tube, will be detected. The slight reduction in reactor coolant 
system flow, due to sleeving, is considered to have an insignificant 
impact on steam generator operation during normal operation and 
accident conditions and is clearly bounded by tube plugging 
evaluations. The technical specifications will continue to contain 
reporting requirements for tubes which have had their degradation 
spanned regardless of whether the tube is plugged or sleeved.
    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, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 14, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) to revise the Azimuthal Power 
Tilt limit from less than or equal to 0.10 (10%) to less than or equal 
to 0.03 (3%) and to revise the action statement for control element 
assembly (CEA) misalignment to allow 24 hours to restore the tilt to 
less than 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:
    The additional time for recovery from a CEA misalignment is 
acceptable for the following reasons:
    Consistent with the safety analyses, this TS, places a limit on 
tilt for steady state operation as an initial condition for the 
safety analyses. It is not a limit to be applied during transients. 
This is because accident analyses are initiated from steady state 
conditions and are not required to assume a core power distribution 
transient simultaneous with or immediately prior to the accident. 
This would in effect be two accidents occurring simultaneously.
    The probability of having an accident immediately after a CEA 
drop during the 24 hour period allowed for tilt to be restored to 
less than 3% is very low.
    Technical Specification 3/4.1.3 addresses the CEA misalignments 
and requires a 30% power reduction within one hour if the CEA cannot 
be restored to its proper position.
    Reducing power as required in TS 3.2.3 action b.2 will tend to 
increase the azimuthal tilt, making the transient worse. A lower 
power reduces the rate at which xenon near the dropped CEA can be 
burned out. Maintaining power will quicken the process and keep the 
tilt as low as possible.
    The additional time is only allowed for a confirmed CEA 
misalignment which operators have procedures to respond to. The 
change in tilt is expected and is not indicative of anomalous core 
power distribution behavior that might require more immediate 
action.
    This change conservatively reduces the Azimuthal Power Tilt 
technical specification limit to agree with the assumptions used in 
the safety analysis. The lower tilt represents a more even power 
distribution in the core. A CEA drop event may temporarily cause the 
azimuthal power tilt to exceed the 3% limit. However, for the 
reasons identified above and since the probability of having another 
event within the 24 hours allowed for recovery after the CEA drop is 
extremely low, this change does not involve a significant increase 
in the probability or consequences of any accident.
    The change in technical specification limit on tilt does not 
involve any change to any equipment or the manner in which the plant 
will be operated. This change will further restrict unevenness in 
the core power distribution. Therefore, this change does not create 
the possibility of a new or different kind of accident previously 
evaluated.
    The proposed change incorporates an Azimuthal Power Tilt 
technical specification limit to agree with the assumptions used in 
the safety analysis. Implementation of this change will preserve the 
margin of safety and be consistent with the safety analyses. 
Therefore, this change does not involve a significant reduction in 
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, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: December 14, 1993
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TSs) by removing the reactor 
vessel material specimen withdrawal schedule and by updating the 
reactor coolant system pressure-temperature (P-T) curves. The specimen 
withdrawal schedule will be relocated to the Updated Final Safety 
Analysis Report (UFSAR).
    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:
    Although the Reactor Vessel material specimens withdrawal 
schedule will be removed from the Technical Specifications, the 
Technical Specifications bases will continue to provide background 
information on the use of the data obtained from material specimens. 
Also, updates to the schedule will continue to be submitted to the 
NRC for approval prior to implementation.
    Operating the plant in accordance with the new, updated P-T 
Curves will assure preserving the structural integrity of the 
reactor vessel over the life of the plant. The pressure and 
temperature limits were developed in accordance with 10 CFR [Part] 
50 Appendix G requirements.
    Removing the requirements associated with the previous exemption 
to Appendix H (TS 4.4.8.1.2 items a & b) is purely an administrative 
change.
    Therefore, the proposed changes will not significantly increase 
the probability or consequences of any accident previously 
evaluated.
    Removal of the Reactor Vessel material specimen schedule from 
the Technical Specifications has no impact on accidents at the 
plant. Updates to the schedule will still be required to be 
submitted to the NRC prior to implementation per Section II.B.3 of 
Appendix H to 10 CFR Part 50.
    Also, updates to the P-T Curves will not create a new or 
different type [of] accident. The reactor vessel beltline P-T limits 
were revised applying the general guidance of the ASME Code, 
Appendix G procedures with the necessary margins of safety for 
heatup, cooldown and inservice hydro test conditions.
    The change to TS 4.4.8.1.2 items a & b is a purely 
administrative.
    Therefore, the proposed changes will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    Removal of the schedule for Reactor Vessel material specimen 
withdrawal from the Technical Specifications does not impact the 
margin of safety. The schedule will continue to receive NRC review 
and approval prior to implementation of updates to the schedule.
    Updates to the P-T Curves are provided to preserve the margin to 
[sic] safety to assure that when stressed under operating, 
maintenance and testing the boundary behaves in a non-brittle manner 
and the probability of rapidly propagating fracture is minimized.
    The change to TS 4.4.8.1.2 items a & b is a purely 
administrative.
    Therefore, the proposed changes will not result in a significant 
reduction in the margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of New Orleans 
Library, Louisiana Collection, Lakefront, New Orleans, Louisiana 70122
    Attorney for licensee: N.S. Reynolds, Esq., Winston & Strawn 1400 L 
Street N.W., Washington, D.C. 20005-3502
    NRC Project Director: William D. Beckner

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

    Date of amendment request: November 30, 1993
    Description of amendment request: The purpose of the request is to 
change the plant Technical Specifications (TS) by removing the 
protective and maximum allowable setpoint limits for axial power 
imbalance and the trip setpoint for nuclear overpower based on reactor 
coolant system (RCS) flow (flux-to-flow) from the TS and relocating 
them to the existing TMI-1 Core Operating Limits Report (COLR). The 
proposed change is in accordance with Generic Letter 88-16 guidance 
with regard to placing cycle-dependent parameters into the COLR and the 
NRC-approved Babcock and Wilcox Fuel Company (BWFC) Topical Report BAW-
10179P-A, ``Safety Criteria and Methodology for Acceptable Cycle Reload 
Analyses.'' The TMI-1 Cycle 10 COLR, submitted to the NRC on November 
7, 1993, includes these protective and maximum allowable setpoint 
limits and nuclear overpower trip setpoints to support this Technical 
Specifications 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. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability of occurrence or consequences of an accident previously 
evaluated. The proposed amendment relocates protective and maximum 
allowable setpoint limits from Technical Specifications, and design 
nuclear power peaking factors and the maximum allowable local linear 
heat rate limit from Technical Specification Bases, to the TMI-1 
Core Operating Limits Report in accordance with NRC-approved Topical 
Report BAW-10179P-A. The proposed amendment provides continued 
control of the values of these limits and assures these values 
remain consistent with all applicable limits of the safety analyses 
addressed in the TMI-1 FSAR [Final Safety Analysis Report]. The 
Technical Specifications retain the requirement to maintain the 
plant within the appropriate bounds of these limits. Therefore, the 
proposed amendment has no effect on the probability of occurrence or 
consequences of an accident previously evaluated.
    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 accident previously evaluated. The 
proposed amendment relocates protective and maximum allowable 
setpoint limits, design nuclear power peaking factors and maximum 
allowable local linear heat rate limit to the TMI-1 Core Operating 
Limits Report. The Technical Specifications retain the requirement 
to maintain the plant within the appropriate bounds of these limits. 
Therefore, the proposed amendment has no effect on the possibility 
of creating a new or different kind of accident from any accident 
previously evaluated.
    3. Operation of the facility in accordance with the proposed 
amendment would not involve a significant reduction in a margin of 
safety. The proposed amendment provides continued control of the 
values of these limits and assures these values remain consistent 
with all applicable limits of the safety analyses addressed in the 
TMI-1 FSAR. Therefore, it is concluded that operation of the 
facility in accordance with the proposed amendment does not involve 
a significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, 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: John F. Stolz

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

    Date of amendment request: December 14, 1993
    Description of amendment request: The amendment would revise 
Technical Specification (TS) 3/4.8.2, ``DC Sources,'' to delete two 
notes that indicate that two 125-volt full capacity battery chargers 
are required when the Uninterruptible Power Supply is powered by its 
backup DC power supply. These notes apply to the Divisions I and II DC 
sources during operating and shutdown conditions. The licensee has 
determined that only one battery charger is required to meet current 
design requirements and the criteria delineated in Regulatory Guide 
1.32, ``Criteria for Safety-Related Electric Power Systems for Nuclear 
Power Plants.'' The amendment would also revise TS 3/4.8.2 to increase 
the minimum allowable electrolyte temperature for the 125-volt 
batteries from 60 deg.F to 65 deg.F. This proposed change would 
establish consistency between the TSs, the Updated Safety Analysis 
Report, and applicable battery capacity calculations. The amendment 
would also make administrative changes to TS 3/4.8.4, ``Electrical 
Equipment Protective Devices,'' and the TS Bases.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The existing ``notes'' at the bottom of pages 3/4 8-14 and 3/4 
8-19 indicate that two (2) 125-volt full capacity chargers are 
required when the Uninterruptible Power Supply is powered by its 
backup DC power supply. These ``notes'' were based on an overly 
conservative calculation which determined that both chargers were 
required to supply adequate power to connected loads including an 
Uninterruptible Power Supply. More recent calculations indicate that 
one (1) charger is adequate to supply power to connected loads and 
an Uninterruptible Power Supply. Based on these more recent 
calculations, Niagara Mohawk proposes to delete these ``notes.'' 
Because it has been determined adequate power will be provided to 
connected loads with one (1) charger, deletion of these ``notes'' 
will not affect the reliability of connected loads nor their ability 
to perform their intended function. Therefore, this change will not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    Existing Technical Specification 4.8.2.1 requires that the 125-
volt batteries be demonstrated to be operable by verifying that the 
average electrolyte temperature of one out of five connected cells 
is above 60 deg.F. The Nine Mile Point Unit 2 Updated Safety 
Analysis Report and current battery capacity calculations assume a 
battery electrolyte temperature of at least 65 deg.F. The change 
from 60 deg.F to 65 deg.F is conservative in that battery capacity 
is increased. Increasing the capacity of the batteries will not 
adversely affect the reliability of connected loads nor their 
ability to perform their intended function. Therefore, this change 
will not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The changes made to Technical Specification 3/4.8.4 and to the 
Bases of Specifications 3/4.6.3 and 3/4.8.4 are administrative 
changes and do not affect plant systems or operation. Accordingly 
these changes will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    The function of the battery chargers is to provide adequate 
power to connected loads. Since it has been determined that one (1) 
battery charger is sufficient to provide adequate power, deletion of 
these ``notes'' which require that two (2) chargers be available 
does not affect the capability of the chargers to perform their 
function. The proposal to change the required electrolyte 
temperature from 60 deg.F to 65 deg.F increases the capacity of the 
batteries and therefore improves the capability of the batteries to 
perform their intended function. The remaining changes are 
administrative changes and do not affect plant systems or operation.
    The proposed changes do not introduce any new accident 
precursors and do not involve any physical alterations to plant 
configurations which could initiate a new or different kind of 
accident. The changes do not adversely affect the design or 
performance characteristics of the batteries, battery chargers or 
connected loads. The proposed change to increase the required 
electrolyte temperature will increase battery capacity. Therefore, 
the proposed amendment will not create the possibility of a new or 
different kind of accident from any previously evaluated.
    The operation of Nine Mile Point Unit 2, in accordance with the 
proposed amendment, will not involve a significant reduction in a 
margin of safety.
    Niagara Mohawk proposes to delete the existing ``notes'' that 
indicate that two (2) 125-volt full capacity chargers are required 
when the Uninterruptible Power Supply is powered by its backup DC 
power supply. Niagara Mohawk engineering has determined that one (1) 
battery charger is adequate to meet the maximum DC load demands 
including the Uninterruptible Power Supply. The proposed change to 
increase the required electrolyte temperature will increase battery 
capacity. The remaining changes are administrative.
    These changes will not adversely affect the design or 
performance characteristics of the batteries, battery chargers, or 
connected loads nor will they affect the capability of the 
batteries, battery chargers, or connected loads to perform their 
intended function. Therefore, the proposed changes do not involve a 
significant reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Project Director: Robert A. Capra

