[Federal Register Volume 68, Number 236 (Tuesday, December 9, 2003)]
[Notices]
[Pages 68654-68675]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-30246]


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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, November 14, through November 26. The last 
biweekly notice was published on November 25, 2003 (68 FR 66131).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
Involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By January 8, 2004, 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 PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for

[[Page 68655]]

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 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent 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 PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island 
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania

    Date of amendment request: August 6, 2003.
    Description of amendment request: This amendment would revise the 
Technical Specifications (TSs) to incorporate reference to the 10 CFR 
50.55a, Codes and Standards, criteria for the inservice reactor 
building tendon surveillance requirements, to incorporate an 
administrative change to the TS Definition 1.22 to be consistent with 
10 CFR 20.1003, as well as other administrative corrections from 
previously issued TS amendments.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed revision to Technical Specification 4.4.2.1 and 
associated Bases Section incorporates reference to the criteria of 
10 CFR 50.55a, ``Codes and standards,'' in addition to the existing 
criteria of Regulatory Guide 1.35. This change provides consistency 
between the Technical Specification tendon surveillance program 
criteria and the regulatory requirements specified in 10 CFR 
50.55a(b)(2)(vi). These

[[Page 68656]]

regulatory requirements and the associated surveillance program 
ensure that the reactor building tendon prestressing system is 
capable of maintaining the structural integrity of the containment 
during operating and accident conditions. The reactor building 
prestressing system is not an initiator of any accident. Therefore, 
this change is not related to the probability of any accident 
previously evaluated. This change ensures that the containment 
tendon surveillance program addresses the appropriate regulatory 
criteria. This change does not result in any reduction in the 
effectiveness of the existing surveillance program. The tendon 
surveillance program will continue to ensure that the containment 
structure is capable of performing its intended safety function in 
the event of a design basis accident. Therefore, this change has no 
affect on the consequences of an accident previously evaluated.
    The proposed changes to Technical Specification Definition 1.22, 
Technical Specification 3.1.6.6 and associated Bases, and Technical 
Specification 3.24 Bases are only administrative changes or 
corrections and have no affect on plant design or operations. 
Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed revision to Technical Specification 4.4.2.1 and 
associated Bases Section incorporates references to the criteria of 
10 CFR 50.55a, ``Codes and standards,'' in addition to the existing 
criteria of Regulatory Guide 1.35. This change provides consistency 
between the Technical Specification tendon surveillance program 
criteria and the regulatory requirement specified in 10 CFR 
50.55a(b)(2)(vi). The proposed Technical Specification change does 
not result in any reduction in effectiveness of the existing tendon 
surveillance program. The tendon surveillance program will continue 
to satisfy the applicable Technical Specification and regulatory 
required criteria, thus ensuring that the containment structure will 
perform its design safety function. This change has no affect on the 
design and operation of plant structures, systems, and components. 
This change does not introduce any new accident precursors and does 
not involve any alterations to plant configurations, which could 
initiate a new or different kind of accident.
    The proposed changes to Technical Specification Definition 1.22, 
Technical Specification 3.1.6.6 and associated Bases, and Technical 
Specification 3.24 Bases are only administrative changes or 
corrections and have no affect on plant design or operations.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed revision to Technical Specification 4.4.2.1 and 
associated Bases Section incorporates reference to the criteria of 
10 CFR 50.55a, ``Codes and standards,'' in addition to the existing 
criteria of Regulatory Guide 1.35. This change provides consistency 
between the Technical Specification tendon surveillance program 
criteria and the regulatory requirement specified in 10 CFR 
50.55a(b)(2)(vi). The containment examination and inspection 
requirements specified in 10 CFR 50.55a(b)(2)(vi) meet the same 
standards as the criteria specified in Regulatory Guide 1.35. The 
proposed Technical Specification change does not result in any 
reduction in effectiveness of the existing tendon surveillance 
program. The tendon surveillance program will continue to satisfy 
the applicable Technical Specification and regulatory required 
criteria, thus ensuring that the containment structure will perform 
its design safety function in accordance with existing margins of 
safety for containment integrity.
    The proposed changes to Technical Specification Definition 1.22, 
Technical Specification 3.1.6.6 and associated Bases, and Technical 
Specification 3.24 Bases are only administrative changes or 
corrections and have no affect on plant design or operations.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration. 
Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice President, 
General Counsel and Secretary, Exelon Generation Company, LLC, 300 
Exelon Way, Kennett Square, PA 19348 NRC Section Chief: Richard J. 
Laufer.

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

    Date of amendments request: August 22, 2003.
    Description of amendments request: The amendments would revise 
three different sections in the Updated Final Safety Analysis Report 
(UFSAR) for PVNGS [Palo Verde Nuclear Generating Station], Units 1, 2, 
and 3. This request would revise the sections of the UFSAR which 
describe the maximum fuel pin pressurization criteria used for fuel 
handling accident safety analyses. This change is necessitated due to 
the combination of higher core burnup designs, fuel which contains 
erbia poison, and the recent introduction of ZIRLO cladded fuel to the 
PVNGS reactors.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed change would revise sections of the PVNGS UFSAR, 
which describe the maximum fuel pin pressurization criteria used for 
fuel handling accident safety analyses.
    No additional equipment is being added as a result of the 
proposed change. None of the failure modes and effects analyses are 
impacted by the proposed change since no structures, systems, or 
components (SSCs) are being modified, system lineups remain the 
same, and operator actions for fuel handling accident are not 
changing. No manual actions are being substituted for automatic 
actions. The SSCs relied upon to mitigate the event are not 
changing. Specifically, the fuel building, BOPESFAS (Balance of 
Plant-Engineered Safety Features Actuation System), radiation 
monitor setpoints, etc. . . are not impacted. The methodology 
changes will have no impact on the likelihood of a malfunction of 
any SSCs.
    No departures from the design or testing and performance 
standards outlined in any 10 CFR [Part] 50, Appendix A, General 
Design Criteria (GDC) will result from the proposed activity. The 
proposed UFSAR changes will not make any SSCs more likely to fail 
(no direct effects). Even with higher fuel pin pressures, the use of 
ZIRLO cladding provides more margin to design stress limits (liftoff 
pressure) than Zircaloy cladding. Regardless of the fuel type (and 
hence cladding type), the design stress and code allowable limits 
will not be exceeded. Palo Verde Nuclear Generating Station (PVNGS) 
``Fuel Mishandling Accident Evaluation with ZIRLO Fuel Rods'' 
concluded that the analysis of record for fuel handling events 
involving fuel assemblies containing ZIRLO cladding would remain 
bounding. No physical changes to any SSCs will be performed as a 
result of the proposed changes. In addition, system/equipment 
redundancy requirements are maintained with the proposed UFSAR 
changes.
    Fuel handling accident analyses must ensure doses at the site 
boundary and control room remains well within 10 CFR Part 100 and 10 
CFR [Part] 50 Appendix A, GDC 19 exposure guideline. Restricting the 
peak assembly average fuel pin pressure to <1200 psig will still 
result in acceptable doses. Therefore, no indirect effects on SSCs 
associated with dose limitations are impacted.
    Consequences mean dose at the Exclusion Area Boundary (EAB), Low 
Population Zone (LPZ), and Control Room; therefore, an increase in 
consequences must involve an increase in radiological doses to the 
public or to control room operators. No changes to the

[[Page 68657]]

dose exposure as a result of a fuel handling accident are proposed 
for the methodology change and Regulatory Guide 1.25 deviation 
requested. Therefore, there are no radiological consequence changes 
for this event.
    The fuel handling accident event does involve fuel barrier 
failure and does involve consequences, however no changes to the 
fuel handling dose calculation are required since the 
decontamination factor will remain unchanged even with maximum fuel 
pin pressure exceeding 1200 psig. Activities affecting on-site dose 
consequences that may require prior NRC approval are those that 
impede required actions inside or outside the control room to 
mitigate the consequences of reactor accidents.
    The proposed change does not modify any operator actions and 
hence will not impede required actions inside or outside the control 
room to mitigate the consequences of reactor accidents. The proposed 
change will not prevent or degrade the effectiveness of actions 
described or assumed in an accident discussed in the UFSAR. The 
proposed change does alter assumptions previously made in evaluating 
the radiological consequences of an accident described in the UFSAR, 
however the altered assumption is a methodology change. If the 
proposed methodology change were not applied, the calculated dose 
would increase. The peak assembly average pin pressure concept would 
allow the decontamination factor (DF) to remain the same and 
therefore consequences would remain unchanged. The proposed change 
does not play a direct role in mitigating the radiological 
consequences of an accident described in the UFSAR. The radiological 
consequences of the accident described in the UFSAR are bounding for 
the proposed activity (e.g., the results of the UFSAR analysis bound 
those that would be associated with the proposed change).
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The accident affected by the proposed change is the fuel 
handling accident (UFSAR Section 15.7.4). The proposed change does 
not involve any new equipment and does not operate any existing 
equipment in a different or more severe manner than what has 
previously been analyzed. PVNGS evaluations concluded all analyses 
of record for fuel handling events involving fuel assemblies 
containing ZIRLO cladding will remain bounding. The material 
strength of ZIRLO is significantly higher than that for Zircaloy-4. 
Since the allowable stresses for ZIRLO cladding are significantly 
higher than for Zircaloy-4, the same number of fuel rods (or fewer) 
will be damaged by the same accident scenarios as previously 
evaluated. Regardless of the fuel type (and hence cladding type), 
the design stress and code allowable limits will not be exceeded. 
Slight changes in the maximum fuel pin pressure during fuel movement 
will have no impact on the possibility of creating an accident of a 
different type as long as the design pressure structural limits of 
the fuel assembly are not approached. PVNGS calculation documents 
minimum liftoff pressures will not be challenged regardless of the 
fuel type or cladding type. Maintaining peak assembly average fuel 
pin pressure below 1200 psig will not challenge the liftoff pressure 
design basis limit for the cladding. The peak pin internal pressures 
for the hot rods never exceed the clad liftoff pressure and 
therefore the fuel pins will not be more likely to fail. Vendor 
calculation shows that the ZIRLO cladding fuel design results in a 
greater margin to the design pressure limit of the fuel cladding and 
also documents liftoff pressures are not exceeded for PVNGS fuel 
designs.
    The design function of the SSCs required to function during a 
fuel handling accident is to provide protection to ensure fuel 
damage is limited to 236 fuel pins (one fuel assembly) and ensuring 
doses do not exceed established limits. These are indirect affects. 
This change will not make a SSC more likely to fail (no direct 
affects). In fact, ZIRLO cladding fuel is less likely to fail than 
the original Zircaloy-4 cladding fuel. No physical changes to the 
SSCs will be performed as a result of the proposed change. This 
proposed change does not change the failure modes for the SSCs 
required to operate for the fuel handling accident. The cladding 
calculations document design stress or code allowable limits will 
not be exceeded. Hence, system/equipment redundancy requirements are 
maintained. Fuel handling accident analyses must ensure doses at the 
site boundary remain within acceptable design limits. The cladding 
calculations document fuel pin pressures do not exceed the design 
pressure ratings for the fuel assembly. Therefore, no indirect 
effects on SSCs associated fuel clad pressure boundary exist. None 
of the failure modes and effects analyses are impacted by this 
methodology change since no SSCs are being modified, system lineups 
remain the same, and operator actions are not changing.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The Nuclear Reactor Regulation (NRR) Safety Evaluation, 
``Related to Task Interface Agreement 99-03 Regarding Potential 
Nonconservative Assumptions for Fuel Handling Accident, McGuire 
Nuclear Station, dated November 24, 1999,'' states in part, ``The 
NRR staff has concluded that the increased rod pressures associated 
with extended bumup fuel can be expected to decrease the value of 
the iodine DF. However, the NRR staff believes that the iodine DF 
value of 100 provided in Regulatory Guide 1.25 has sufficient margin 
to compensate for the increases in rod gas pressures at current 
allowable bumup levels and for the expected increases in gap release 
fractions. Conservatisms in the assessment of the amount of fuel 
damage provide additional margin. Design basis fuel handling 
accidents are not considered to have a high risk significance. On 
the basis of these findings, the staff concludes that there is 
reasonable assurance that adequate protection of the public from the 
effects of design basis fuel handling accidents involving fuel with 
peak rod average bumups as high as 62 GWD/MTU will continue.''
    To assess the margin of safety, the methodology specified in 
Regulatory Guide 1.183, [``]Alternative Radiological Source Terms 
for Evaluating Design Basis Accidents at Nuclear Power Reactors,[''] 
was evaluated. This regulatory guide suggests a DF of 200 for 
iodine. This DF is well above the DF of 100 specified by Regulatory 
Guide 1.25.
    APS [Arizona Public Service] proposes that ample margin is 
retained to justify the continued used [use] of an overall 
decontamination factor of 100 at a peak assembly average fuel pin 
pressure of 1200 psig.
    Therefore, APS has concluded that the proposed license amendment 
request does not involve a significant reduction in a margin of 
safety.
    Based on the above, APS concludes that the [activities 
associated with] the proposed amendment(s) present no significant 
hazards consideration under the standards set forth in 10 CFR 50.92 
[``Issuance of Amendment,''] (c) and, accordingly, a finding of ``no 
significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendments request: September 17, 2003.
    Description of amendments request: The amendments would revise 
sections of the Technical Specifications (TS) to support replacement of 
the part length control element assemblies (PLCEAs) with a new design 
that contains neutron absorber over the entire control section of the 
CEA. The replacements are referred to as part strength control element 
assemblies (PSCEAs). Additionally, a change is proposed to TS 3.1.5--
``Control Element Assembly (CEA) Alignment,'' Condition B, to eliminate 
a potential condition which could cause an unwarranted plant shutdown.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the