Pacific Gas and Electric Company, Docket No. 50-133, Humboldt Bay 
Power Plant, Unit 3, Humboldt County, California

    Date of application for amendment: July 7, 1993 (Reference LAR 93-
01)
    Brief description of amendment: The proposed amendment would revise 
the Technical Specifications (TS) for the Humboldt Bay Power Plant Unit 
No. 3. This proposed revision would change TS VII.H.3, ``Semiannual 
Radioactive Effluent Release Report,'' to extend the reporting period 
from semiannually to annually and to change the report submission date 
from 60 days after January 1 and July 1 of each year to before April 1 
of each year.
    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:
    a. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    The proposed TS changes are administrative in nature. The 
proposed changes to TS VII.H.3 are consistent with 10 CFR 50.36a 
report requirements. The proposed changes do not affect accident 
evaluations. The proposed changes are administrative in nature, 
should result in improved administrative practices, and do not 
affect plant operations.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    b. Does the change create the possibility of a new or different 
kind of accident from any accident previously evaluated?
    The proposed changes are administrative in nature, do not result 
in physical alterations or changes to the operation of the plant, 
and cause no change in the method by which any safety-related system 
performs its function.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any accident previously 
evaluated.
    c. Does the change involve a significant reduction in a margin 
of safety?
    These administrative changes do not alter the basic regulatory 
requirements and do not affect any safety analyses.
    The proposed change to TS VII.H.3. does not alter any 
administrative controls over radioactive effluent, nor does the 
proposed change involve any physical alterations to the plant with 
respect to radioactive effluents. Therefore, the proposed change 
would not affect the meaning, application, and function of the TS 
requirements.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment requests involve no significant hazards consideration.
    Local Public Document Room location: Humboldt County Library, 636 F 
Street, Eureka, California 95501
    Attorney for licensee: Christopher J. Warner, Esq., Pacific Gas & 
Electric Company, P.O. Box 7442, San Francisco, California 94120
    NRC Branch Chief: John H. Austin

Portland General Electric Company, et al., Docket No. 50-344, 
Trojan Nuclear Plant, Columbia County, Oregon

    Date of amendment request: January 27, 1993
    Description of amendment request: The proposed amendment, by 
Portland General Electric Company, PGE or the licensee, would change 
the Trojan Nuclear Plant (Trojan) Appendix A Technical Specifications 
to reflect the permanently defueled status of the facility. The 
permanent cessation of power generation at Trojan and the May 5, 1993 
amendment to the license which granted the licensee a Possession Only 
License for the facility has rendered many of the existing provisions 
of the current Appendix A Technical Specifications inappropriate. PGE 
has developed Permanently Defueled Technical Specifications (PDTS) for 
Trojan using NUREG-1431, ``Standard Technical Specifications, 
Westinghouse Plants,'' as a basis for the PDTS scope and format.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided an analysis of the issue of no significant hazards 
consideration based upon the following:
    1. Operation of the facility in accordance with the proposed 
amendment would not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    The proposed amendment shows that the worst case design basis 
accident for this plant, in its permanently shutdown defueled state, 
is a postulated spent fuel handling accident at the Trojan facility. 
The licensee has also identified a second postulated design basis 
accident scenario, a loss of forced cooling to the spent fuel pool. 
Other Trojan Final Safety Analysis Report (FSAR) accident scenarios 
addressed in Chapter 15 are no longer applicable to Trojan in the 
permanently defueled mode. The proposed amendment does not lessen 
any of the requirements associated with handling spent fuel and 
therefore, the probability of a fuel handling accident occurring is 
unchanged. The licensee has analyzed the a loss of forced spent fuel 
pool cooling accident and has shown that this scenario would not 
result in a radiological release. The proposed amendment does not 
change the consequences of the accident since it does not affect the 
magnitude, detection, or mitigation of either accident scenario. 
Additionally, the ability of the spent fuel pool to withstand other 
applicable FSAR events, natural phenomena, and fires is either 
unchanged from the existing licensing basis or is improved during 
the permanently defueled condition.
    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 accident previously evaluated.
    Maintaining the permanently defueled facility in accordance with 
the PDTS does not create the possibility of a new or different kind 
of accident from any previously considered. Most of the existing 
plant systems and functions will not be operational in the 
permanently defueled condition since power operations are prohibited 
and all of the fuel at Trojan is stored in the spent fuel pool. 
However, all structures, systems and components that are necessary 
for safe fuel handling and storage activities will be maintained 
operable during the permanently defueled condition. The proposed 
PDTS provide operation and surveillance requirements and 
administrative controls which are sufficient to ensure that the 
required structures, systems and components will be maintained 
operable in the permanently defueled condition.
    3. Operation of the facility in accordance with the proposed 
amendment does not involve a significant reduction in a margin of 
safety.
    The proposed PDTS are sufficient to ensure no reduction in a 
margin of safety, in part, because of the reduced range of design 
basis accidents against which the facility must be protected now 
that the facility is prohibited from power operations and is 
permanently defueled. Only a fuel handling accident or a loss of 
forced cooling to the spent fuel pool are relevant during the 
permanently defueled condition. The margins of safety for both of 
these accidents will remain the same or improve by maintaining the 
facility in accordance with the proposed PDTS. None of the other 
Chapter 15 FSAR accidents are applicable since power operations are 
prohibited and the facility is permanently defueled. Additionally, 
the margins of safety for other applicable FSAR events, natural 
phenomena, and fires are either unchanged from the existing 
licensing basis or are improved during the permanently defueled 
condition.
    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: Branford Price Millar Library, 
Portland State University, 934 S.W. Harrison Street, P.O. Box 1151, 
Portland, Oregon 97207.
    Attorney for licensees: Leonard A. Girard, Esq., Portland General 
Electric Company, 121 S.W. Salmon Street, Portland, Oregon 97204.
    NRR Project Director: Seymour H. Weiss

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

    Date of amendment request: December 20, 1993
    Description of amendment request: The licensee has requested an 
amendment to the Technical Specifications (TS) to revise Section 3.3.D 
(Weld Channel and Penetration Pressurization System) to allow certain 
portions of the Weld Channel Pressurization System (WCPS) to be 
disconnected if they are determined to be inoperable and not 
practicably accessible for repair. The WCPS continuously pressurizes 
channels over welds in the steel liner of the containment building. To 
be disconnected, an inoperable portion of the WCPS must be covered by 
concrete such that repairs would involve removal of part of the 
containment structure or the inoperable portion of the system is 
located behind plant equipment inside the containment building such 
that repairs would involve relocation of the equipment. The licensee 
has requested this TS amendment since one portion of the WCPS has 
become inoperable and, since it is buried in concrete below the 
containment floor, cannot be practically repaired. In addition, 
administrative changes would be made to TS Section 3.3.B (Containment 
Cooling and Heat Removal) and TS Section 3.3.E (Component Cooling 
System) to correct typographical errors. Specifically, TS Section 
3.3.B.3.b references 3.3.A.3 when is should reference 3.3.B.1 and TS 
Section 3.3.E.3.b references 3.3.A.3 when it should reference 3.3.E.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:
    Consistent with the criteria of 10 CFR 50.92, the enclosed 
application is judged to involve no significant hazards based on the 
following information:
    (1) Does the proposed license amendment involve a significant 
increase in the probability or consequences of an accident 
previously evaluated?
    Response:
    The proposed license amendment does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated. The consequences of an accident previously 
evaluated would not be affected by the disconnection of portions of 
the WCPS [Weld Channel Pressurization System] because the accident 
analyses do not assume the operation of any portion of the WC & PPS 
[Weld Channel & Penetration Pressurization System]. Additionally, 
operation of the WC & PPS is not taken credit for in any offsite 
accident dose calculations. The probability of an accident 
previously evaluated would not be increased because the 
disconnection of any portion of the WCPS could not initiate an 
accident. The administrative changes correct errors in the technical 
specifications and technical specification bases. These 
administrative changes have no affect on the probability or 
consequences of an accident previously evaluated.
    (2) Does the proposed license amendment create the possibility 
of a new or different kind of accident from any accident previously 
evaluated?
    Response:
    The proposed license amendment does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The allowance for the disconnection of sections of the 
WCPS will allow the Authority [Power Authority of the State of New 
York] to avoid repairs that can potentially degrade containment 
integrity or the condition of vital equipment. The proposed license 
amendment does not create the possibility of a new accident because 
the disconnection of any portion of the WCPS could not initiate an 
accident. The administrative changes correct errors in the technical 
specifications and technical specification bases. These 
administrative changes can not [cannot] create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    (3) Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response:
    The proposed amendment would not involve a significant reduction 
in a margin of safety. The allowance for the disconnection of 
sections of the WCPS will allow the Authority to avoid repairs that 
can potentially degrade containment integrity or the condition of 
vital equipment. The WC & PPS will still provide continuous 
pressurization and monitoring of leak-tightness for the zones 
incorporated into the containment penetrations and for at lease 80% 
of the channels over the welds in the steel inner of the containment 
building. The WC & PPS will continue to provide assurance that the 
containment leak rate in the event of an accident is lower than that 
assumed in the accident analyses because the accident analyses do 
not assume that any section of the WC & PPS is operating. The 
administrative changes correct errors in the technical 
specifications and technical specification bases. These 
administrative changes have no affect on any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10601.
    Attorney for licensee: Mr. Charles M. Pratt, 10 Columbus Circle, 
New York, New York 10019.
    NRC Project Director: Robert A. Capra