[[Page 68658]]

issue of no significant hazards consideration, which is presented 
below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The physical difference between the 4-finger full strength 
control element assemblies (FSCEAs) and the PSCEAs involves using 
Inconel rather than B4C (boron carbide) over 100% of the 
active control section of each CEA finger. In addition, the PSCEAs 
use Inconel tubing to encase solid Inconel slugs, which cover the 
entire control section of the control element assembly (CEA). The 
current PLCEAs (also have only 4-fingers) use solid Inconel rods for 
only the lower half of each finger and B4C pellets in the 
top 15 inches (10%) of the control section of the CEA. Although 
failure of the solid Inconel region due to neutron fluence would be 
less likely than a typical clad design, the differences in swelling 
between the Inconel slugs encased by Inconel clad for the PSCEAs 
will be minor and result in a minimal impact on clad integrity. With 
the exception of the neutron absorber, the cladding design used for 
the PSCEAs is similar to the cladding of the full strength CEAs 
(FSCEAs). The geometry, cladding materials, and the spider assembly 
that supports the CEA fingers are essentially the same for the 4-
finger FSCEAs and the PSCEAs. The principal difference results from 
the Inconel slugs contained in the PSCEAs being heavier than the 
B4C pellets used in the FSCEAs. Even though the weight of 
a 4-finger PSCEA is greater than the weight of a 4-finger PLCEA or a 
4-finger FSCEA, this weight difference is bounded by the 12-finger 
FSCEAs which are operated by the same CEA drive mechanism system.
    The PSCEAs use Inconel as a neutron absorber in the entire 
control section of each CEA finger and will be operationally used 
the same way as the PLCEAs. In particular, the insertion restraints 
that are defined by the power dependent insertion limits (PDILs) for 
the PLCEAs will remain the same for the PSCEAs. This existing 
requirement will not result in any significant operational impact on 
the PSCEAs since the solid Inconel cylinder in the bottom 50% 
(operating range of the PDILs) of the PLCEAs has essentially the 
same reactivity worth as that of the PSCEAs.
    In addition, renaming the full length CEAs and part length CEAs 
to full strength CEAs and part strength CEAs, respectively, and 
providing definition for the PSCEAs will not impact the safe 
operation of the plant. The terminology will be appropriately 
changed in any related document, equipment tag, or indication on a 
control panel.
    The PLCEAs are not credited in the accident analyses for 
accident mitigation. The PSCEA design eliminates an accident 
scenario involving the insertion of a PLCEA past the PDIL, which 
results in an axial shift in power due to the upper region of the 
PLCEAs which has no neutron absorber. This condition will not occur 
with the PSCEAs because they are filled with neutron absorber over 
100% of the control section of each finger.
    Concerning TS Limiting Condition for Operation (LCO) 3.1.5, 
Condition B, proposed change; there are three position indicator 
channels available for each CEA. Current TS Bases state that, ``At 
least two of the following three CEA position indicator channels 
shall be OPERABLE for each CEA.'' Additionally the TS Bases states, 
``If only one CEA position indicator channel is OPERABLE, continued 
operation in MODES 1 and 2 may continue, provided, within 6 hours, 
at least two position indicator channels are returned to OPERABLE 
status; or within 6 hours and once per 12 hours, verify that the CEA 
group with the inoperable position indicators are either fully 
withdrawn or fully inserted while maintaining the insertion limits 
of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8.'' The TS Bases make no 
restriction or condition limiting only one CEA within a subgroup to 
having only one CEA position indication channel. Current analyses 
already assume that more than one CEA in a subgroup could have only 
one position indicator OPERABLE. Modifying the wording for Condition 
B, of LCO 3.1.5, will not affect the likelihood or consequences of a 
CEA drop, slip, ejection, or misalignment. This change will still 
require at least one position indication channel be available for 
each CEA.
    Consequently, the proposed change does not involve a significant 
increase in the probability or consequences of an accident.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes do not introduce any new mode of plant 
operation and the PSCEAs, like the PLCEAs, are not relied upon for 
accident mitigation. The PSCEAs will be operated in exactly the same 
manner in which the PLCEAs are operated. The existing operating 
restrictions for the PLCEAs will apply to the PSCEAs. In particular, 
the power dependent insertion limit (PDIL) restrictions identified 
in the Core Operating Limits Report (COLR) will remain the same for 
the PSCEAs. The PSCEA design uses Inconel over the entire control 
section of each CEA finger, which will prevent the potential 
undesired flux redistribution currently associated with the 
misoperation of PLCEAs. Therefore, the analysis associated with the 
undesired flux redistribution misoperation for the PLCEAs will be 
eliminated from PVNGS [Palo Verde Nuclear Generating Station] safety 
analyses. PSCEA misoperation events are bounded by the existing 
PLCEA and FSCEA misoperation safety analyses.
    In addition, renaming (within the Technical Specifications) the 
``full length CEAs'' and ``part length CEAs'' to ``full strength 
CEAs'' and ``part length or part strength CEAs,'' respectively, and 
providing a definition for the PSCEAs will not impact the safe 
operation of the plant. The terminology will be appropriately 
changed in any related document, equipment tag, or indication on a 
control panel.
    Concerning TS LCO 3.1.5, Condition B proposed change, CEA 
position indication channels have no control function and provide 
input to the CEA Calculators (CEACs) and Core Protection Calculators 
(CPCs) for generation of a penalty factor. This change will still 
require at least one position indication channel be available for 
each CEA. Allowing Condition `B' of LCO 3.1.5 to apply to more than 
one CEA per group does not create the possibility of a different 
type of malfunction than previously evaluated in the UFSAR [Updated 
Final Safety Analysis Report].
    Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The design of the PSCEAs is very similar to the FSCEAs except 
for the neutron absorber within each finger of a PSCEA. The PSCEAs 
do not have as strong of a neutron absorber (Inconel) as that which 
is contained in the FSCEAs (B4C). There is a weight 
difference which results from the Inconel slugs contained in the 
PSCEAs being heavier than the B4C pellets used in the 
FSCEAs. Even though the weight of the 4-finger PSCEAs is greater 
than the weight of the 4-finger PLCEAs, the CEA drive mechanism and 
support components shall operate within their design bases. 
Therefore, the PSCEAs can be considered adequate for safety-related 
applications. Consequently, the differences in design between the 
current PLCEAs and the PSCEAs do not adversely impact safe 
operation.
    The PLCEAs are not relied upon for shutdown margin or accident 
mitigation and no new requirements will apply to the PSCEAs. 
However, the design of the PSCEAs is effectively eliminating the 
concern associated with the insertion of the PLCEAs past the PDILs 
which could result in an undesirable shift in neutron flux to the 
top of the core due to the region within the PLCEAs that do not have 
neutron absorber. The PSCEAs have neutron absorber throughout their 
entire control section, which prevents a neutron flux shift to the 
top of the core if inserted past the PDIL, when compared to that of 
the PLCEAs.
    In addition, renaming the ``full length CEAs'' and ``part length 
CEAs'' to ``full strength CEAs'' and ``part length or part strength 
CEAs,'' respectively, and providing definition for the PSCEAs will 
not impact the safe operation of the plant. The terminology will be 
appropriately changed in any related document, equipment tag, or 
indication on a control panel.
    Concerning TS LCO 3.1.5, Condition B, proposed change, the 
current licensing bases already consider having more than one CEA in 
a CEA group with only one available position indication. The TS 
Bases for LCO 3.1.5, Condition B state that, ``At least two of the 
following three CEA position indicator channels shall be OPERABLE 
for each CEA.'' Additionally the Bases states, ``If only one CEA 
position indicator channel is OPERABLE, continued operation in MODES 
1 and 2 may continue, provided, within 6 hours, at least two 
position indicator channels are returned to OPERABLE status; or 
within 6 hours and once per 12 hours, verify that the CEA group with 
the inoperable position indicators are either fully

[[Page 68659]]

withdrawn or fully inserted while maintaining the insertion limits 
of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8.'' The TS Bases make no 
restriction or condition limiting only one CEA within a subgroup, to 
having only one CEA position indication channel OPERABLE. Therefore, 
modifying the wording for LCO 3.1.5, Condition B, does not involve a 
significant reduction in the margin of safety since loss of 
indication to more than one CEA is already considered in the 
licensing bases.
    Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    Based on the above, APS [Arizona Public Service] concludes that 
the activities associated with the proposed amendment(s) present no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92 [``Issuance of Amendment,''] (c) and, accordingly, a 
finding of ``no significant hazards consideration'' is justified.

    The NRC staff has reviewed the licensee's analysis and, based on 
that review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: October 7, 2003.
    Description of amendment request: The licensee is proposing to 
revise Technical Specification (TS) Section 5.5.6, ``Pre-Stressed 
Concrete Containment Tendon Surveillance Program,'' for consistency 
with the requirements of 10 CFR 50.55a(g)(4) for components 
classified as Code Class CC. The proposed revision to TS 5.5.6 is to 
indicate that the Containment Tendon Surveillance Program, 
inspection frequencies, and acceptance criteria shall be in 
accordance with Section XI, Subsection IWL of the American Society 
of Mechanical Engineers Boiler and Pressure Vessel Code and the 
applicable addenda as required by 10 CFR 50.55a, except where an 
exemption or relief has been authorized by the NRC. The licensee has 
also proposed to delete the provisions of Surveillance Requirement 
3.0.2 from this specification.
    In addition, the licensee is proposing to revise TS 5.5.16, 
``Containment Leakage Rate Testing Program,'' to add exceptions to 
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test 
Program.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed changes would revise Technical Specification (TS) 
Section 5.5.6, ``Pre-Stressed Concrete Containment Tendon 
Surveillance Program,'' and Section 5.5.16, ``Containment Leakage 
Rate Testing Program,'' for consistency with the requirements of 10 
CFR 50.55a(g)(4) for components classified as Code Class CC. The 
revised requirements do not affect the function of the containment 
post-tensioning system components. The post-tensioning systems are 
passive components whose failure modes could not act as accident 
initiators or precursors. The improved inspections required by the 
American Society of Mechanical Engineers (ASME) Code serve to 
maintain containment response to accident conditions, by causing the 
identification and repair of defects in the containment.
    The proposed change affects the frequency of visual examinations 
that will be performed for the concrete surfaces of the containment 
for the purpose of the Containment Leakage Rate Testing Program. In 
addition, the proposed change allows those examinations to be 
performed during power operation as opposed to during a refueling 
outage. The frequency of visual examinations of the concrete 
surfaces of the containment and the mode of operation during which 
those examinations are performed has no relationship to or adverse 
impact on the probability of any of the initiating events assumed in 
the accident analyses. The proposed change would allow visual 
examinations that are performed pursuant to NRC approved ASME Code 
Section XI requirements (except where relief has been granted by the 
NRC) to meet the intent of visual examinations [as] required by 
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test 
Programs,'' without requiring additional visual examinations 
pursuant to the Regulatory Guide. The intent of early detection of 
deterioration will continue to be met by the more rigorous 
requirements of the ASME Code[-]required visual examinations. As 
such, the safety function of the containment as a fission product 
barrier is maintained.
    The proposed amendment does not impact any accident initiators, 
analyzed events, or assumed mitigation of accident or transient 
events. The proposed changes do not involve the addition or removal 
of any equipment or any design changes to the facility.
    Therefore, this proposed change does not represent a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change revises the Technical Specification 
administrative controls programs for consistency with the 
requirements of 10 CFR 50.55a(g)(4) for components classified as 
Code Class CC. The function of the containment post-tensioning 
system components are not altered by this change. The improved 
inspections required by the American Society of Mechanical Engineers 
(ASME) Code serve to maintain containment response to accident 
conditions, by causing the identification and repair of defects in 
the containment. In addition, the change affects the frequency of 
visual examinations that will be performed for the concrete surface 
containments. The proposed change also allows those examinations to 
be performed during power operation as opposed to during a refueling 
outage. Therefore, this change updates the Technical Specifications 
to meet the current regulations and eliminates duplication of 
requirements. The safety function of the containment as a fission 
product barrier will be maintained.
    Therefore, this proposed change does not create the possibility 
of an accident of a different kind than previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed change revises the improved Standard Technical 
Specification administrative controls programs for consistency with 
the requirements of 10 CFR 50.55a(g)(4) for components classified as 
Code Class CC. The function of the containment post-tensioning 
system components are not altered by this change. The change also 
affects the frequency of visual examinations that will be performed 
for the concrete surface containments. In addition, the proposed 
change allows those examinations to be performed during power 
operation as opposed to during a refueling outage. The change 
ensures that containment integrity [will be maintained] and ensures 
that the safety function of the containment as a fission product 
barrier will be maintained.
    Therefore, this proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendments request involves no significant hazards consideration.
    Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona 
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix, 
Arizona 85072-2034.
    NRC Section Chief: Stephen Dembek.

Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power 
Station, Unit No. 1, New London County, Connecticut

    Date of amendment request: September 18, 2003.
    Description of amendment request: The licensee is proposing to 
revise the Design Features Technical Specification 4.2, ``Fuel 
Storage.'' The licensee's technical specification change implements the 
following proposed changes:
    (1) Eliminates all credit for Boraflex as a neutron absorber.

[[Page 68660]]

    (2) Reduces the number of fuel assemblies allowed to be stored in 
the spent fuel pool (SFP) from 3229 to 2959. The fuel will be 
prohibited from being stored in 270 specific storage rack locations. 
This is necessary to support the elimination of all credit for 
Boraflex.
    (3) Changes the required spent fuel pool keff to <=0.95. 
This is necessary to support the elimination of all credit for 
Boraflex.
    (4) Eliminates the design features requirements on new fuel 
storage, since Millstone Unit No. 1 (MP1) is a plant that has ceased 
power operation and will no longer receive new fuel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    Accidents previously evaluated are the fuel handling accidents[, 
as] described in the Decommissioned Safety Analysis Report (DSAR), 
and a seismic event, which is considered as part of the spent fuel 
rack design.
    Since there are no changes to plant hardware, nor any changes in 
how fuel is moved, there are no changes to the probability of a fuel 
handling accident. The consequences of a fuel handling accident are 
not affected, since none of the inputs to the fuel handling accident 
is affected.
    The proposed changes affect the criticality analysis of the 
spent fuel storage racks. The spent fuel racks will continue to be 
able to perform their design function, which is to maintain the 
stored fuel in a sub-critical and cooled condition under all normal 
and postulated accident conditions. There are no physical hardware 
changes to the plant from these proposed changes. The revised 
criticality analysis submitted with these proposed changes 
demonstrates that fuel will be maintained in a sub-critical 
condition during all normal and postulated accident conditions, 
including the seismic event. Since there is no change in the ability 
of the fuel storage racks to maintain a sub-critical condition due 
to a seismic event, there is no change in the probability or 
consequences of this accident.
    Reducing the amount of fuel storage is a conservative action, 
and the spent fuel racks were designed and licensed to allow empty, 
partially filled, or completely full storage racks. Thus the fuel 
racks will continue to be able to perform their design function to 
maintain the fuel in a coolable condition.
    The change to the new fuel storage racks is to delete the 
Technical Specification requirements for the new fuel storage 
keff limits. Since MP1 is a plant that has ceased power 
operation and will no longer receive new fuel, there is no need for 
these Technical Specification requirements. There are no new fuel 
related accidents previously analyzed, therefore this change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    In summary, the proposed changes do not involve an increase in 
the probability or consequences of an accident previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    Since there are no changes to the plant equipment, there is no 
possibility of a new or different kind of accident being initiated 
or affected by equipment issues.
    Reducing the number of fuel assemblies to be stored in the pool, 
and discontinuing credit for Boraflex are conservative changes that 
do not introduce any new or different kind of failure modes.
    The changes made primarily affect the nuclear criticality 
analysis and do not create a new or different kind of accident. 
Changes in eliminating Boraflex credit, restricting fuel in certain 
storage locations, and changing the allowable keff limit 
are all impacts to the nuclear criticality analysis for the SFP. The 
SFP criticality analysis is part of the basic design of the system 
and is not an accident. The ability to maintain the SFP 
keff less than or equal to 0.95, as well as within the 10 
CFR Part 50 Appendix A, ``General Design Criteria for Nuclear Power 
Plants,'' Criterion 62 ``Prevention of Criticality in Fuel Storage 
and Handling'' (Reference 6) criteria of sub-critical, have been 
evaluated. Criticality impacts are more appropriately discussed 
under the margin of safety criterion.
    The change to the new fuel storage racks is to delete the 
Technical Specification requirements for the new fuel storage 
keff limits. Since MP1 is a plant that has ceased power 
operation and will no longer receive new fuel, there is no need for 
these Technical Specification requirements. Since Millstone 1 
currently has no new fuel and new fuel cannot be received, this 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    In summary, the proposed changes do not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The margin of safety relevant to the SFP is defined as (1) SFP 
keff remains sub-critical by an acceptable margin, and 
(2) the spent fuel in the SFP remains adequately cooled so that the 
fission product barriers remain intact.
    The industry and regulatory accepted value for [the] required 
sub-criticality margin[s] in the SFP is to ensure that the 
keff of the SFP remains <=0.95 under all normal and 
postulated accident conditions. This is documented in the Standard 
Review Plan, Regulatory Guide 1.13, and ANSI/ANS-57.2, ``American 
National Standard Design Requirements for LWR Spent Fuel Storage 
Facilities at Nuclear Power Plants.'' The current MP1 Technical 
Specifications require a more conservative value of 0.90 for SFP 
keff. The proposed Design Features Technical 
Specification changes the maximum SFP keff from 0.90 to 
0.95. This is not a significant reduction in the margin to [of] 
safety since the proposed value of 0.95 is consistent with the 
accepted regulatory guidance for [the] sub-criticality margin. The 
proposed criticality analysis demonstrates that the SFP 
keff remains <=0.95 on a 95/95 basis under all normal and 
postulated accident conditions, thus the required margin of 
criticality safety has been maintained.
    The proposed changes conservatively reduce the amount of fuel 
that can be stored, and therefore do not affect the SFP cooling 
analysis. Therefore, the spent fuel in the SFP remains adequately 
cooled so that the fission product barriers remain intact.
    The removal of Technical Specification requirements for the new 
fuel storage keff limits does not affect the margin of 
safety since new fuel can no longer be received.
    Therefore, based on the above, 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.
    Attorney for licensee: Lilliam M. Cuoco, Esq., Senior Nuclear 
Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road, 
Waterford, Connecticut 06385.
    NRC Section Chief: Stephen Dembek.

Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee 
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.

    Date of amendment request: October 16, 2001; as supplemented by 
letters dated May 20, September 12, and November 21, 2002; and January 
27, September 22, and November 20, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications to incorporate changes resulting 
from the use of an alternate source term and the implementation of 
several plant modifications. Publications of the Proposed No 
Significant Hazards Consideration Determination and Opportunity for 
Hearing have already appeared in the Federal Register on January 22, 
2002 (67FR2922) and October 14, 2003 (68FR59215). The November 20, 
2003, submittal contained a revised No Significant Hazards 
Consideration Determination.
    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:


[[Page 68661]]


    Standards for determining whether a license amendment involves 
no significant hazards considerations are contained in 
10CFR50.92(c). The TS [Technical Specification] changes and 
modifications as proposed in this LAR [license amendment request] 
have been evaluated in accordance with 10 CFR 50.92 and determined 
not to involve any significant hazards considerations.
    The proposed LAR includes (1) implementing the AST [alternate 
source term] for accident analysis as described in Regulatory Guide 
1.183; (2) removing the PRVS [penetration room ventilation system] 
and relaxing the SFPVS [spent fuel pool ventilation system] TS 
because they are no longer credited for Control Room and off-site 
doses; (3) revising the CRVS [control room ventilation system] to 
allow for a one time completion time extension on Conditions B and C 
when entering the conditions to support implementation of the 
Control Room intake/booster fan modification; (4) lowering the 
Reactor Building leakage rate from 0.25 w%/day to 0.20 w%/day; (5) 
revising the VFTP [ventilation filter testing program] radioactive 
methyl iodide removal acceptance criterion for SFPVS and CRVS 
Booster Fan trains; and (6) adoption of TSTF [Technical 
Specification Task Force]-51.
    Plant modifications are also being proposed in concert with the 
proposed TS changes. They include relocating the existing Control 
Room outside air intake from the roof of the Auxiliary Building to 
the roof of the Turbine Building and installing dual intakes for 
each Control Room; re-routing HPI [high-pressure injection]/LPI 
[low-pressure injection] relief valve discharge back into the 
Reactor Building and replacing the existing Caustic Addition system 
with a passive system.

    As a result of this evaluation, Duke has concluded:

    1. The proposed amendment will not involve a significant 
increase in the probability of consequences of an accident 
previously evaluated.
    The AST and those plant systems affected by implementing the 
proposed changes to the TS are not assumed to initiate design basis 
accidents. The AST does not affect the design or operations of the 
facility. Rather, the AST is used to evaluate the consequences of a 
postulated accident. The implementation of the AST has been 
evaluated in the revisions to the analysis of the design basis 
accidents for ONS [Oconee Nuclear Station]. Based on the results of 
these analyses, it has been demonstrated that, with the requested 
changes, the dose consequences of these events meet the acceptance 
criteria of 10 CFR 50.67 and Regulatory Guide 1.183. Therefore, the 
proposed amendment will not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The proposed amendment will not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    The AST and those plant systems affected by implementing the 
proposed changes to the TS are not assumed to initiate design basis 
accidents. The systems affected by the changes are used to mitigate 
the consequences of an accident that has already occurred. The 
proposed TS changes and modifications do not significantly affect 
the mitigative function of these systems. Consequently, these 
systems do not alter the nature of events postulated in the Safety 
Analysis Report nor do they introduce any unique precursor 
mechanisms. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed amendment will not involve a significant 
reduction in the margin of safety.
    The implementation of the AST, proposed changes to the TS and 
the implementation of the proposed modifications have been evaluated 
in the revisions to the analysis of the consequences of the design 
basis accidents for the ONS. Based on the results of these analyses, 
it has been demonstrated that with the requested changes the dose 
consequences of these events meet the acceptance criteria of 10 CFR 
50.67 following the provisions of Regulatory Guide 1.183. Thus, the 
proposed amendment will not involve a significant reduction in the 
margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005,
    NRC Section Chief: John A. Nakoski.

Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station, 
Unit 2, Oconee County, South Carolina

    Date of amendment request: October 28, 2003.
    Description of amendment request: The proposed amendment would 
revise the licensing basis in the Updated Final Safety Analysis Report 
to support installation of a passive low-pressure injection (LPI) cross 
connect inside containment. The proposed changes would revise the 
licensing basis for selected portions of the core flood and LPI piping 
to allow exclusion of the dynamic effects associated with postulated 
pipe rupture of that piping by application of leak-before-break 
methodology. A similar amendment was approved for Unit 1 by NRC letter 
dated September 29, 2003.
    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:

    Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the 
determination that this amendment request involves a No Significant 
Hazards Consideration by applying the standards established by the 
NRC regulations in 10 CFR 50.92. This ensures 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:
    The proposed LAR [license amendment request] modifies the Unit 2 
licensing basis to allow the dynamic effects associated with 
postulated pipe rupture of selected portions of the Unit 2 LPI [low-
pressure injection]/Core Flood (CF) piping to be excluded from the 
design basis. The proposed design allowances for these selected 
portions of piping continue to allow the LPI system design to meet 
GDC [General Design Criterion] 4 requirements related to 
environmental and dynamic effects. The proposed LAR will continue to 
ensure that ONS [Oconee Nuclear Station] can meet design basis 
requirements associated with the LPI safety function. The addition 
of the crossover line will enhance the ability of the control room 
operator to mitigate the consequences of specific events for which 
LPI is credited. Therefore, the proposed LAR does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (2) Create the possibility of a new or different kind of 
accident from any kind of accident previously evaluated:
    The proposed LAR modifies the Unit 2 licensing basis to allow 
the dynamic effects associated with postulated pipe rupture of 
selected portions of the Unit 2 LPI/Core Flood (CF) piping to be 
excluded from the design basis. The proposed design allowances for 
these selected portions of piping continue to allow the LPI system 
design to meet GDC 4 requirements related to environmental and 
dynamic effects. The systems affected by the changes are used to 
mitigate the consequences of an accident that has already occurred. 
The proposed licensing basis change does not affect the mitigating 
function of these systems. Consequently, these changes do not alter 
the nature of events postulated in the Safety Analysis Report nor do 
they introduce any unique precursor mechanisms. Therefore, the 
proposed amendment will not create the possibility of a new or 
different kind of accident from any accident previously evaluated.
    (3) Involve a significant reduction in the margin of safety:
    The proposed licensing basis change does not unfavorably affect 
any plant safety limits, set points, or design parameters. The 
change also do [SIC] not unfavorably affect the fuel, fuel cladding, 
RCS [reactor coolant system], or containment integrity. Therefore, 
the proposed licensing basis change, which adds new design 
allowances associated with the passive LPI cross connect 
modification, do [SIC] not involve a significant reduction in the 
margin of safety.


[[Page 68662]]


    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.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

Energy Northwest, Docket No. 50-397, Columbia Generating Station, 
Benton County, Washington

    Date of amendment request: November 4, 2003.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
technical specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4 exceptions in individual TSs would be 
eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect 
the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated November 4, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Thomas C. Poindexter, Esq., Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana

    Date of amendment request: October 21, 2003.
    Description of amendment request: The proposed change would remove 
MODE restrictions that currently prevent performance of Surveillance 
Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division III DC 
electrical power subsystem while in MODE 1, 2, or 3. These 
surveillances verify that the battery capacity is adequate to perform 
its required functions. The changes would allow the performance of SR 
3.8.4.7 and SR 3.8.4.8 during normal plant operations rather than only 
during refueling outages.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The power supplied by the battery is used as a source of control 
and motive power for the HPCS [High Pressure Core Spray] system 
logic, HPCS diesel-generator set control and protection, and other 
Division III related controls. The loads supplied by this system are 
loads associated with Division III of the Emergency Core Cooling 
System (ECCS).
    The battery testing period is within the period of time that the 
system will already be out of service for other planned maintenance. 
The battery test does not increase unavailability of the supported 
system or represent any change in risk above the current practice of 
planned system maintenance outages as currently allowed by the TS 
[Technical Specification]. Any risk associated with the testing of 
the Division III batteries will be enveloped by the risk management 
of the system outage.
    The out of service condition is controlled and evaluated for 
safety implications in accordance with 10 CFR 50.65 [``Requirements 
for monitoring the effectiveness of maintenance at nuclear power 
plants'']. The HPCS system reliability and availability are 
monitored and evaluated in relationship to Maintenance Rule goals to 
ensure that total outage times do not degrade operational safety 
over time.

[[Page 68663]]

    Therefore, the proposed change will have no effect on the 
probability or consequences of any previously evaluated accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The request involves the testing of the HPCS battery on-line 
while the system is already out of service. The testing will not add 
additional out of service time. Testing during this period has no 
influence on, nor does it contribute in any way to, the possibility 
of a new or different kind of accident or malfunction from those 
previously analyzed. The method of performing this test is not 
changed. No new accident modes are created by testing during the 
period when the system is already unavailable. Because the system is 
already out of service, no safety-related equipment or safety 
functions are altered as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The battery testing will be performed when the HPCS system is 
already out of service for maintenance. The out of service condition 
is controlled and evaluated for safety implications in accordance 
with 10 CFR 50.65. The batteries are not expected to be unavailable 
for more than 36 hours. This testing period is within the period of 
time that the system will already be out of service for other 
planned maintenance. Therefore, the battery test does not increase 
unavailability of the supported system or represent any change in 
risk above the current practice of planned system maintenance 
outages as currently allowed by the TS. Timing of this test has no 
effect on any fission product barrier.
    Therefore, the propose change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005.
    NRC Section Chief: Robert A. Gramm.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of amendment request: October 21, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification Section 5.5.7, ``Steam Generator (SG) 
Tube Surveillance Program,'' to allow a one-time extension of the 
frequency for examination of the SG tubes. Specifically, the amendment 
would extend the examination, currently due no later than November 17, 
2004, to June 17, 2006.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    There is no direct increase in SG leakage because the proposed 
change does not alter the plant design. The scope of the inspection 
performed during the first refueling outage subsequent to SG 
replacement (last outage), exceeded the technical specification 
requirements for the first two refueling outages combined, after 
replacement. More tubes were inspected than were required by the 
technical specifications. Indian Point 2 does not have an active SG 
damage mechanism and will meet the current industry examination 
guidelines without performing inspections during the next refueling 
outage. The results of the Condition Monitoring Assessment 
subsequent to the last outage, demonstrated that all performance 
criteria were met during the last operating period. The results of 
the aforementioned Operational Assessment show that all performance 
criteria will be met over the proposed operating period.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change will not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of the inspections performed during the last (first after 
SG replacement) refueling outage significantly exceeds the Technical 
Specification requirements for the scope of the first two refueling 
outages combined subsequent to SG replacement.
    The proposed change does not affect the SG design, the method of 
operation, or reactor coolant chemistry controls. No new equipment 
is being introduced, and installed equipment is not being operated 
in a new or different manner. The proposed change involves a one-
time extension of the SG tube inservice inspection frequency, and 
therefore will not give rise to new failure modes. In addition, the 
proposed change does not impact any other plant system or 
components.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    SG tube integrity is a function of design, environmental, and 
current physical condition. Extending the SG tube inservice 
inspection frequency by one operating cycle will not alter the 
function or design of the SGs. Inspections conducted prior to 
placing the SGs into service (pre-service inspection) and inspection 
during the first refueling outage following SG replacement, 
demonstrate that the SGs do not have fabrication damage or an active 
damage mechanism. The scope of those inspections significantly 
exceeds those required by the technical specifications. These 
inspection results were comparable to similar inspection results for 
the same model SG installed at other plants, and subsequent 
inspections at those plants provided results that support the 
extension request. The improved design of the replacement SGs also 
provides assurance that significant tube degradation is not likely 
to occur over the proposed operating period.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. John Fulton, Assistant General Counsel, 
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY 
10601.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: June 30, 2003, as supplemented by letter 
dated November 20, 2003.
    Description of amendment request: The proposed amendment would (1) 
reorganize the Arkansas Nuclear One, Unit No. 2 (ANO-2) Technical 
Specifications (TSs) Section 6.0, Administrative Controls, (2) modify 
the ANO-2 Facility Operating License, and actions and surveillance 
requirements (SRs) of various other TSs, to support the reorganization 
of Section 6.0, and (3) modify several actions and SRs that are related 
to systems that are shared by ANO-2 and Arkansas Nuclear One, Unit No. 
1 (ANO-1). These changes are being proposed so that the philosophy and 
location of the TSs in Section 6.0 reflect the recently approved 
conversion of the ANO-1 TSs to the Improved Technical Specifications 
(ITS) and the subsequent amendments to the ANO-1 ITS. This amendment 
request supersedes the

[[Page 68664]]

previous application related to the revision of TS Section 6.0 dated 
January 31, 2002, as supplemented on June 26 and July 18, 2002. The 
January 31, 2002, application was previously noticed in the Federal 
Register on March 19, 2002 (67 FR 12602), and the June 30, 2003, 
application was previously noticed in the Federal Register on July 22, 
2003 (68 FR 43385).
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.

Administrative Changes

    The proposed changes involve reformatting and rewording of the 
existing TSs. The reformatting and rewording process involves no 
technical changes to existing requirements. As such, the proposed 
changes are administrative in nature and do not impact initiators of 
analyzed events or assumed mitigation of accident or transient 
events.

Less Restrictive--Administrative Deletion of Requirements

    The proposed changes relocate requirements from the TSs to other 
license basis documents which are under licensee control. The 
documents containing the relocated requirements will be maintained 
using the provisions of applicable regulatory requirements.

More Restrictive Changes

    The proposed changes provide more stringent requirements for the 
ANO-2 TSs. These more stringent requirements are not assumed to be 
initiators of analyzed events and will not alter assumptions 
relative to mitigation of accident or transient events. The more 
stringent requirements are imposed to ensure process variables, 
structures, systems, and components are maintained consistent with 
the safety analyses and licensing basis and to provide greater 
consistency with the ANO-1 TS and NUREG 1432.