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

    Date of amendment request: December 8, 1993. The December 8, 1993 
request supersedes an earlier request dated November 17, 1992, which 
was previously noticed (58 FR 52994). This notice supersedes the 
previous notice.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications for Salem Units 1 and 2 to 
incorporate the guidance provided by the staff in Generic Letter 90-06 
(GL 90-06) as follows:
    1. Specification 3/4.4.3 and 3/4.4.5 for Salem 1 and Salem 2, 
respectively, ``RELIEF VALVES'', will incorporate the guidance of GL 
90-06 with the following exceptions:
    a. The surveillance requirement to test the emergency power supply 
for the power operated relief valves (PORVs) and block valves has not 
been incorporated. The PORVs and block valves are powered from the 
emergency busses.
    b. The entry conditions for one or both PORVs inoperable will not 
be based on excessive seat leakage alone. Entry conditions will be 
based on the capability of the PORV to be manually cycled consistent 
with the Action Statements contained in NUREG-1431, Standard Technical 
Specifications for Westinghouse Plants.
    c. With both PORVs inoperable in Modes 1, 2, or 3 and not capable 
of being manually cycled, or both block valves are inoperable, an 
allowed outage time of 6 hours to restore on block valve or PORV to 
operable status has been requested.
    2. Specification 3.4.9.3 and 3.4.10.3 for Salem 1 and Salem 2, 
respectively, ``OVERPRESSURE PROTECTION SYSTEMS'' will incorporate the 
guidance of GL 90-06.
    3. In addition to the guidance provided by GL 90-06, the following 
changes have also been proposed.
    a. For Salem 1, the reference to the specific American Society of 
Mechanical Engineers (ASME) valve category would be deleted from 
Specification 3.4.9.3.1.
    b. Specification 3.5.3. ``ECCS SUBSYSTEMS - TAVE < 350 deg.F'' 
would be revised to clarify the applicability of the surveillance 
requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. Will not involve a significant increase in the probability or 
consequences of an accident or malfunction of equipment important to 
safety previously evaluated. This change will instill administrative 
restrictions on the Low Temperature Overpressure Protection System 
and the PORVs thereby improving reliability and availability to 
respond to a Steam Generator Tube Rupture and overpressure 
transient. The proposed amendment requires that power be maintained 
to block valves that are closed should a PORV be inoperable, but 
still capable of being manually cycled. This change ensures that the 
block valves can be opened on demand from the control room. Power is 
maintained to the block valves so that it is operable and may be 
subsequently opened to allow the PORV to be used to control reactor 
pressure. The capability to manually cycle the PORV is consistent 
with the Action Statement contained in NUREG-1431. This change 
actually improves overall plant safety. Therefore, the proposed 
amendment does not involve a procedural or physical change to any 
structure, system or component that significantly affects accident/
malfunction probabilities or consequences previously evaluated in 
the UFSAR.
    2. Will not create the possibility of a new or different kind of 
accident from any accident previously evaluated. The proposed 
amendment does not involve any physical changes to plant structures, 
components, or systems. With the exception of maintaining power to a 
block valve closed to isolate a PORV that is inoperable but capable 
of being manually cycled, which does not create the possibility of a 
new or different kind of accident, the proposed change will not 
impose any different requirements on plant operation. Therefore, the 
proposed changes do not create the possibility of a new or different 
accident from any previously evaluated.
    3. Will not involve a significant reduction in a margin of 
safety. The proposed changes actually increase the overall margin of 
safety by improving the availability and reliability of the PORVs 
and Block valves in response to Steam Generator Tube Rupture events, 
and the PORVs in response to overpressure transients.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Charles L. Miller

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

    Date of amendment request: January 25, 1993
    Description of amendment request: The amendment revises the 
pressure-temperature limit curves for heatup, cooldown, hydrostatic 
tests and criticality from 10 effective full power years (EFPY) to 15 
EFPY.
    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 Technical Specification changes:
    1. Do not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed changes assure that existing safety limits are not 
exceeded due to changing Reactor Vessel conditions. These changes 
reflect the latest material testing results per 10CFR50, Appendix H. 
Clearly defining the pressure and temperature limits decreases the 
probability of nonductile failures.
    Therefore, it may be concluded that the proposed changes do not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    2. Do not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    No physical plant modifications or new operating configurations 
result from these changes. These changes do not adversely affect the 
design or operation of any system or component important to safety, 
rather, they establish limits to assure that operation remains 
within acceptable boundaries.
    Therefore, no new or different accident from any previously 
evaluated will be created.
    3. Do not involve a significant reduction in a margin of safety.
    Capsule ``X'' analysis concluded that the Reactor Vessel has 
sufficient fracture toughness for continued safe operation, 
providing operation remains within acceptable limits. The heatup and 
cooldown curves define these acceptable limits. The proposed changes 
maintain existing margins of safety by modifying operating 
requirements based on specimen Capsule ``X'' analysis. This analysis 
accounts for the actual chemistry data associated with the limiting 
weld, rather than the previously used default values, thus providing 
some additional operating margin with no reduction in the margin of 
safety. Actual weld chemistry data is specified in existing 
Technical Specification Table B 3/4.4-1.
    Therefore, it may be concluded that 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: Salem Free Public library, 112 
West Broadway, Salem, New Jersey 08079
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
    NRC Project Director: Charles L. Miller

Southern Nuclear Operating Company, Inc., Docket No. 50-348, Joseph 
M. Farley Nuclear Plant, Unit 1, Houston County, Alabama

    Date of amendment request: December 9, 1993
    Description of amendment request: The proposed amendment would 
modify Technical Specification (TS) 3/4.4.6, Steam Generators, and TS 
3/4.4.9, Specific Activity, and their associated bases. The steam 
generator plugging/repair limit will be modified to establish a 
methodology for determining serviceability for tubes with outside 
diameter stress corrosion cracking at the tube support plate that more 
realistically assesses structural integrity. For Unit 1, the 
operational leakage requirement will be modified to reduce the total 
allowable primary-to-secondary leakage for any one steam generator from 
500 gallons per day to 140 gallons per day. In addition, the TS limit 
for specific activity of dose equivalent I131 and its transient 
dose equivalent I131 reactor coolant specific activity will be 
reduced by a factor of 4 in order to increase the allowable leakage in 
the event of a steam line break.
    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 Farley Unit 1 in accordance with the proposed 
license amendment does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Testing of model boiler specimens for free standing tubes at 
room temperature conditions show burst pressures as high as 
approximately 5000 psi [per square inch] for indications of outer 
diameter stress corrosion cracking with voltage measurements as high 
as 26.5 volts. Burst testing performed on pulled tubes with up to 
7.5 volt indications show burst pressures in excess of 5900 psi at 
room temperature. As stated earlier, tube burst criteria are 
inherently satisfied during normal operating conditions by the 
presence of the tube support plate. Furthermore, correcting for the 
effects of temperature on material properties and minimum strength 
levels (as the burst testing was done at room temperature), tube 
burst capability significantly exceeds the R.G. [Regulatory Guide] 
1.121 criterion requiring the maintenance of a margin of 1.43 times 
the steam line break pressure differential on tube burst if through-
wall cracks are present without regard to the presence of the tube 
support plate. Based on the existing data base this criterion is 
satisfied with bobbin coil indications with signal amplitudes over 
twice the 2.0 volt interim repair criteria, regardless of the 
indicated depth measurement. This structural limit is based on a 
lower 95 [percent] confidence level limit of the data. The 2.0 volt 
criteria provides an extremely conservative margin of safety to the 
structural limit considering expected growth rates of outside 
diameter stress corrosion cracking at Farley. Alternate crack 
morphologies can correspond to a voltage so that a unique crack 
length is not defined by a burst pressure to voltage correlation. 
However, relative to expected leakage during normal operating 
conditions, no field leakage has been reported from tubes with 
indications with a voltage level of under 7.7 volts for a 3/4 inch 
tube with a 10 volt correlation to 7/8 inch tubing (as compared to 
the 2.0 volt proposed interim tube repair limit). Thus, the proposed 
amendment does not involve a significant increase in the probability 
or consequences of an accident.
    Relative to the expected leakage during accident condition 
loadings, the accidents that are affected by primary-to-secondary 
leakage and steam release to the environment are Loss of External 
Electrical Load and/or Turbine Trip, Loss of All AC Power to Station 
Auxiliaries, Major Secondary System Pipe Failure, Steam Generator 
Tube Rupture, Reactor Coolant Pump Locked Rotor, and Rupture of a 
Control Rod Drive Mechanism Housing. Of these, the Major Secondary 
System Pipe Failure is the most limiting for Farley in considering 
the potential for off-site doses. The offsite dose analyses for the 
other events which model primary-to secondary leakage and steam 
release from the secondary side to the environment assume that the 
secondary side remains intact. The steam generator tubes are not 
subjected to a sustained increase in differential pressure, as is 
the case following a steam line break event. This increase in 
differential pressure is responsible for the postulated increase in 
leakage and associated offsite doses following a steam line break 
event. In addition, the steam line break event results in a bypass 
of containment for steam generator leakage. Upon implementation of 
the interim repair criteria, it must be verified that the expected 
distribution of cracking indications at the tube support plate 
intersections are such that primary-to-secondary leakage would 
result in site boundary dose within the current licensing basis. 
Data indicate that a threshold voltage of 2.8 volts would result in 
through-wall cracks long enough to leak at steam line break 
conditions. Application of the proposed repair criteria requires 
that the current distribution of a number of indications versus 
voltage be obtained during the refueling outages. The current 
voltage is then combined with the rate of change in voltage 
measurement and a voltage measurement uncertainty to establish an 
end of cycle voltage distribution and, thus, leak rate during steam 
line break pressure differential. The leak rate during a steam line 
break is further increased by a factor related to the probability of 
detection of the flaws. If it is found that the potential steam line 
break leakage for degraded intersections planned to be left in 
service coupled with the reduced specific activity levels allowed 
result in radiological consequences outside the current licensing 
basis, then additional tubes will be plugged or repaired to reduce 
steam line break leakage potential to within the acceptance limit. 
Thus, the consequences of the most limiting design basis accident 
are constrained to present licensing basis limits.
    2) The proposed license amendment does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    Implementation of the proposed interim tube support plate 
elevation steam generator tube repair criteria does not introduce 
any significant changes to the plant design basis. Use of the 
criteria does not provide a mechanism which could result in an 
accident outside of the region of the tube support plate elevations. 
Neither a single or multiple tube rupture event would be expected in 
a steam generator in which the repair criteria has been applied 
(during all plant conditions). The bobbin probe signal amplitude 
repair criteria is established such that operational leakage of 
excessive leakage during a postulated steam line break condition is 
not anticipated. Southern Nuclear has previously implemented a 
maximum leakage rate limit of 140 gpd [gallons per day] per steam 
generator on Unit 1. The R.G. 1.121 criterion for establishing 
operational leakage rate limits that require plant shutdown are 
based upon leak-before-break considerations to detect a free span 
crack before potential tube rupture. The 140 gpd limit provides for 
leakage detection and plant shutdown in the event of the occurrence 
of an unexpected single crack resulting in leakage that is 
associated with the longest permissible crack length. R.G. 1.121 
acceptance criteria for establishing operating leakage limits are 
based on leak-before-break considerations such that plant shutdown 
is initiated if the leakage associated with the longest permissible 
crack is exceeded. The longest permissible crack is the length that 
provides a factor of safety of 1.43 against bursting at steam line 
break pressure differential. A voltage amplitude approximately 9 
volts for typical outside diameter stress corrosion cracking 
corresponds to meeting this tube burst requirement at the 95 
[percent] prediction interval on the burst correlation. Alternate 
crack morphologies can correspond to a voltage so that a unique 
crack length is not defined by the burst pressure versus voltage 
correlation. Consequently, typical burst pressure versus through-
wall crack length correlations are used below to define the 
``longest permissible crack'' for evaluating operating leakage 
limits.
    The single through-wall crack lengths that result in tube burst 
at 1.43 times steam line break pressure differential and steam line 
break conditions are about 0.53 inch and 0.84 inch, respectively. 
Normal leakage for these crack lengths would range from about 0.4 
gallons per minute to 4.5 gallons per minute respectively while 
lower 95 [percent] confidence level leak rates would range from 
about 0.06 gallons per minute to 0.6 gallons per minute, 
respectively.
    An operating leak rate of 140 gpd per steam generator has been 
implemented on Unit 1. This leakage limit provides for detection of 
0.4 inch long cracks at nominal leak rates and 0.6 inch long cracks 
at the lower 95 [percent] confidence level and for three times 
normal operating pressure differential at less than nominal leak 
rates.
    Based on the above, the implementation of interim plugging 
criteria will not create the possibility of a new or different kind 
of accident from any previously evaluated.
    3) The proposed license amendment does not involve a significant 
reduction in margin of safety.
    The use of the interim tube support plate elevation repair 
criteria is demonstrated to maintain steam generator tube integrity 
commensurate with the requirements of R.G. 1.121. R.G. 1.121 
describes a method acceptable to the NRC staff for meeting [General 
Design Criteria] 2, 14, 15, 31, and 32 by reducing the probability 
of the consequences of steam generator tube rupture. This is 
accomplished by determining the limiting conditions of degradation 
of steam generator tubing, as established by inservice inspection, 
for which tubes with unacceptable cracking should be removed from 
service. Upon implementation of the criteria, even under the worst 
case conditions, the occurrence of outside diameter stress corrosion 
cracking at the tube support plate elevations is not expected to 
lead to a steam generator tube rupture event during normal or 
faulted plant conditions. The most limiting effect would be a 
possible increase in leakage during a steam line break event. 
Excessive leakage during a steam line break event, however, is 
precluded by verifying that, once the criteria are applied, the 
expected end of cycle distribution of crack indications at the tube 
support plate elevations would result in minimal, and acceptable 
primary to secondary leakage during the event and hence help to 
demonstrate radiological conditions are less than an appropriate 
fraction of the 10 CFR [Part] 100 guideline.
    The margin to burst for the tubes using the interim repair 
criteria is comparable to that currently provided by existing 
technical specifications.
    In addressing the combined effects of LOCA [loss-of-coolant 
accident] + SSE [safe shutdown earthquake] on the steam generator 
component (as required by GDC 2), it has been determined that tube 
collapse may occur in the steam generators at some plants. This is 
the case as the tube support plates may become deformed as a result 
of lateral loads at the wedge supports at the periphery of the plate 
due to either the LOCA rarefaction wave and/or SSE loadings. Then 
the resulting pressure differential on the deformed tubes may cause 
some of the tubes to collapse.
    There are two issues associated with steam generator tube 
collapse. First, the collapse of steam generator tubing reduces the 
RCS [reactor coolant system] flow area through the tubes. The 
reduction in flow area increases the resistance to flow of steam 
from the core during a LOCA which, in turn, may potentially increase 
Peak Clad Temperature (PCT). Second, there is a potential the 
partial through-wall cracks in tubes could progress to through-wall 
cracks during tube deformation or collapse or that short through-
wall indications would leak at significantly higher leak rates than 
included in the leak rate assessments.
    Consequently, a detailed leak-before-break analysis was 
performed and it was concluded that the leak-before-break 
methodology (as permitted by GDC 4) is applicable to the Farley Unit 
1 [RCS] primary loops and, thus, the probability of breaks in the 
primary loop piping is sufficiently low that they need not be 
considered in the structural design basis of the plant. Excluding 
breaks in RCS primary loops, the LOCA loads from the large branch 
line breaks were analyzed at Farley Unit 1 and were found to be of 
insufficient magnitude to result in steam generator tube collapse or 
significant deformation.
    Regardless of whether or not leak-before-break is applied to the 
primary loop piping at Farley, any flow area reduction is expected 
to be minimal (much less than 1 [percent]) and PCT margin is 
available to account for this potential effect. Based on analyses 
results, no tubes near wedge locations are expected to collapse or 
deform to the degree that secondary to primary in-leakage would be 
increased over current expected levels. For all other steam 
generator tubes, the possibility of secondary-to-primary leakage in 
the event of a LOCA + SSE event is not significant. In actuality, 
the amount of secondary-to-primary leakage in the event of a LOCA + 
SSE is expected to be less than that currently allowed, i.e., 500 
gpd per steam generator. Furthermore, secondary-to-primary in-
leakage for the same pressure differential since the cracks would 
tend to tighten under a secondary-to-primary pressure differential. 
Also the presence of the tube support plate is expected to reduce 
the amount of in-leakage.
    Addressing the R.G. 1.83 considerations, implementation of the 
tube repair criteria is supplemented by 100 [percent] inspection 
requirements at the tube support plate elevations having outside 
diameter stress corrosion cracking indications, reduced operating 
leak rate limits, eddy current inspection guidelines to provide 
consistency in voltage normalization, and rotating pancake coil 
inspection requirements for the larger indications left in service 
to characterize the principal degradation mechanism as outside 
diameter stress corrosion cracking.
    As noted previously, implementation of the tube support plate 
elevation repair criteria will decrease the number of tubes which 
must be taken out of service with tube plugs or repaired. The 
installation of steam generator tube plugs or tube sleeves would 
reduce the RCS flow margin, thus implementation of the interim 
repair criteria will maintain the margin of flow that would 
otherwise be reduced through increased tube plugging or sleeving.
    Based on the above, it is concluded that the proposed change 
does not result in a significant reduction in margin with respect to 
plant safety as defined in the Final Safety Analysis Report or any 
bases of the plant Technical Specifications.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: Houston-Love Memorial Library, 
212 W. Burdeshaw Street, Post Office Box 1369, Dothan, Alabama 36302
    Attorney for licensee: James H. Miller, III, Esq., Balch and 
Bingham, Post Office Box 306, 1710 Sixth Avenue North, Birmingham, 
Alabama 35201
    NRC Project Director: S. Singh Bajwa