Less Restrictive Changes

    (1) A note will be added that allows three (3) hours to perform 
the channel functional test on the control room radiation monitors 
without entering the associated Actions.
    The control room area radiation monitor is used to support 
mitigation of the consequences of an accident; however, it is not 
considered the initiator of any previously analyzed accident. Also, 
the addition of the Note to allow time for testing reduces the 
potential for initiation of a previously analyzed accident due to 
reduced potential for shutdowns and startups due to incomplete or 
missed surveillances. As such, the proposed revision to include an 
allowance for testing does not significantly increase the 
probability of any accident previously evaluated. This change does 
not result in any hardware changes, but does allow operation for a 
limited time with an inoperable monitor for the purposes of testing. 
Since the capability of the control room area radiation monitor to 
provide the required information continues to be verified, and the 
time allowed for inoperability for testing is short, the change will 
not reduce the capability of required equipment to mitigate the 
event. Also, the consequences of an event occurring during the 
proposed operation of the unit during the allowed inoperability for 
testing are the same as the consequences of an event occurring while 
operating under the current TS Actions. Therefore, this change does 
not involve a significant increase in the consequences of any 
accident previously evaluated.
    (2) This change will allow the control room boundary to be 
opened intermittently under administrative controls, and will allow 
both trains of the CREVS [control room emergency ventilation system] 
to be inoperable due to control room boundary inoperability for a 
period of 24 hours.
    Neither CREVS nor the control room boundary is the initiator of 
any accident analyzed in the SAR [Safety Analysis Report]. 
Therefore, this change does not result in a significant increase in 
the probability of an accident previously evaluated.
    The CREVS and the control room boundary are intended to provide 
a habitable environment for the control room operators in the event 
of an accident that results in the release of radioactivity to the 
environment. The allowance to open the control room boundary 
intermittently is acceptable, because of the administrative controls 
that will be implemented to ensure that the opening can be rapidly 
closed when the need for control room isolation is indicated, 
restoring the control room habitability envelope. Allowing both 
CREVS trains to be inoperable for 24 hours due to an inoperable 
control room boundary is acceptable because of the low probability 
of an accident requiring control room isolation during any given 24 
hour period, because entry into this condition is expected to be an 
infrequent occurrence, and because preplanned compensatory measures 
to protect the control room operators from potential hazards are 
implemented. Therefore, this change will not result in a significant 
increase in the probability [consequences] of an accident previously 
evaluated.
    (3) An allowance will be added to allow use of a ``simulated'' 
or ``actual'' signal when testing the automatic isolation feature of 
the control room air filtration system.
    The phrase ``actual or simulated'' in reference to the automatic 
initiation signal, has been added to the system functional test 
surveillance test description. This does not impose a requirement to 
create an ``actual'' signal, nor does it eliminate any restriction 
on producing an ``actual'' signal. The proposed change does not 
affect the procedures governing plant operations and the 
acceptability of creating these signals; it simply would allow such 
a signal to be utilized in evaluating the acceptance criteria for 
the system functional test requirements. Therefore, the change does 
not involve a significant increase in the probability of an accident 
previously evaluated. Since the function of the system functional 
test remains unaffected the change does not involve a significant 
increase in the consequences of an accident previously evaluated.
    (4) An allowance for the diesel fuel storage tanks to contain 
less than 22,500 gallons of fuel for up to 48 hours as long as the 
individual volume is greater than 17,446 gallons will be added. The 
lower value when summed with the contents of the other tank ensures 
six days of fuel oil is available. During the 48 hours, the diesel 
generator is capable of performing its intended function. There is a 
low probability that an event would occur for which the diesel 
generator would be required during this short period of time when 
the lower fuel oil volume is allowed.
    The AC Sources are used to support mitigation of the 
consequences of an accident and can be involved in the initiation of 
the accident analyzed in SAR. Equipment powered by the AC Sources, 
which may be considered as an initiator, continues to be assured of 
electrical power. The proposed increased restoration time involves 
parameters unrelated to initiating the failure of the AC Sources. As 
such the proposed time allowance for restoration of limited levels 
of readiness parameter degradation will not increase the probability 
of any accident previously evaluated. The proposed changes allow 
additional time for restoration of parameters that have been 
identified as not immediately affecting the capability of the power 
source to provide its required safety function. The identified 
parameters are capable of being replenished during operation of the 
diesel generators, and the short additional allowable action time 
continues to provide adequate assurance of operable required 
equipment. Therefore, this change does not involve a significant 
increase in the probability of or the consequences of any accident 
previously evaluated.
    (5) Seven days will be allowed to restore the stored diesel fuel 
oil total particulates to within the required limits prior to 
declaring the associated diesel inoperable.
    The testing of diesel generator fuel oil is not considered an 
initiator, or a mitigating factor, in any previously evaluated 
accident. The presence of particulates does not mean failure of the 
fuel oil to burn properly in the diesel engine. In addition, 
particulate concentration is unlikely to change significantly 
between surveillance intervals (31 days). Therefore, the change does 
not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (6) An allowance for the person who is satisfying the 
requirement of the radiation protection staff position and for the 
person filling the Shift Technical Advisor (STA) position to be 
vacant for not more than two hours in order to provide for 
unexpected absences is being added. This is consistent with the 
allowance permitted for the control room operator as reflected in 
existing TSs.

[[Page 68665]]

    This change does not result in any changes in hardware or 
methods of operation. The change allowing the absence of the STA or 
the radiation protection technician is not considered in the safety 
analysis, and cannot initiate or affect the mitigation of an 
accident in any way. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (7) The STA will be allowed to support the shift crew rather 
than only the shift supervisor. This provides more flexibility and 
does not dilute the function of the STA.
    This change does not result in any changes in hardware or 
methods of operation. The change in the support relationship between 
the STA and the control room staff is not considered in the safety 
analysis, and cannot initiate or affect the mitigation of an 
accident in any way. Therefore, this change does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    (8) The Occupational Radiation Exposure Report will be submitted 
by April 30 of each calendar year instead of prior to March 1.
    This change does not result in any changes in hardware or 
methods of operation. The change in date for submittal of ``after 
the fact'' information is not considered in the safety analysis, and 
cannot initiate or affect the mitigation of an accident in any way. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (9) An allowance is proposed that will revise the high radiation 
areas to include additional previously approved methods for 
implementation of alternatives to the ``control device'' or ``alarm 
signal'' requirements of 10 CFR [Part] 20. These alternatives 
provide adequate control of personnel in high radiation areas as 
evidenced by NRC issuance of NUREG-1432.
    The controls for access to a high radiation area are not 
considered as initiators, or as a mitigation factor, in any 
previously evaluated accident. Therefore, the change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (10) An allowance to require periodic testing of stored fuel for 
the particulates only is proposed.
    The testing of diesel generator fuel oil is not considered an 
initiator or a mitigating factor in any previously evaluated 
accident. Therefore, the change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    (11) The removal of the requirement to notify the Vice 
President, Operations ANO within 24 hours of violating a safety 
limit.
    Notification of the Vice President, Operations ANO when a safety 
limit is violated is not considered an initiator or a mitigating 
factor in any previously evaluated accident. Therefore, the change 
does not involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    (12) The Radioactive Effluent Release Report will be submitted 
by May 1 of each calendar year instead of prior to March 1.
    This change does not result in any changes in hardware or 
methods of operation. The change in date for submittal of ``after 
the fact'' information is not considered in the safety analysis, and 
cannot initiate or affect the mitigation of an accident in any way. 
Therefore, this change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    (13) A change that allows a 25% extension of the frequency in 
accordance with SR 4.0.2 for the integrated leak tests of each 
system outside containment that could contain highly radioactive 
fluids.
    The extension of the testing frequency, up to 25% of the test 
interval, is not considered an initiator or a mitigating factor in 
any previously evaluated accident. Therefore, the change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.
    (14) A change that allows the OSRC [Onsite Safety Review 
Committee] review of the desirability of maintaining a channel in 
the bypassed condition to be at or before the next regularly 
scheduled meeting.
    The proposed change is not considered an initiator or a 
mitigating factor in any previously evaluated accident. Therefore, 
the change does not involve a significant increase in the 
probability or consequences of any accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.

Administrative Changes

    The proposed changes do not necessitate a physical alteration of 
the plant (no new or different type of equipment will be installed) 
or changes in parameters governing normal plant operations. The 
proposed changes will not impose any different requirements.

Less Restrictive--Administrative Deletion of Requirements

    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operations. The proposed changes will not impose any different 
requirements and adequate control of the information will be 
maintained.

More Restrictive Changes

    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed changes do impose different requirements. 
However, these changes do not impact the safety analysis and 
licensing basis.

Less Restrictive Changes

    (1) A note will be added that allows three (3) hours to perform 
the channel functional test on the control room radiation monitors 
without entering the associated Actions.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will still ensure proper 
surveillances are required for the equipment considered in the 
safety analysis. Thus, this change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    (2) This change will allow the control room boundary to be 
opened intermittently under administrative controls, and will allow 
both trains of the control room ventilation system (CREVS) to be 
inoperable due to a control room boundary inoperability for a period 
of 24'hours.
    The proposed change does not necessitate a physical alteration 
of the unit (no new or different type of equipment will be 
installed) or changes in parameters governing normal unit operation. 
Prompt and appropriate compensatory actions will still be taken in 
the event of an accident. Thus, this change does not create the 
possibility of a new or different kind of accident from any 
previously evaluated.
    (3) An allowance will be added to allow use of a ``simulated'' 
or ``actual'' signal when testing the automatic isolation feature of 
the control room air filtration system.
    The possibility of a new or different kind of accident from any 
accident previously evaluated is not created because the proposed 
change introduces no new mode of plant operation and it does not 
involve physical modification to the plant.
    (4) An allowance for the diesel fuel storage tanks to contain 
less than 22,500 gallons of fuel for up to 48 hours as long as the 
individual volume is greater than 17,446 gallons will be added. The 
lower value when summed with the contents of the other tank ensures 
six days of fuel oil is available. During the 48 hours, the diesel 
generator is capable of performing its intended function. There is a 
low probability that an event would occur for which the diesel 
generator would be required during this short period of time when 
the lower fuel oil volume is allowed.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will continue to ensure operable 
safety equipment is available. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (5) Seven days will be allowed to restore the stored diesel fuel 
oil total particulates to within the required limits prior to 
declaring the associated diesel inoperable.
    No changes are proposed in the manipulation of the plant 
structures, systems, or components, or in the design of the plant 
structures, systems, or components. The presence of particulates 
does not mean failure of the fuel oil to burn properly in the diesel 
engine. In addition, particulate concentration is unlikely to change 
significantly between surveillance intervals (31 days). Therefore, 
the change does not

[[Page 68666]]

create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (6) An allowance for the person who is satisfying the 
requirement of the radiation protection staff position and for the 
person filling the Shift Technical Advisor (STA) position to be 
vacant for not more than two hours in order to provide for 
unexpected absences is proposed. This is consistent with the 
allowance permitted for the control room operator as reflected in 
existing TSs.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will impact only the STA and 
radiation protection staffing positions and does not directly impact 
the operation of the plant. Thus, this change does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    (7) The STA will be allowed to support the shift crew rather 
than only the shift supervisor. This provides more flexibility and 
does not dilute the function of the STA.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will impact only the support 
relationship the STA provides the control room staff and does not 
directly impact the operation of the plant. Thus, this change does 
not create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    (8) The Occupational Radiation Exposure Report will be submitted 
by April 30 of each calendar year instead of prior to March 1.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will impact only the administrative 
requirements for submittal of information and does not directly 
impact the operation of the plant. Thus, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (9) An allowance is proposed that will revise the high radiation 
areas to include additional previously approved methods for 
implementation of alternates to the ``control device'' or ``alarm 
signal'' requirements of 10 CFR [Part] 20. These alternatives 
provide adequate control of personnel in high radiation areas as 
evidenced by NRC issuance of NUREG-1432.
    No changes are proposed in the manipulation of the plant 
structures, systems, or components, or in the design of the plant 
structures, systems, or components. Therefore, the change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (10) An allowance to require periodic testing of stored fuel for 
the particulates only is proposed.
    No changes are proposed in the manipulation of the plant 
structures, systems, or components, or in the design of the plant 
structures, systems, or components. Therefore, the change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.
    (11) The removal of the requirement to notify the Vice 
President, Operations ANO within 24 hours of violating a safety 
limit.
    No changes are proposed that result in the manipulation or the 
design of plant structures, systems, or components. Therefore, the 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (12) The Radioactive Effluent Release Report will be submitted 
by May 1 of each calendar year instead of prior to March 1.
    The proposed change does not necessitate a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or changes in parameters governing normal plant 
operation. The proposed change will impact only the administrative 
requirements for submittal of information and does not directly 
impact the operation of the plant. Thus, this change does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    (13) A change that allows a 25% extension of the frequency in 
accordance with SR 4.0.2 for the integrated leak tests of each 
system outside containment that could contain highly radioactive 
fluids.
    No changes are proposed that result in the manipulation or the 
design of plant structures, systems, or components. Therefore, the 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    (14) A change that allows the OSRC review of the desirability of 
maintaining a channel in the bypassed condition to be at or before 
the next regularly scheduled meeting.
    No changes are proposed that result in the manipulation or the 
design of plant structures, systems, or components. Therefore, the 
change does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    Response: No.

Administrative Changes

    The proposed changes will not reduce the margin of safety 
because they have no impact on any safety analysis assumptions. The 
changes are administrative in nature.