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

    Date of amendment request: December 10, 1993
    Description of amendment request: The proposed changes would modify 
the surveillance frequency of the auxiliary feedwater system pumps and 
valves from monthly to quarterly. Various administrative changes are 
being proposed such as 1) punctuation and grammar, 2) correction of 
system or component names, and 3) capitalization of defined words.
    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 Surry Power Station in accordance 
with the proposed Technical Specifications changes will not:
    1. Involve a significant increase in the probability of 
occurrence or consequences of an accident previously evaluated.
    Changing the surveillance test frequency of the Auxiliary 
Feedwater System pumps and valves does not significantly affect the 
probability of occurrence or consequences of any previously 
evaluated accidents. The probability of an accident occurrence is 
not increased in itself by the proposed changes in surveillance 
testing of the Auxiliary Feedwater System pumps and valves. 
[A]uxiliary feedwater pump testing is performed through a full-flow 
test line, thereby not affecting normal plant operations. Changes to 
the testing therefore do not affect the probability of an accident 
occurrence. Redundant trains of the Auxiliary Feedwater System 
remain available during surveillance testing, therefore, the 
consequences of an accident are unchanged by the proposed changes in 
the Auxiliary Feedwater System surveillance test frequencies. 
Quarterly testing of the pumps and valves will continue to assure 
that the Auxiliary Feedwater System is capable of performing its 
intended functions for either unit if called upon. Consistent with 
Generic Letter 93-05, ``Line-Item Technical Specifications 
Improvements to Reduce Surveillance Requirements for Testing During 
Power Operation,'' the new testing frequency should reduce Auxiliary 
Feedwater System unavailability resulting from failures and 
equipment degradation during testing, thereby resulting in improved 
system reliability. Invoking ASME Section XI as the acceptance 
criteria for testing the Auxiliary Feedwater Pumps is an enhancement 
to the acceptance criteria presently specified. Furthermore, the 
operability requirements for the Auxiliary Feedwater System remain 
unchanged. Therefore, the probability or consequences of any 
previously analyzed accident are not increased by the proposed 
changes in surveillance requirements for the Auxiliary Feedwater 
System.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Changes in test frequency and acceptance criteria for the 
Auxiliary Feedwater System pumps and valves do not involve any 
physical modification of the plant or result in a change in a method 
of operation. Quarterly testing of the pumps and valves during both 
operation and shutdown will continue to assure that the Auxiliary 
Feedwater System will be capable of performing its intended function 
for either unit. Invoking ASME Section XI acceptance criteria for 
testing the Auxiliary Feedwater System pumps is an enhancement to 
the acceptance criteria presently specified. The operability 
requirements for the Auxiliary Feedwater System remain unchanged. 
Furthermore, new or different failure modes are not introduced by 
these changes in surveillance requirements. Therefore, a new or 
different type of accident is not created by these proposed changes 
in surveillance requirements for the Auxiliary Feedwater System.
    3. Involve a significant reduction in a margin of safety.
    Changing the surveillance requirements of the Auxiliary 
Feedwater System pumps and valves does not affect any safety limits 
or limiting safety system settings. System operating parameters are 
unaffected. This reduced pump and valve testing frequency should 
reduce Auxiliary Feedwater System unavailability due to actual 
testing, as well as failures and equipment degradation during 
testing. Thus reduced testing results in an improved system 
reliability. Quarterly testing of the pumps and valves during 
operation and shutdown will continue to assure that the Auxiliary 
Feedwater System will be capable of performing its intended 
functions for either unit. Therefore, the reduction in surveillance 
testing requirements for the Auxiliary Feedwater System pumps and 
valves does not reduce any margin of safety.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 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: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185.
    Attorney for licensee: Michael W. Maupin, Esq., Hunton and 
Williams, Riverfront Plaza, East Tower, 951 E. Byrd Street, Richmond, 
Virginia 23219.
    NRC Project Director: Herbert N. Berkow

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

    Date of amendment request: September 17, 1993
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications 
(TS) by incorporating technical and administrative changes to TS 4.5, 
Emergency Core Cooling System and Containment Air Cooling System Tests; 
TS 4.7, Main Steam Isolation Valves; and Table TS 4.1-3, Minimum 
Frequencies for Equipment Tests. Changes are proposed for the safety 
injection (SI) system automatic initiation test; the internal 
containment spray system (ICS) flow blockage test; the SI, ICS and 
residual heat removal pumps' periodic tests; the main steam isolation 
valves' test; and the periodic control rod functional test.
    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:
    (a) TS 4.5.a.1.A
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability of an accident previously evaluated is not 
increased by the TS change. The changes do not affect any structure, 
system, or component that initiates an accident analyzed in the 
Updated Safety Analysis Report (USAR). The probability of an 
accident occurring is independent of the availability of emergency 
core cooling components used to mitigate an accident.
    The consequences of an accident previously evaluated will not be 
increased by this TS change. Revising the TS wording to clarify that 
the pumps may be operated during the periodic surveillance tests 
does not decrease their availability and therefore does not decrease 
their ability to mitigate the consequences of accidents previously 
evaluated in the USAR.
    Clarifying the TS wording will not increase 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.
    A new or different kind of accident from those previously 
evaluated in the USAR will not be created by this TS change.
    The automatic SI actuation is designed to respond to various 
events analyzed in the USAR which take credit for SI in the event 
mitigation. The test required by TS 4.5.a.1.A verifies that the 
valves, pump circuit breakers, and automatic circuitry receive the 
SI signal in the proper sequence. The procedural prerequisites for 
the performance of this test ensure that an adequate flow path and 
overpressure protection are available for the pumps during the test.
    This proposed amendment does not alter the plant configuration, 
operating setpoints, or overall plant performance. It simply 
provides clarification that the pumps may start and operate in 
conjunction with the automatic circuitry test; however, the pumps 
are not required to start and operate for this test. Therefore, the 
possibility or a new or different accident from any previously 
evaluated is not created.
    3) involve a significant reduction in the margin of safety.
    The intent of the TS, i.e., verify each component receives the 
SI signal in the proper sequence, is not altered by this proposed 
change. Clarifying that the pumps are allowed to operate during the 
performance of this surveillance requirement does not adversely 
affect the ability of the surveillance to demonstrate system 
actuation. Therefore, this proposed change will not reduce the 
margin of safety.
    (b) TS 4.5.a.2.B
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This proposed change does not affect the probability of an 
accident previously evaluated. The ICS system is designed to respond 
to various events analyzed in the USAR which take credit for spray 
for event mitigation. The probability of an accident occurring is 
independent of the availability of containment air cooling systems 
used to mitigate an accident.
    This proposed change does not affect the consequences of an 
accident previously evaluated. The proposed change allows the use of 
surveillance methods that provide assurance of system capability to 
perform required design functions in the event of an accident. The 
configuration of containment air cooling components used to mitigate 
the consequences of accidents previously evaluated is not being 
changed.
    Therefore, the proposed changes do not increase the probability 
or consequences of accidents previously evaluated.
    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    A new or different kind of accident from those previously 
evaluated will not be created by this TS change. The proposed 
changes do not alter the operation, function or modes of plant or 
equipment operation. Allowing the use of equivalent surveillance 
methods does not create the probability of a new or different kind 
of accident.
    3) involve a significant reduction in the margin of safety.
    This proposed change does not alter the intent of the 
specification. The use of equivalent test methods to demonstrate 
component functionality does not reduce the margin of safety. The 
proposed change does remove a requirement to verify 100% (i.e. 168) 
of the installed nozzles are unobstructed; however, as noted in the 
safety evaluation, the original system was overdesigned by 
approximately 5%. Verification of the number of required nozzles to 
satisfy design assumptions is assured by the design requirements 
stated in the USAR. Since the minimum design requirements stated in 
the USAR are demonstrated by surveillance, the margin of safety is 
not reduced. Consistent with Section 3.6.6A and associated basis 
statements of NUREG 1431, ``Westinghouse Standard Technical 
Specifications,'' Revision 0, performing surveillance at ten year 
intervals assures the required number of spray nozzles are capable 
of functioning in the event of an accident. Therefore, there is no 
adverse effect on public health and safety.
    (c) TS 4.5.b.1.B
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change will not increase the probability of 
accidents previously evaluated. The probability of an accident 
occurring is independent of the availability of the systems used to 
mitigate an accident.
    The proposed change will not increase the consequences of an 
accident previously evaluated. The consequences of previously 
evaluated accidents requiring SI, RHR, and containment spray for 
event mitigation are independent of the surveillance testing of 
these systems.
    The SI, RHR, and ICS pumps are designed to operate in response 
to various events analyzed in the USAR which initiate on a SI or 
high-high containment pressure signal. The proposed change is 
intended to allow the use of full flow quarterly testing for the ICS 
pumps and retain the existing quarterly mini-flow requirements for 
the SI and RHR pumps. Surveillance testing verifies proper operation 
of the components and thereby ensures the consequences of the 
accident are consistent with the evaluations of the USAR.
    Therefore, the proposed change will not result in 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.
    A new or different kind of accident from those previously 
evaluated will not be created by this TS change. The proposed 
amendment does not alter the plant configuration, operating 
setpoints, or overall plant performance.
    3) involve a significant reduction in the margin of safety.
    The margin of safety will not be reduced by verifying the pump's 
performance through full flow testing or miniflow testing. The 
intent of the TS, i.e. verifying no degradation of the ICS pump is 
satisfied by performing a full flow test on a quarterly basis.
    (d) Administrative changes to TS Section 4.5
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated, 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 the margin of safety.
    The proposed changes are administrative in nature and do not 
alter the intent or interpretation of the TS. Therefore, no 
significant hazards exist.
    (e) TS Section 4.7
    The proposed changes were reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The probability of an accident previously evaluated will not be 
increased by this TS change. The MSIVs are designed to limit the 
cooldown rate of the RCS and prevent structural damage to the 
containment, resulting from a SLB incident, by closing to isolate 
the SGs. The probability of a SLB occurring is independent of MSIV 
operability.
    The consequences of an accident previously evaluated will not be 
increased by this TS change. This TS change allows testing of the 
MSIVs when operating temperature and pressure conditions exist that 
are consistent with the conditions under which the acceptance 
criteria were generated. The acceptable closure time for the MSIVs 
is not being modified by this TS change, therefore the consequences 
of accidents relying on the closure of the MSIVs are not increased.
    An evaluation of the Updated Safety Analysis Report (USAR) SLB 
analyses was performed. The existing core power and reactor coolant 
system transient analyses assumed initial hot shutdown conditions 
for all cases analyzed since this represents the most conservative 
initial conditions for the accident. The analysis of the containment 
pressure transient also assumed initial hot shutdown conditions (at 
which time the steam pressure is highest and there is the greatest 
inventory of water in the steam generator). The containment pressure 
analysis also conservatively delayed the MSIV closures such that 
steam flow from both steam generators existed for the first ten 
seconds. It has been determined that the proposed change in mode for 
surveillance testing is enveloped by the existing analyses, 
therefore the consequences of an accident remained unchanged.
    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    A new or different kind of accident from those previously 
evaluated will not be created by this TS change. The proposed 
amendment does not alter the plant configuration, operating 
setpoints or overall plant performance.
    3) involve a significant reduction in the margin of safety.
    The margin of safety will not be reduced by this TS change. 
Changing the plant conditions under which this surveillance is 
performed does not alter the acceptance criterion for closure time. 
Changing the operating mode for the surveillance test is based on 
engineering judgement which recognizes the importance of 
demonstrating the capability of the components to perform in a steam 
environment.
    (f) Table TS 4.1-3
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    1) involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    This proposed change will not alter the intent of this 
specification which is to ensure that the control rods are operable 
and capable of performing their safety-related function during a 
USAR analyzed event. Control rods fully inserted into the core are 
already performing their safety-related function and therefore, by 
definition, are operable. Revising the specification clarifies that 
the control rods fully inserted into the core need not be tested. 
Therefore, the probability or consequences of an accident previously 
evaluated will not be increased.
    2) create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The proposed amendment does not alter the plant configuration, 
operating setpoints, or overall plant performance. Therefore, a new 
or different kind of accident from those previously evaluated will 
not be created by this TS change.
    3) involve a significant reduction in the margin of safety.
    Rods which are fully inserted into the core are already 
performing their safety function and therefore, by definition, are 
already operable. Exempting these control rods from a partial 
movement test will not reduce the margin of safety.
    (g) Administrative changes to Table TS    4.1-3
    The proposed change was reviewed in accordance with the 
provisions of 10 CFR 50.92 to show no significant hazards exist. The 
proposed change will not:
    (1) involve a significant increase in the probability or 
consequences of an accident previously evaluated, 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 the margin of safety.
    The proposed changes are administrative in nature and do not 
alter the intent or interpretation of the TS. Therefore, no 
significant hazards exist.
    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Local Public Document Room location: University of Wisconsin 
Library Learning Center, 2420 Nicolet Drive, Green Bay, Wisconsin 
54301.
    Attorney for licensee: Bradley D. Jackson, Esq., Foley and Lardner, 
P. O. Box 1497, Madison, Wisconsin 53701-1497.
    NRC Project Director: John N. Hannon

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

    Date of amendment request: November 23, 1993
    Description of amendment request: The proposed amendments would 
change the operating conditions and limiting conditions for operation 
for containment systems, and revise corresponding definitions and 
tests. In addition, related bases would be updated.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed amendments will not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    The proposed changes will add operating conditions and limiting 
conditions for operation to Section 15.3.6, ``Containment System,'' 
of the Point Beach Nuclear Plant (PBNP) Technical Specifications 
(TS). This change also proposes revisions to Sections 15.1, 
``Definitions,'' and 15.4.4, ``Containment Tests,'' to support the 
changes to Section 15.3.6. Additionally, the change will add 
explanatory text to the bases for Section 15.3.6 to support the 
revisions.
    The proposed revisions will add more specific limiting 
conditions for operation (LCOs) for containment isolation valves, 
air locks, overall containment air leakage, and internal pressure. 
The proposed LCOs state more clearly the requirements for 
operability and the actions to be taken if the requirements are not 
met. The proposed LCOs reflect the intent of the Westinghouse 
Improved Standard Technical Specifications, NUREG-1431. None of the 
existing LCOs are being removed.
    As most of the existing LCOs in Section 15.3.6 do not specify 
completion times for required actions, Section 15.3.0, ``General 
Considerations,'' applies. Section 15.3.0 states that if an LCO does 
not prescribe a time period, the affected unit shall be placed in 
hot shutdown within three hours of discovering the situation. If the 
conditions which prompted the shutdown cannot be corrected, the unit 
shall be placed in cold shutdown within 48 hours of discovering the 
situation. The proposed revisions to Section 15.3.6 specify 
appropriate completion times for each LCO to eliminate the need to 
default to the generic time requirements given in Section 15.3.0. If 
the required actions and associated completion times are not met, 
the plant must be brought to at least hot shutdown within 6 hours 
and to cold shutdown within 36 hours. The slightly longer time to 
hot shutdown is more than offset by the shorter time to cold 
shutdown, thereby reducing the consequences of a release from 
containment. These times are reasonable, based on operating 
experience, to reach the required plant conditions from full power 
conditions in an orderly manner without challenging plant systems.
    There is no physical change to the facility, its systems, or its 
operation. Therefore, there is no increase in the probability or 
consequences of an accident previously evaluated.
    2. The proposed amendments will not create the possibility of a 
new or different kind of accident from any accident previously 
analyzed.
    The proposed changes will add operating conditions and limiting 
conditions for operation to Section 15.3.6, ``Containment System,'' 
of the Point Beach Nuclear Plant (PBNP) Technical Specifications 
(TS). This change also proposes revisions to Sections 15.1, 
``Definitions,'' and 15.4.4, ``Containment Tests,'' to support the 
changes to Section 15.3.6. Additionally, the change will add 
explanatory text to the bases for Section 15.3.6 to support the 
revisions.
    The proposed revisions will add more specific LCOs for 
containment isolation valves, air locks, overall containment air 
leakage, and internal pressure. The proposed LCOs state more clearly 
the requirements for operability and the actions to be taken if the 
requirements are not met. The proposed LCOs reflect the intent of 
the Westinghouse Improved Standard Technical Specifications, NUREG-
1431. None of the existing LCOs are being removed.
    As most of the existing LCOs in Section 15.3.6 do not specify 
completion times for required actions, Section 15.3.0, ``General 
Considerations,'' applies. Section 15.3.0 states that if an LCO does 
not prescribe a time period, the affected unit shall be placed in 
hot shutdown within three hours of discovering the situation. If the 
conditions which prompted the shutdown cannot be corrected, the unit 
shall be placed in cold shutdown within 48 hours of discovering the 
situation. The proposed revisions to Section 15.3.6 specify 
appropriate completion times for each LCO to eliminate the need to 
default to the generic time requirements given in Section 15.3.0. If 
the required actions and associated completion times are not met, 
the plant must be brought to at least hot shutdown within 6 hours 
and to cold shutdown within 36 hours. The slightly longer time to 
hot shutdown is more than offset by the shorter time to cold 
shutdown, thereby reducing the consequences of a release from 
containment. These times are reasonable, based on operating 
experience, to reach the required plant conditions from full power 
conditions in an orderly manner without challenging plant systems.
    There is no physical change to the facility, its system, or its 
operation. Thus, a new or different kind of accident cannot occur.
    3. The proposed amendments will not involve a significant 
reduction in the margin of safety.
    The proposed changes will add operating conditions and limiting 
conditions for operation to Section 15.3.6, ``Containment System,'' 
of the Point Beach Nuclear Plant (PBNP) Technical Specifications 
(TS). This change also proposes revisions to Sections 15.1, 
``Definitions,'' and 15.4.4, ``Containment Tests,'' to support the 
changes to Section 15.3.6. Additionally, the change will add 
explanatory text to the bases for Section 15.3.6 to support the 
revisions.
    The proposed revisions will add more specific LCOs for 
containment isolation valves, air locks, overall containment air 
leakage, and internal pressure. The proposed LCOs state more clearly 
the requirements for operability and the actions to be taken if the 
requirements are not met. The proposed LCOs reflect the intent of 
the Westinghouse Improved Standard Technical Specifications, NUREG-
1431. None of the existing LCOs are being removed.
    As most of the existing LCOs in Section 15.3.6 do not specify 
completion times for required actions, Section 15.3.0, ``General 
Considerations,'' applies. Section 15.3.0 states that if an LCO does 
not prescribe a time period, the affected unit shall be placed in 
hot shutdown within three hours of discovering the situation. If the 
conditions which prompted the shutdown cannot be corrected, the unit 
shall be placed in cold shutdown within 48 hours of discovering the 
situation. The proposed revisions to Section 15.3.6 specify 
appropriate completion times for each LCO to eliminate the need to 
default to the generic time requirements given in Section 15.3.0. If 
the required actions and associated completion times are not met, 
the plant must be brought to at least hot shutdown within 6 hours 
and to cold shutdown within 36 hours. The slightly longer time to 
hot shutdown is more than offset by the shorter time to cold 
shutdown, thereby reducing the consequences of a release from 
containment. These times are reasonable, based on operating 
experience, to reach the required plant conditions in an orderly 
manner without challenging plant systems.
    There is no physical change to the facility, its systems, or its 
operation. Thus, a significant reduction in a margin of safety 
cannot occur. In fact, by adding more specific LCOs for containment 
integrity and allowing for a more orderly unit shutdown when 
required, an increased margin of safety may be realized.
    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: John N. Hannon

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 20555, and at the local public document 
rooms for the particular facilities involved.

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

    Date of application for amendment: October 22, 1993, as revised 
October 28, 1993.
    Brief description of amendment: The amendment (1) removes the 
title-specific organizational listing of the Plant Nuclear Safety 
Committee (PNSC) membership in TS 6.5.2.2 and replaces it with a 
functional description of the PNSC composition, (2) adds specific 
member qualification requirements in TS 6.5.2.3, and (3) revises 
Section 6.5.2.2 to stipulate that PNSC members shall be designated by 
the plant general manager and shall be limited in number to between 
seven and nine members.
    Date of issuance: December 28, 1993
    Effective date: December 28, 1993
    Amendment No.: 41
    Facility Operating License No. NPF-63. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62152) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Cameron Village Regional 
Library, 1930 Clark Avenue, Raleigh, North Carolina 27605.

Commonwealth Edison Company, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket 
Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 
2, Will County, Illinois; Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois; and Docket 
Nos. 50-295 and 50-304, Zion Nuclear Power Station Units 1 and 2, 
Lake County, Illinois

    Date of application for amendments: November 10, 1993
    Brief description of amendments: The amendments revised the 
``Radioactive Effluent Controls Program'' described in Section 6.0 of 
the Braidwood, Byron, LaSalle, and Zion Technical Specifications (TS) 
to be consistent with the revised 10 CFR Part 20. The changes 
specifically address the limitation on radioactive material release of 
liquid and gaseous effluents.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62152) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 30, 1993, and an 
environmental assessment noticed in the Federal Register December 30, 
1993 (58 FR 69412).
    Date of issuance: December 30, 1993
    Effective date: December 30, 1993
    Amendment Nos.: For Byron, 57 and 57; for Braidwood, 45 and 45; for 
LaSalle, 93 and 77; for Zion, 152 and 140.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72, NPF-77, 
NPF-11, NPF-18, DPR-39, and DPR-48. The amendments revise the Technical 
Specifications. No significant hazards consideration comments received: 
No
    Local Public Document Room locations: For Byron, the Byron Public 
Library, 109 N. Franklin, P.O. Box 434, Byron, Illinois 61010; for 
Braidwood, the Wilmington Township Public Library, 201 S. Kankakee 
Street, Wilmington, Illinois 60481; for LaSalle, the Public Library of 
Illinois Valley Community College, Rural Route No. 1, Oglesby, Illinois 
61348; for Zion, the Waukegan Public Library, 128 N. County Street, 
Waukegan, Illinois 60085.