Less Restrictive--Administrative Deletion of Requirements

    The proposed changes will not reduce a margin of safety because 
they have no impact on any safety analysis assumptions. In addition, 
the requirements to be transposed from the TSs to other license 
basis documents, which are under licensee control, are the same as 
the exiting TSs. The documents containing the relocated requirements 
will be maintained using the provisions of applicable regulatory 
requirements.

More Restrictive Changes

    The imposition of more stringent requirements prevents a 
reduction in the margin of plant safety by:
    (a) increasing the scope of the specification to include 
additional plant equipment,
    (b) providing additional actions,
    (c) decreasing restoration times, or
    (d) imposing new surveillances.
    The changes are consistent with the safety analysis and 
licensing basis.

Less Restrictive Changes

    (1) A note will be added that allows three (3) hours to perform 
the channel functional test on the control room radiation monitors 
without entering the associated Actions.
    The margin of safety for the control room area radiation monitor 
is based on availability and capability of the instrumentation to 
provide the required information to the operator. The frequency is 
based on unit operating experience that demonstrates channel failure 
is rare, and on the use of less formal but more frequent checks of 
channels during normal operational use of the displays associated 
with the required channels. Therefore, the availability and 
capability of the control room area radiation monitor continues to 
be assured by the proposed Surveillance Requirements and this change 
does not involve a significant reduction in a margin of safety.
    (2) This change will allow the control room boundary to be 
opened intermittently under administrative controls, and will allow 
both trains of the control room ventilation system (CREVS) to be 
inoperable due to control room boundary inoperability for a period 
of 24 hours.
    This change does not involve a significant reduction in a margin 
of safety since: (1) Administrative controls will be in place to 
ensure that an open control room boundary can be rapidly closed when 
a need for control room isolation is indicated; and (2) an 
inoperable control room boundary that renders both trains of CREVS 
inoperable is an infrequent occurrence, the probability of an 
accident requiring control room isolation during any given 24 hour 
period is low, and preplanned compensatory measures to protect the 
control room operators from potential hazards are implemented.
    (3) An allowance will be added to use a simulated or actual 
signal when testing the automatic isolation feature of the control 
room air filtration system.
    Use of an actual signal instead of the existing requirement 
which limits use to a simulated signal, will not affect the 
performance of the surveillance test. OPERABILITY is adequately 
demonstrated in either case since the system itself can not 
discriminate between ``actual'' or ``simulated'' signals. Therefore, 
the change does not involve a significant reduction in a margin of 
safety.
    (4) An allowance for the diesel fuel storage tanks to contain 
less than 22,500 gallons of fuel for up to 48 hours as long as the

[[Page 68667]]

individual volume is greater than 17,446 gallons. The lower value 
when summed with the contents of the other tank ensures six days of 
fuel oil is available. During the 48 hours, the diesel generator is 
capable of performing its intended function. There is a low 
probability that an event would occur for which the diesel generator 
would be required during this short period of time when the lower 
fuel oil volume is allowed.
    The parameter limits provide substantial margin to the parameter 
values that would be absolutely necessary for diesel generator 
operability. When the parameters are less than their limits this 
margin is reduced. However, the availability of AC Sources continues 
to be assured since the allowed time for parameters to be less than 
their limits is short and the allowed levels for the parameters are 
adequate to provide the immediately needed power availability. 
Further, the parameters can be restored to within limits during the 
proposed time provided should they be required. Therefore, this 
change does not result in a significant reduction in [a] margin of 
safety.
    (5) Seven days will be allowed to restore the stored diesel fuel 
oil total particulates to within the required limits prior to 
declaring the associated diesel inoperable.
    The proposed change allows the stored diesel fuel oil total 
particulates to be outside the required limits for seven days before 
declaring the associated diesel inoperable. The presence of 
particulates does not mean failure of the fuel oil to burn properly 
in the diesel engine. In addition, particulate concentration is 
unlikely to change significantly between surveillance intervals (31 
days). The seven day allowance provides an appropriate backstop to 
ensure the particulate level is restored to within limits in a 
reasonable time period. Since the diesel is still capable of 
performing its function the margin of safety is not reduced.
    (6) An allowance for the person who is satisfying the 
requirement of the radiation protection staff position and for the 
person filling the Shift Technical Advisor (STA) position to be 
vacant for not more than two hours in order to provide for 
unexpected absences is proposed. This is consistent with the 
allowance permitted for the control room operator as reflected in 
existing TSs.
    The margin of safety is not dependent on the presence of the STA 
or the radiation protection technician. Therefore, this change does 
not involve a significant reduction in a margin of safety.
    (7) The STA will be allowed to support the shift crew rather 
than only the shift supervisor. This provides more flexibility and 
does not dilute the function of the STA.
    The margin of safety is not dependent upon who the STA supports. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.
    (8) The Occupational Radiation Exposure Report will be submitted 
by April 30 of each calendar year instead of prior to March 1.
    The margin of safety is not dependent on the submittal of 
information. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    (9) An allowance is proposed that will revise the high radiation 
areas to include additional previously approved methods for 
implementation of alternatives to the ``control device'' or ``alarm 
signal'' requirements of 10 CFR [Part] 20. These alternatives 
provide adequate control of personnel in high radiation areas as 
evidenced by NRC issuance of NUREG-1432.
    The requirements for control of high radiation areas provide for 
the use of alternates to the ``control device'' or ``alarm signal'' 
requirements of 10 CFR 20.1601. This change provides such 
alternative methods for controlling access. These methods and 
additional administrative requirements have been determined to 
provide adequate controls to prevent unauthorized and inadvertent 
access to such areas. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    (10) An allowance to require periodic testing of stored fuel for 
the particulates only is proposed.
    The testing of stored diesel generator fuel oil is revised to 
require the periodic testing of the stored fuel oil only for 
particulates (replacing the periodic testing per ASTM-D975) once 
every 31 days. The change reflects industry-standard acceptable DG 
fuel oil testing programs. Over the storage life of ANO-2 DG fuel 
oil, the properties tested by ASTM-D975 are not expected to change 
and performing these tests once on the new fuel oil provides 
adequate assurance of the proper initial quality of fuel oil. The 
periodic testing for particulates monitors a parameter that reflects 
degradation of fuel oil and can be trended to provide increased 
confidence that the stored DG fuel oil will support DG operability. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.
    (11) The removal of the requirement to notify the Vice 
President, Operations ANO within 24 hours of violating a safety 
limit.
    The margin of safety is not dependent upon notification of the 
Vice President, Operations ANO upon the violation of a TS safety 
limit. Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.
    (12) The Radioactive Effluent Release Report will be submitted 
by May 1 of each calendar year instead of prior to March 1.
    The margin of safety is not dependent on the submittal of 
information. Therefore, this change does not involve a significant 
reduction in a margin of safety.
    (13) A change that allows a 25% extension of the frequency in 
accordance with SR 4.0.2 for the integrated leak tests of each 
system outside containment that could contain highly radioactive 
fluids.
    The proposed allowance allows a possible increase in performance 
interval. However, the test will still be performed at reasonable 
intervals to ensure the intent of the surveillance is maintained. 
Therefore, this change does not involve a significant reduction in a 
margin of safety.
    (14) A change that allows the OSRC review of the desirability of 
maintaining a channel in the bypassed condition to be at or before 
the next regularly scheduled meeting.
    The proposed change allows the OSRC review to occur earlier than 
previously required if an OSRC meeting is called before the next 
regularly scheduled meeting. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: October 22, 2003.
    Description of amendment request: The licensee proposes to change 
the existing pressure/temperature limits (P/T) from 16 to 32 effective 
full power years (EFPY). In addition, the maximum heatup rate will be 
changed to 60 [deg]F per hour and the maximum cooldown rate to 100 
[deg]F per hour for all reactor coolant system temperatures. For 
inservice hydrostatic pressure and leak testing, the maximum heatup and 
cooldown rates will be changed to 60 [deg]F and 100 [deg]F, 
respectively.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously analyzed?
    Response: No.
    The probability of occurrence of an accident previously 
evaluated for Waterford 3 is not altered by the proposed amendment 
to the TSs [Technical Specifications]. The accidents currently 
analyzed in the Waterford 3 Final Safety Analysis Report (FSAR) 
remain the same considering the results of the proposed changes to 
the P/T limits and the LTOP [low temperature overpressure] enable 
temperature. The new P/T and LTOP enable temperature limits were 
based on the NRC [Nuclear Regulatory Commission] accepted 
methodologies along with the ASME [American Society of Mechanical 
Engineers] Code [Boiler and Pressure Vessel Code] alternatives. The 
proposed changes do not impact the integrity of the reactor coolant 
pressure boundary (RCPB) (i.e., there is no change to the

[[Page 68668]]

operating pressure, materials, loadings, etc.). The proposed change 
does not affect the probability nor consequences of any design basis 
accident (DBA). The proposed P/T limit curves, maximum heatup and 
cooldown rates, and the LTOP enable temperature are not considered 
to be an initiator or contributor to any accident currently 
evaluated in the Waterford 3 FSAR. The new limits ensure the long 
term integrity of the RCPB.
    Fracture toughness test data are obtained from material 
specimens contained in capsules that are periodically withdrawn from 
the reactor vessel. These data permit determination of the 
conditions under which the vessel can be operated with adequate 
safety margins against non-ductile fracture throughout its service 
life. During the spring 2002 Waterford 3 refueling outage, a reactor 
vessel specimen capsule was withdrawn and analyzed to predict the 
fracture toughness requirements using projected neutron fluence 
calculations. For each analyzed transient and steady state 
condition, the allowable pressure is determined as a function of 
reactor coolant temperature considering postulated flaws in the 
reactor vessel beltline, inlet nozzle, outlet nozzle, and closure 
head.
    The predicted radiation induces ``RTNDT was calculated using the 
respective reactor vessel beltline materials copper and nickel 
contents and neutron fluence applicable to 32 EFPY including an 
estimated increase in flux due to proposed power uprates. The RTNDT 
and, in turn, the operating limits for Waterford 3 were adjusted to 
account for the effects of irradiation on the fracture toughness of 
the reactor vessel materials. Therefore, new operating limits will 
be established which are represented in the revised operating curves 
for heatup/criticality, cooldown, and inservice hydrostatic testing 
contained in the TSs.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes to the P/T and LTOP enable temperature will 
not create a new accident scenario. The requirements to have P/T 
limits and LTOP protection are part of the licensing basis for 
Waterford 3. The approach used to develop the new P/T limits and 
LTOP enable temperature meets NRC and ASME regulations and 
guidelines. The data analysis for the vessel specimen removed during 
the last Waterford 3 refueling outage confirms that the vessel 
materials are responding as predicted.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The existing P/T curves and LTOP enable temperature in the TSs 
are reaching their expiration period for the number of years at 
effective full power operation. The revision of the P/T limits and 
curves will ensure that Waterford 3 continues to operate within the 
operating margins allowed by 10 CFR 50.60 and the ASME Code. The 
material properties used in the analysis are based on results 
established through Westinghouse material reports for copper and 
nickel content. The application of ASME Code Case N-641 presents 
alternative procedures for calculating P/T and LTOP temperatures in 
lieu of that established for ASME Section XI, Appendix G-2215. The 
Code alternative allows certain assumptions to be conservatively 
reduced. However, the procedures allowed by Code Case N-641 still 
provide significant conservatism and ensure an adequate margin of 
safety in the development of P/T operating and pressure test limits 
to prevent non-ductile fractures.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: N.S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick 
Generating Station, Unit Nos. 1 and 2, Montgomery County, Pennsylvania

    Date of amendment request: September 8, 2003.
    Description of amendment request: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of Title 10 of the Code of Federal 
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition 
for Operation (LCO) 3.0.4, exceptions in individual TSs, would be 
eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect 
the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated September 8, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions

[[Page 68669]]

of the TS. The TS allow operation of the plant without the full 
complement of equipment through the conditions for not meeting the 
TS LCO. The risk associated with this allowance is managed by the 
imposition of required actions that must be performed within the 
prescribed completion times. The net effect of being in a TS 
condition on the margin of safety is not considered significant. The 
proposed change does not alter the required actions or completion 
times of the TS. The proposed change allows TS conditions to be 
entered, and the associated required actions and completion times to 
be used in new circumstances. This use is predicated upon the 
licensee's performance of a risk assessment and the management of 
plant risk. The change also eliminates current allowances for 
utilizing required actions and completion times in similar 
circumstances, without assessing and managing risk. The net change 
to the margin of safety is insignificant. Therefore, this change 
does not involve a significant reduction in a margin of safety.