Commonwealth Edison Company, Docket Nos. 50-237 and 50-249, Dresden 
Nuclear Power Station, Units 2 and 3, Grundy County, Illinois

    Date of application for amendments: June 2, 1993
    Brief description of amendments: The amendments would include 
Commonwealth Edison Company's Topical Report NFSR-0091, in Section 6.6 
of the Technical Specifications. Topical Report NFSR-0091 has been 
reviewed and approved by the NRC staff.
    Date of issuance: December 28, 1993
    Effective date: Immediately, to be implemented within 45 days.
    Amendment Nos.: 124 and 118
    Facility Operating License Nos. DPR-19 and DPR-25. The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57847). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Morris Public Library, 604 
Liberty Street, Morris, Illinois 60450.

Consolidated Edison Company of New York, Docket No. 50-247, Indian 
Point Nuclear Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: May 5, 1993, as revised by 
letter dated August 27, 1993, and supplemented by letter dated December 
21, 1993.
    Brief description of amendment: The amendment revises surveillance 
intervals for the Containment Pressure Channels, the Steam Pressure 
Channels, and the Reactor Coolant Temperature Channels to accommodate a 
24-month refueling cycle. These revisions are being made in accordance 
with the guidance provided by Generic Letter 91-04, ``Changes in 
Technical Specification Surveillance Intervals to Accommodate a 24-
Month Fuel Cycle.''
    Date of issuance: December 28, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 167
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52981) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: March 23, 1993
    Brief description of amendment: The amendment revises the 
operability requirements and the surveillance frequency for periodic 
testing of suppression chamber (torus) to drywell vacuum breakers. All, 
vice 10 of 12, vacuum breakers are required to be operable and periodic 
testing will be done each cold shutdown (if not performed within the 
last 92 days) vice monthly.
    Date of issuance: January 4, 1994
    Effective date: January 4, 1994, with full implementation within 45 
days.
    Amendment No.: 96
    Facility Operating License No. NPF-43. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: May 26, 1993 (58 FR 
30191) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 4, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Monroe County Library System, 
3700 South Custer Road, Monroe, Michigan 48161.

Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, 
Unit No. 2, Pope County, Arkansas

    Date of application for amendment: July 22,1993
    Brief description of amendment: The amendment relocated the 
containment isolation valve table from the containment systems 
specification to plant procedures, in accordance with Generic Letter 
91-08.
    Date of issuance: December 22, 1993
    Effective date: December 22, 1993
    Amendment No.: 154
    Facility Operating License No. NPF-6. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 1, 1993 (58 
FR 46231) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 22, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Tomlinson Library, Arkansas 
Tech University, Russellville, Arkansas 72801

Entergy Operations, Inc., System Energy Resources, Inc., South 
Mississippi Electric Power Association, and Mississippi Power & 
Light Company, Docket No. 50-416, Grand Gulf Nuclear Station, Unit 
1, Claiborne County, Mississippi

    Date of application for amendment: September 23, 1993
    Brief description of amendment: The amendment deleted the 
surveillance requirement for jet pump differential pressure measurement 
for operating loops when power is less than or equal to 25% rated 
thermal power (RTP) and replaces the requirement with a provision 
which, during operation at more than 25% RTP, allows continued 
operation if the jet pump surveillance is performed within 24 hours 
after RTP exceeds 25% or within 4 hours after an associated 
recirculation loop is placed in service.
    Date of issuance: January 4, 1994
    Effective date: January 4, 1994
    Amendment No: 110
    Facility Operating License No. NPF-29. Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52984) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 4, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Judge George W. Armstrong 
Library, Post Office Box 1406, S. Commerce at Washington, Natchez, 
Mississippi 39120.

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

    Date of application for amendment: August 16, 1988, as supplemented 
September 19, 1988; February 9, March 31, June 26, October 10, and 
November 22, 1989; June 21, October 15, and November 7, 1990; February 
19, April 19, June 21, August 28, and October 9, 1991; January 13, 
1992; January 18, May 28, October 24, and November 12, 1993.
    Brief description of amendment: This amendment replaces the TMI-2 
Appendix A and B Technical Specifications with the Post-Defueling 
Monitored Storage (PDMS) Technical Specifications to facilitate long 
term storage of the facility.
    Date of issuance: December 28, 1993
    Effective date: December 28, 1993
    Amendment No.: 48
    Facility Operating License No. DPR-73: The amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 25, 1991 (56 FR 
19128). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Government Publications 
Section, State Library of Pennsylvania, Walnut Street and Commonwealth 
Avenue, Box 1601, Harrisburg, Pennsylvania 17105.

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 application for amendment: October 21, 1993
    Brief description of amendments: The amendments change the 
technical specifications by revising Technical Specification 6.5.2.8, 
``Audits,'' by removing the prescriptive frequency of the required 
audits. The frequency has been removed in order to allow performance-
based inspection frequencies. This is consistent with NUREG-1431, 
``Standard Technical Specifications - Westinghouse Plants,'' September 
1992.
    Date of issuance: December 27, 1993
    Effective date: December 27, 1993, to be implemented within 10 days 
of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 56; Unit 2 - Amendment No. 
45
    Facility Operating License Nos. NPF-76 and NPF-80. Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59752). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 27, 1993. 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, Texas 77488

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: September 15, 1993, as supplemented by 
letter dated November 30, 1993.
    Brief description of amendments: The amendments consist of changes 
to the technical specifications to implement the new requirements of 10 
CFR Part 20 which were issued on May 21, 1991. The changes are either 
editorial changes that provide consistency between the technical 
specifications and the revised 10 CFR Part 20 or are changes to the 
effluent limits cross referenced to the former 10 CFR Part 50, Appendix 
I limits.
    Date of issuance: December 30, 1993
    Effective date: December 30, 1993, to be implemented on January 1, 
1994.
    Amendment Nos.: Unit 1 - Amendment No. 57; Unit 2 - Amendment No. 
46
    Facility Operating License Nos. NPF-76 and NPF-80: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57854) The November 30, 1993, submittal provided additional clarifying 
information and did not change the original proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendments is contained in a Safety Evaluation dated 
December 30, 1993, and an environmental assessment noticed in the 
Federal Register December 27, 1993 (58 FR 68445). 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, Texas 77488

Iowa Electric Light and Power Company, Docket No. 50-331, Duane 
Arnold Energy, Center, Linn County, Iowa

    Date of application for amendment: July 28, 1993
    Brief description of amendment: The amendment revised the Technical 
Specifications by incorporating the new requirements of 10 CFR Part 20. 
The revision includes changing the release rate limits for gaseous and 
liquid effluents, the monitoring and reporting requirements, the 
definitions and the record retention requirements.
    Date of issuance: December 27, 1993
    Effective date: December 27, 1993
    Amendment No.: 194
    Facility Operating License No. DPR-49. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 15, 1993 (58 
FR 48384) The Commission's related evaluation of the amendment is 
contained in an Environmental Assessment dated December 23, 1993 (58 FR 
68179), and a Safety Evaluation dated December 27, 1993. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Cedar Rapids Public Library, 
500 First Street, S. E., Cedar Rapids, Iowa 52401.

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

    Date of application for amendment: September 10, 1993, as 
supplemented November 12, 1993
    Brief description of amendment: The amendment changes the Technical 
Specifications to provide consistency with the guidance of Generic 
Letter 90-09 that relates to the revision of the surveillance 
requirements for snubbers.
    Date of issuance: December 28, 1993
    Effective date: December 28, 1993
    Amendment No.: 71
    Facility Operating License No. DPR-21. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: September 29, 1993 (58 
FR 50969). The November 12, 1993, letter provided clarifying 
information that did not change the initial proposed no significant 
hazards consideration determination. The Commission's related 
evaluation of the amendment is contained in a Safety Evaluation dated 
December 28, 1993. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Learning Resources Center, 
Thames Valley State Technical College, 574 New London Turnpike, 
Norwich, Connecticut 06360.

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 application for amendments: July 7, 1993 (Reference LAR 93-
05)
    Brief description of amendments: The amendments revise the combined 
Technical Specifications (TS) for the Diablo Canyon Power Plant Unit 
Nos. 1 and 2 by changing the gaseous effluent limit of TS 6.8.4.g., 
``Radioactive Effluent Controls Program,'' and the Bases for TS 3/
4.11.1.4, ``Liquid Holdup Tanks,'' to conform to recent revisions to 10 
CFR 20.
    Date of issuance: January 6, 1994
    Effective date: January 6, 1994
    Amendment Nos.: 84, 85
    Facility Operating License Nos. DPR-80 and DPR-82 The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: August 18, 1993 (58 FR 
43930) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 6, 1994, and an 
environmental assessment noticed in the Federal Register January 5, 
1994 (59 FR 606). No significant hazards consideration comments 
received: No.
    Local Public Document Room location: California Polytechnic State 
University, Robert E. Kennedy Library, Government Documents and Maps 
Department, San Luis Obispo, California 93407.

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

    Date of application for amendment: October 19, 1993
    Brief description of amendment: The amendment revises Technical 
Specification (TS) 6.5.2.2 to change the membership of the Safety 
Review Committee (SRC). The proposed amendment also modifies TS 
6.5.2.10 to change the time limit for providing the Executive Vice 
President - Nuclear Generation with SRC meeting minutes and reports of 
review.
    Date of issuance: December 28, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 141
    Facility Operating License No. DPR-64: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 10, 1993 (58 
FR 59755) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: White Plains Public Library, 
100 Martine Avenue, White Plains, New York 10610.

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

    Date of application for amendment: September 24, 1993
    Brief description of amendment: The amendment revises Technical 
Specification Table 4.2-2, ``Minimum Test and Calibration Frequency for 
Core and Containment Cooling Systems,'' to delete the requirement for 
calibration of time delay relays and timers in the logic system 
functional test for the containment cooling subsystem.
    Date of issuance: December 28, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 201
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62155) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: October 18, 1993
    Brief description of amendment: The amendment modifies Technical 
Specification (TS) 6.5.2.2 to change the membership requirements for 
the Safety Review Committee (SRC). The amendment also modifies TS 
6.5.2.10 to change the time limit for providing the Executive Vice 
President - Nuclear Generation with SRC meeting minutes and reports of 
review.
    Date of issuance: December 28, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 202
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62156) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: September 24, 1993
    Brief description of amendment: The amendment makes miscellaneous 
administrative changes including typographical and editorial 
corrections to the Appendix A Technical Specifications (TSs) and 
Appendix B Radiological Effluent TSs. These administrative changes do 
not result in any substantive changes to the TSs. The erroneous 
surveillance requirement in Appendix A, TS 4.3.C.3.b, has also been 
deleted.
    Date of issuance: December 29, 1993
    Effective date: As of the date of issuance to be implemented within 
30 days.
    Amendment No.: 203
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62155) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 29, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Reference and Documents 
Department, Penfield Library, State University of New York, Oswego, New 
York 13126.

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

    Date of application for amendment: April 1, 1993, as supplemented 
on July 2, 1993.
    Brief description of amendment: This amendment revises Technical 
Specifications surveillance requirements to extend the surveillance 
test intervals and allowed out-of-service times for the emergency core 
cooling system and reactor core isolation cooling system actuation 
instrumentation.
    Date of issuance: December 27, 1993
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days of the date of issuance.
    Amendment No.: 62
    Facility Operating License No. NPF-57: This amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 28, 1993 (58 FR 
25864) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 27, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: Pennsville Public Library, 190 
S. Broadway, Pennsville, New Jersey 08070

Southern California Edison Company, et al, Docket No. 50-206, San 
Onofre Nuclear Generating Station, Unit No. 1, San Diego County, 
California

    Date of application for amendment: May 12, 1993, and supplemented 
June 30, and November 23, 1993. The supplemental information submitted 
June 30, and November 23, 1993, did not affect the proposed no 
significant hazards consideration determination.
    Brief description of amendment: This amendment adds a new license 
condition 2.C.(9) concerning the San Onofre, Unit 1, Fire Protection 
Program. The licensee proposed this new license condition in accordance 
with Generic Letters 86-10 and 88-12, and as part of a request to 
replace in its entirety the existing set of operating technical 
specifications incorporated in
    Facility Operating License No. DPR-13 as Appendix A with a set of 
permanently defueled technical specifications.
    Date of issuance: December 28, 1993
    Effective date: As of the date of issuance, to be fully implemented 
no later than 120 calendar days from the date of issuance of this 
amendment.
    Amendment No.: 155
    Facility Operating License No. DPR-13: The amendment adds a new 
license condition and replaces in its entirety the existing set of 
Technical Specifications with a set of Permanently Defueled Technical 
Specifications.
    Date of initial notice in Federal Register: June 23, 1993 (58 FR 
34093) and supplemented July 7, 1993 (58 FR 36445). The Commission's 
related evaluation of the amendment is contained in a Safety Evaluation 
dated December 28, 1993. No significant hazards consideration comments 
received: No.
    Local Public Document Room location: Main Library, University of 
California, P.O. Box 19557, Irvine, California 92713

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: September 13, 1993
    Brief description of amendments: The amendments change the 
Technical Specifications to eliminate the low feedwater reactor trip 
and reduce the steam generator low-low water level reactor trip and 
safeguard actuation setpoint from 17 percent to 15 percent of narrow 
range span with a corresponding reduction in allowable value from 16 
percent to 14.4 percent.
    Date of issuance: December 29, 1993
    Effective date: December 29, 1993
    Amendment Nos.: 104 and 97
    Facility Operating License Nos. NPF-2 and NPF-8. Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993 (58 
FR 62157) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated December 29, 1993. 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

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

    Date of application for amendments: June 16, 1993; supplemented 
November 10, 1993 (TS 93-05)
    Brief description of amendments: The amendments change the lifting 
force specified in the surveillance requirements for the ice condenser 
intermediate deck doors to less restrictive values that are based on 
Westinghouse Electric Corporation analysis.
    Date of issuance: January 3, 1994
    Effective date: January 3, 1994
    Amendment Nos.: 175 - Unit 1 and 166 - Unit 2
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the technical specifications.
    Date of initial notice in Federal Register: August 4, 1993 (58 FR 
41513) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 3, 1994. No significant 
hazards consideration comments received: None
    Local Public Document Room location: Chattanooga-Hamilton County 
Library, 1101 Broad Street, Chattanooga, Tennessee 37402

Toledo Edison Company, Centerior Service Company, and The Cleveland 
Electric Illuminating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit No. 1, Ottawa County, Ohio

    Date of application for amendment: August 30, 1993
    Brief description of amendment: This amendment adds a Technical 
Specification (TS) Action statement, which applies when both 
containment hydrogen dilution systems are inoperable, and revises the 
associated TS bases section.
    Date of issuance: December 30, 1993
    Effective date: December 30, 1993
    Amendment No. 183
    Facility Operating License No. NPF-3. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: October 13, 1993 (58 FR 
52995) The commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 30, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: University of Toledo Library, 
Documents Department, 2801 Bancroft Avenue, Toledo, Ohio 43606.

TU Electric Company, Docket Nos. 50-445 and 50-446, Comanche Peak 
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas

    Date of amendment request: September 10, 1992, as supplemented by 
letter dated March 17, 1993.
    Brief description of amendment: The amendments change the ``Z'' 
value for the reactor coolant pump (RCP) undervoltage relay setpoint 
and ``allowable value'' for the RCP underfrequency relay setpoint in 
Technical Specification Table 2.2-1.
    Date of issuance: December 23, 1993
    Effective date: December 23, 1993, to be implemented within 30 days 
of issuance.
    Amendment Nos.: Unit 1 - Amendment No. 22; Unit 2 - Amendment No. 8
    Facility Operating License Nos. NPF-87 and NPF-89: Amendment 
revised the Technical Specifications.
    Date of initial notice in Federal Register: April 14, 1993 (58 FR 
19488). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 23, 1993. No 
significant hazards consideration comments received: No
    Local Public Document Room location: University of Texas at 
Arlington Library, Government Publications/Maps, 701 South Cooper, P. 
O. Box 19497, Arlington, Texas 76019.

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

    Date of application for amendment: November 3, 1992, as clarified 
on December 4, 1992
    Brief description of amendment: The amendment revises the Technical 
Specification Table 4.3-1, ``Reactor Trip System Instrumentation 
Surveillance Requirements,'' Note 5, to reflect that integral bias 
curves, rather than detector plateau curves, are used to calibrate the 
source range instrumentation.
    Date of issuance: December 28, 1993
    Effective date: December 28, 1993
    Amendment No.: 87
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 6, 1993 (58 FR 
601) The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 28, 1993. No significant hazards 
consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendment: February 25, 1992
    Brief description of amendment: The amendment revises the Technical 
Specification Section 6 to change General Manager, Nuclear Operations 
to Vice President, Nuclear Operations.
    Date of issuance: December 28, 1993
    Effective date: December 28, 1993
    Amendment No.: 88
    Facility Operating License No. NPF-30. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 29, 1992 (57 FR 
18179) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated December 28, 1993. No 
significant hazards consideration comments received: No.
    Local Public Document Room location: Callaway County Public 
Library, 710 Court Street, Fulton, Missouri 65251.

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

    Date of application for amendments: September 29, 1993
    Brief description of amendments: These amendments modify the 
required inspection frequency of the low pressure turbine blades and 
make administrative changes to the Technical Specifications.
    Date of issuance: January 5, 1994
    Effective date: January 5, 1994
    Amendment Nos.: 184 and 184
    Facility Operating License Nos. DPR-32 and DPR-37: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 27, 1993 (58 FR 
57860) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 5, 1994. No significant 
hazards consideration comments received: No
    Local Public Document Room location: Swem Library, College of 
William and Mary, Williamsburg, Virginia 23185

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

    Date of application for amendments: November 24, 1992
    Brief description of amendments: The amendments modified Technical 
Specification (TS) Section 15.3.10 (Control Rod and Power Distribution 
Limits). These changes removed reference to the hot shutdown condition 
and made the section only applicable during an approach to criticality. 
The shutdown requirements are still governed by the existing shutdown 
margin requirements specified in Section 15.3.10.A.3.
    Date of issuance: January 3, 1994
    Effective date: January 3, 1994
    Amendment Nos.: 144 and 148
    Facility Operating License Nos. DPR-24 and DPR-27. Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: March 3, 1993 (58 FR 
12270) The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated January 3, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room location: Joseph P. Mann Library, 1516 
Sixteenth Street, Two Rivers, Wisconsin 54241.

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

    Date of amendment request: October 21, 1993
    Brief description of amendment: The amendment revises Technical 
Specification 6.3.1 related to the qualification requirements for the 
position of Manager Operations and revises Technical Specification 
6.5.1.2 to delete specific title designations from the Plant Safety 
Review Committee (PSRC) membership. The qualifications for the Manager 
Operations are revised to require that the individual holds a senior 
reactor operator license or previously held a senior reactor operator 
license at a similar unit (pressurized water reactor). The revision to 
the PSRC membership requirements replaced a list of specific manager 
titles with a phrase designating managers responsible for certain 
technical areas.
    Date of issuance: January 3, 1994
    Effective date: January 3, 1994, to be implemented within 30 days 
of issuance.
    Amendment No.: 70
    Facility Operating License No. NPF-42. Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: November 24, 1993, (58 
FR 62158) The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated January 3, 1994. No significant 
hazards consideration comments received: No.
    Local Public Document Room Locations: Emporia State University, 
William Allen White Library, 1200 Commercial Street, Emporia, Kansas 
66801 and Washburn University School of Law Library, Topeka, Kansas 
66621
    Dated at Rockville, Maryland, this 10th day of January, 1994
    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Acting Director, Division of Reactor Projects - III/IV/V, Office of 
Nuclear Reactor Regulation
[Doc. 94-1100 Filed 1-18-94; 8:45 am]
BILLING CODE 7590-01-F