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York 
and Lancaster Counties, Pennsylvania

    Date of application for amendments: September 26, 2003.
    Description of amendment request: The proposed amendment would 
modify the fire protection plan (FPP). The change to the FPP would 
allow converting the existing carbon dioxide (CO2) fire suppression 
systems, located in the cable spreading room (CSR) and each of the four 
emergency diesel generator rooms, from automatic to manual actuation 
systems.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed activity involves changing the actuation of the 
carbon dioxide (CO2) fire suppression systems from 
automatic to manual. With the exception of the Emergency Diesel 
Generator (EDG) CO2 system itself, the proposed activity 
does not result in any physical changes to safety-related 
structures, systems, or components (SSCs), or the manner in which 
safety-related SSCs are operated, maintained, modified, tested, or 
inspected. The EDG CO2 system is safety related due to a 
potential common mode effect on all four EDGs in the event of a 
seismic event. Eliminating the automatic actuation function of the 
EDG CO2 system will thereby eliminate a potential common 
mode effect on the EDGs. The proposed activity does not degrade the 
performance or increase the challenges of any safety-related SSCs 
assumed to function in the accident analysis. As a result, the 
proposed activity does not introduce any new accident initiators. In 
addition, fires are not an accident that is previously evaluated. 
Regardless, the proposed activity does not change the probability of 
a fire occurring since fire ignition frequency is independent of the 
method of fire suppression in the room. The consequences of the 
proposed activity are bounded by the fire safe shutdown analysis, 
which assumes fire damage throughout the affected fire area. The 
fire safe shutdown analysis for each of the areas addressed by the 
proposed activity demonstrates that safe shutdown can be 
accomplished assuming that no fire suppression is available. In 
addition, the removal of the automatic discharge capability of the 
CO2 system in each of the EDG rooms significantly reduces 
the potential for an inadvertent discharge to shutdown the EDG if 
needed for non-fire accident conditions. Similarly, removal of the 
automatic discharge feature in the CSR significantly reduces the 
potential for an inadvertent discharge that would require (by 
procedure) immediate shutdown of both units, and the potential 
migration of CO2 into the main control room or other 
areas. In the future, CO2 discharge will only occur as a 
deliberate action to the most extreme fires, as one element of an 
overall graded approach to fire fighting in the affected areas.
    Therefore, changing the actuation of the CO2 fire 
suppression systems from automatic to manual does not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed activity involves changing the actuation of the 
CO2 fire suppression systems from automatic to manual. 
With the exception of the Emergency Diesel Generator (EDG) 
CO2 system itself, the proposed activity does not result 
in any physical changes to safety-related structures, systems, or 
components (SSCs), or the manner in which safety-related SSCs are 
operated, maintained, modified, tested, or inspected. The proposed 
activity does not degrade the performance or increase the challenges 
of any safety-related SSCs assumed to function in the accident 
analysis. As a result, the proposed activity does not introduce nor 
increase the number of failure mechanisms of a new or different type 
than those previously evaluated. The fire safe shutdown analysis 
assumes fire damage throughout the area consistent with a complete 
lack of fire suppression capability. The elimination of the 
potential for inadvertent actuation accomplished by changing the 
CO2 systems from automatic to manual prevents the 
CO2 systems from creating a challenge to existing 
accidents.
    Therefore, changing the actuation of the CO2 fire 
suppression system from automatic to manual does not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed activity involves changing the actuation of the 
CO2 fire suppression systems from automatic to manual. 
With the exception of the Emergency Diesel Generator (EDG) 
CO2 system itself, the proposed activity does not result 
in any physical changes to safety-related structures, systems, or 
components (SSCs), or the manner in which safety-related SSCs are 
operated, maintained, modified, tested, or inspected. The proposed 
activity does not degrade the performance or increase the challenges 
of any safety-related SSCs assumed to function in the accident 
analysis. The proposed activity does not impact plant safety since 
the conclusions of the fire safe shutdown analysis remain unchanged.
    Therefore, changing the actuation of the CO2 fire 
suppression system from automatic to manual 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.
    Attorney for Licensee: Mr. Edward Cullen, Vice President and 
General Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
S23-1, Philadelphia, PA 19101.
    NRC Section Chief: James W. Clifford.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: August 25, 2003.
    Description of amendment request: The proposed amendment would 
allow extension of the current Emergency Diesel Generator (EDG) 
Technical Specifications allowed outage time (AOT) from 72 hours to a 
period of 14 days. This proposal would be supported by permanently 
installing a non-safety-related supplemental emergency power system 
(SEPS).
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:


[[Page 68670]]


    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    The proposed changes do not involve a change in the operational 
limits or physical design of the electrical power systems, 
particularly the emergency power systems. The proposed changes do 
not change the function or operation of plant equipment or affect 
the response of that equipment if called upon to operate. The 
proposed AOT extensions to allow for additional operational 
flexibility will not cause a significant increase in the probability 
or consequences of an accident previously evaluated. In actuality, 
the installation of the SEPS will have an overall net reduction in 
core damage frequency. The AOT extensions will lessen the burden of 
time pressure to quickly determine the cause of failure and perform 
corrective actions without needing to place the plant in a transient 
to shutdown because of a short allotted AOT.
    A Probabilistic Risk Assessment (PRA) has been performed to 
quantitatively assess the risk impact of an increase in the Allowed 
Outage Time. The proposed change results in a significant decrease 
in core damage frequency (CDF). Large Early Release Frequency (LERF) 
is dominated by containment bypass and containment isolation 
failures and remains relatively unchanged by the addition the SEPS 
combined with a 14-day AOT.
    Based on the above, the proposed changes will not significantly 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed changes do not involve a change in the operational 
limits or physical design of the electrical power systems, 
particularly the emergency power systems. The proposed changes do 
not change the function or operation of plant equipment or introduce 
any new failure mechanisms. The SEPS and interfacing components with 
the safety-related busses have been designed to ensure independence 
and separation, particularly during faulted conditions. As such, no 
new failure modes are being introduced. The plant equipment will 
continue to respond per the design and analyses and there will not 
be a malfunction of a new or different type introduced by the 
proposed changes.
    The proposed amendment extends the Allowed Outage Times for 
restoring an inoperable EDG to OPERABLE status and extends the 
period for operability verification of redundant features to allow 
for minor repair prior to placing the plant in a shutdown transient. 
The proposed amendment will not result in changes to the type of 
corrective or preventive maintenance activities associated with the 
EDGs. Plant operating procedures and the procedures used to respond 
to abnormal or emergency conditions will be enhanced with the option 
to use the SEPS when deemed necessary. Assumptions made in the 
safety analysis related to EDG availability will also remain 
unchanged. Performance of certain maintenance activities at power 
requires an evaluation to assure plant safety is maintained or 
enhanced, which would include evaluation for new or different plant 
conditions. Therefore, the proposed change will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed changes do not involve a change in the operational 
limits. The proposed changes do not change the function or operation 
of plant equipment or affect the response of that equipment if it is 
called upon to operate. The performance capability of the emergency 
diesel generators will not be affected. Installation of the SEPS 
will have an overall net reduction in core damage frequency. 
Emergency diesel generator reliability and availability will be 
improved by implementation of the proposed changes. In addition, 
administrative controls will ensure there are adequate compensatory 
measures that can be and will be taken during extended EDG 
maintenance activities to reduce overall risk. The results of the 
PRA performed to quantitatively assess the risk impact of an 
increase in the Allowed Outage Time indicate the proposed change 
results in a significant decrease in core damage frequency (CDF) by 
up to 30 percent. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Esquire, Florida Power & Light 
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.

FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No. 
1, Rockingham County, New Hampshire

    Date of amendment request: October 6, 2003.
    Description of amendment request: The proposed amendment would 
revise the Seabrook Station licensing basis to implement the 
alternative source term (AST) methodology of Regulatory Guide (RG) 
1.183 through reanalysis of the radiological consequences of a number 
of the Updated Final Safety Analysis Report Chapter 15 accidents. 
Further, having revised the licensing basis, the amendment would also 
revise the definition of dose equivalent I-131 in Technical 
Specifications Section 1.12.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) Section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. The proposed changes do not involve a significant increase in 
the probability or consequences of an accident previously evaluated.
    Alternative source term calculations have been performed that 
demonstrate the dose consequences remain below limits specified in 
NRC [Nuclear Regulatory Commission] Regulatory Guide 1.183 (July 
2000) and 10CFR50.67. The proposed change does not modify the 
physical design or operation of the plant. The use of AST changes 
only the regulatory assumptions regarding the analytical treatment 
of the design basis accidents and has no direct effect on the 
probability of the accident. AST has been utilized in the analysis 
of the limiting design basis accidents listed above. The results of 
the analyses, which include the proposed change to the Technical 
Specifications, demonstrate that the dose consequences of these 
limiting events are all within the regulatory limits. The proposed 
Technical Specification change to the definition of dose equivalent 
I-131 is consistent with the implementation of AST and the 
requirements of RG 1.183 (July 2000).
    Therefore, the proposed change does not involve a significant 
increase the probability or consequences of an accident previously 
evaluated.
    2. The proposed changes do not create the possibility of a new 
or different kind of accident from any previously evaluated.
    The proposed change does not affect any plant structures, 
systems, or components. The operation of plant systems and equipment 
will not be affected by this proposed change. The alternative source 
term and the dose equivalent I-131 definition change do not have the 
capability to initiate accidents. Therefore, the proposed change 
does not create the possibility of a new or different kind of 
accident from any accident previously evaluated.
    3. The proposed changes do not involve a significant reduction 
in the margin of safety.
    The proposed implementation of the alternative source term 
methodology is consistent with NRC RG 1.183 (July 2000). The 
Technical Specification change to the definition of dose equivalent 
I131 is consistent with the implementation of AST and the 
requirements of RG 1.183 (July 2000). Conservative methodologies, 
per the guidance of RG 1.183 (July 2000), have been used in 
performing the accident analyses. The radiological consequences of 
these accidents are all within the regulatory acceptance criteria 
associated with use of the alternative source term methodology.
    The proposed changes continue to ensure that the doses at the 
exclusion area and low population zone boundaries and in the Control 
Room are within the corresponding regulatory limits of RG 1.183 
(July 2000) and 10CFR50.67. The margin of safety for the 
radiological consequences of these accidents is considered to be 
that provided by meeting the applicable regulatory limits, which are 
set at or below the 10CFR50.67 limits. An acceptable margin of 
safety is inherent in these limits.

[[Page 68671]]

    Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Esquire, Florida Power & Light 
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment requests: September 26, 2003.
    Description of amendment requests: The proposed change allows entry 
into a mode or other specified condition in the applicability of a 
Technical Specification (TS), while in a condition statement and the 
associated required actions of the TS, provided the licensee performs a 
risk assessment and manages risk consistent with the program in place 
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting 
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would 
be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to 
reflect the LCO 3.0.4 allowance.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). 
The licensee affirmed the applicability of the following NSHC 
determination in its application dated September 26, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

    Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident 
Previously Evaluated.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    Criterion 2--The Proposed Change Does Not Create the Possibility 
of a New or Different Kind of Accident from any Previously 
Evaluated.
    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.
    Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in a Margin of Safety.
    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. The TS 
allow operation of the plant without the full complement of 
equipment through the conditions for not meeting the TS LCO. The 
risk associated with this allowance is managed by the imposition of 
required actions that must be performed within the prescribed 
completion times. The net effect of being in a TS condition on the 
margin of safety is not considered significant. The proposed change 
does not alter the required actions or completion times of the TS. 
The proposed change allows TS conditions to be entered, and the 
associated required actions and completion times to be used in new 
circumstances. This use is predicated upon the licensee's 
performance of a risk assessment and the management of plant risk. 
The change also eliminates current allowances for utilizing required 
actions and completion times in similar circumstances, without 
assessing and managing risk. The net change to the margin of safety 
is insignificant. Therefore, this change does not involve a 
significant reduction in a margin of safety.
    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.

    Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and 
Electric Company, P.O. Box 7442, San Francisco, California 94120.
    NRC Section Chief: Stephen Dembek.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: October 23, 2003.
    Description of amendment request: The proposed amendment would 
delete the surveillance requirements associated with the Emergency 
Diesel Generator lockout features.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes to the Technical Specifications (TS) 3/
4.8.1.1, AC Sources--Operating, would delete an unnecessary 
surveillance. The probability of occurrence or the consequences for 
an accident or malfunction of equipment is not increased by the 
proposed changes. In addition, the proposed changes do not alter the 
way any structure, system or component (SSC) functions, do not 
modify the manner in which the plant is operated, and do not 
significantly alter equipment out-of-service time. Deleting the 
surveillance of equipment protection does not change the probability 
or consequences of any accident and dose consequences are 
unaffected. No changes to the design of structures, systems, or 
components (SSC) are made and there are no effects on accident 
mitigation.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The possibility of a new or different kind of accident from any 
accident or malfunction in the Hope Creek Updated Final Safety 
Analysis Report (UFSAR) is not created. The Emergency Diesel 
Generators are accident mitigation equipment and cannot initiate an 
accident. The proposed changes to the TS do not change the design 
function or operation of any SSCs. The TS, as amended, would

[[Page 68672]]

continue to provide assurance of EDG operability.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed changes are procedural in nature and make no 
changes that affect the ability of plant SSCs to perform their 
design basis accident functions. In addition, the proposed changes 
do not change the margin of safety since no SSCs are changed. The 
results of accident analysis remain unchanged by the proposed 
changes to TS.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station, 
Salem County, New Jersey

    Date of amendment request: October 24, 2003.
    Description of amendment request: The proposed change to Technical 
Specifications will revise surveillance requirements associated with 
reactor protection system instrumentation, control rod block 
instrumentation, source range monitors, and power distribution limits, 
to minimize unnecessary testing.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment would revise the Technical Specification 
(TS) Surveillance Requirements (SRs) for certain Reactor Protection 
System and Control Rod Block Instrumentation, the source range 
monitors and power distribution limits, consistent with NUREG-1433, 
``Standard Technical Specifications (STS) General Electric Plants, 
BWR [Boiling Water Reactor]/4,'' Revision 2. No changes are being 
made to any instrumentation setpoints or plant components. The 
revised SRs continue to assure that the necessary quality of systems 
and components is maintained, that facility operation will be within 
safety limits, and that the Limiting Conditions for Operation will 
be met.
    Since the proposed changes do not affect any accident initiator 
and since the associated equipment will remain capable of performing 
its design function, the proposed change does not involve a 
significant increase in the probability or radiological consequences 
of an accident previously evaluated.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed changes do not change the design function or 
operation of any plant equipment. No new failure mechanisms, 
malfunctions, or accident initiators are being introduced by the 
proposed changes. Therefore, the proposed changes do not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    No changes are being made to any plant instrumentation setpoints 
or to the required level of redundancy. No changes are being made to 
any power distribution limits.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: September 12, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) Sections 1.1, 3.7.10, 3.7.12, 
3.7.13, 3.7.14, 3.9.4, 5.5.2, and 5.5.10, and the associated Bases 
Sections to implement an alternate source term at North Anna Power 
Station, Units 1 and 2. The proposed changes would implement NUREG-
1465, ``Accident Source Terms for Light-Water Nuclear Power Plants,'' 
dated February 1995, as the design-basis source term, achieve a 
consistent design basis for all accident dose assessments, increase 
operational flexibility by allowing for increased emergency core 
cooling system leakage and unfiltered control room in-leakage, and 
eliminate the surveillance requirement to test the bottled air flow 
rate.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
    We have reviewed the proposed TS changes relative to the 
requirements of 10 CFR 50.92 and determined that a significant hazards 
consideration is not involved. Specifically, operation of North Anna 
Power Station with the proposed changes will not:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed amendment does not involve a significant increase 
in the probability or consequence of an accident previously 
analyzed. The North Anna MCR/ESGR [main control room/emergency 
switchgear room] EVS [emergency ventilation system], PREACS [pump 
room exhaust air cleanup system], and MCE [MCR]/ESGR Bottled Air 
systems only function following the initiation of a design basis 
radiological accident. Therefore, the changes to these 
specifications, the definition of currently irradiated fuel, and the 
increase [of] the depressurization time of [the] containment 
following a design basis LOCA [loss-of-coolant accident] will not 
increase the probability of any previously analyzed accident. These 
systems are not initiators of any design bases accident.
    Revised dose calculations, which take into account the changes 
proposed by this [these] amendment[s] and the use of the alternative 
source term[,] have been performed for the North Anna design basis 
radiological accidents. The results of these revised calculations 
indicate that public and control room doses will not exceed the 
limits specified in 10 CFR 50.67 and Regulatory Guide 1.183. There 
is not a significant increase in predicted dose consequences for any 
of the analyzed accidents. Therefore, the proposed changes do not 
involve a significant increase in the consequences of any previously 
analyzed accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The implementation of the proposed changes does not create the 
possibility of an accident of a different type than was previously 
evaluated in the UFSAR [Updated Final Safety Analysis Report]. 
Although the proposed changes could affect the operation of the MRC 
[MCR]/ESGR EVS following a design basis radiological accident, none 
of these changes can initiate a new or different kind of accident 
since they are only related to system capabilities that provide 
protection from accidents that have already occurred. These changes 
do not alter the nature of events postulated in the UFSAR nor do 
they introduce any unique precursor mechanisms.

[[Page 68673]]

Therefore, the proposed changes do not create the possibility of a 
new or different kind of accident from those previously analyzed.
    3. Involve a significant reduction in the margin of safety.
    The implementation of the proposed changes does not reduce the 
margin of safety. The proposed changes for the MCR/ESGR EVS, PREACS, 
and MCE [MCR]/ESGR Bottled Air System do not affect the ability of 
these systems to perform their intended safety functions to maintain 
dose less than the required limits during design basis radiological 
events. The revised dose calculations also indicate that the change 
to the containment depressurization times will continue to maintain 
the dose to the public and control room operators less than the 
required limits.
    The radiological analysis results, when compared with the 
revised TEDE [total effective dose equivalent] acceptance criteria, 
meet the applicable limits. These acceptance criteria have been 
developed for application to analyses performed with alternative 
source terms. These acceptance criteria have been developed for the 
purpose of use in design basis accident analyses such that meeting 
the stated limits demonstrates adequate protection of public health 
and safety. It is thus concluded that the margin of safety will not 
be reduced by the implementation of the changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel, 
Dominion Resources Services, Inc., Millstone Power Station, Building 
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
    NRC Section Chief: John A. Nakoski.

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, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by e-mail to [email protected].

AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek 
Nuclear Generating Station, Ocean County, New Jersey

    Date of application for amendment: June 2, 2003.
    Brief description of amendment: The amendment revised the Technical 
Specifications, Sections 3.7.B.1 and 3.7.C.2. Section 3.7.B.1 required 
that the reactor may remain in operation ``for a period not to exceed 7 
days in any 30 day period if a startup transformer is out of service.'' 
Section 3.7.C.2 required that the reactor may be in operation ``for a 
period not to exceed 7 days in any 30 day period if a diesel generator 
is out of service.'' The amendment deleted the phrase ``in any 30 day 
period'' from these two sections.
    Date of Issuance: November 24, 2003.
    Effective date: November 24, 2003 and shall be implemented within 
30 days of issuance.
    Amendment No.: 239.
    Facility Operating License No. DPR-16: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40709).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated November 24, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear 
Generating Unit No. 2, Westchester County, New York

    Date of application for amendment: March 27, 2002, as supplemented 
on May 30, July 10, October 10, October 28, November 26, and December 
18, 2002, and on January 6, January 27, February 26, April 8, May 19, 
June 23, June 26, July 15, August 6, September 11, October 8, and 
October 14, 2003.
    Brief description of amendment: This amendment converts the current 
Technical Specifications (TS) to a set of Improved TS based on NUREG-
1431, Revision 2, ``Standard Technical Specifications for Westinghouse 
Plants,'' Revision 2, dated June 2001.
    Date of issuance: November 21, 2003.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 238.
    Facility Operating License No. DPR-26: Amendment replaced the 
current Technical Specifications (TSs) with the Improved TSs in their 
entirety and revised the license.
    Date of initial notice in Federal Register: September 26, 2003 (68 
FR 55660).
    The supplemental letters that were received subsequent to the 
issuance of the Federal Register notice provided clarifying information 
that did not change the no significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 21, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear 
Power Station, Plymouth County, Massachusetts

    Date of application for amendment: March 19, 2003.
    Brief description of amendment: This amendment deletes Technical 
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby 
eliminates the requirements to have and maintain the post accident 
sampling system at the Pilgrim Nuclear Power Station.

[[Page 68674]]

    Date of issuance: November 14, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 204.
    Facility Operating License No. DPR-35: Amendment revised the TSs.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34663).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 14, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, 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

    Date of application for amendments: October 16, 2002, as 
supplemented by letters dated June 20, 2003 and October 14, 2003.
    Brief description of amendments: The amendments revise the 
completion time of Required Action A.1 of Technical Specification 
3.8.7, ``Inverters-Operating,'' from the current 24 hours to 7 days for 
one inoperable instrument bus inverter. This provides greater 
operational flexibility for online maintenance of an instrument bus 
inverter with the potential to reduce the duration of refueling 
outages.
    Date of issuance: November 19, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 135/135, 129/129.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75874).
    The supplemental letters contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 2003. No significant hazards 
consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: October 24, 2002 and as 
supplemented by letter dated June 20, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification 5.5.13, ``Primary Containment Leakage Rate Testing 
Program,'' to reflect a one-time deferral of the primary containment 
Type A test to no later than June 13, 2009 for Unit 1 and no later than 
December 7, 2008 for Unit 2.
    Date of issuance: November 19, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 162, 148.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: December 10, 2002 (67 
FR 75876).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 19, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle 
County Station, Units 1 and 2, LaSalle County, Illinois

    Date of application for amendments: May 19, 2003.
    Brief description of amendments: The amendments revise Appendix A, 
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the change will decrease the frequency 
associated with TS Surveillance Requirement (SR) 3.7.7.1 for Turbine 
Bypass Valve (BPV) testing from 7 to 31 days. The change is consistent 
with the testing frequency contained in NUREG-1434, ``Standard 
Technical Specifications General Electric Plants, BWR/6,'' Revision 2, 
dated June 2001, for BPV testing. The 7-day frequency associated with 
SR 3.7.7.1 was established in the LaSalle County Station (LSCS) TS 
during conversion to improved Standard Technical Specifications (STS) 
format due to the testing frequency contained in the LSCS custom TS and 
the difficulties experienced with other Electro-Hydraulic Control (EHC) 
system valves to consistently pass their surveillance tests. LSCS has 
recently re-evaluated the performance of these valves and has 
determined that the current performance of these valves supports 
decreasing the testing frequency of the BPVs from 7 to 31 days.
    Date of issuance: November 13, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment Nos.: 163/148.
    Facility Operating License Nos. NPF-11 and NPF-18: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: June 24, 2003 (68 FR 
37577).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 13, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin

    Date of application for amendment: December 19, 2002, as 
supplemented July 25, 2003.
    Brief description of amendment: The proposed amendment would revise 
the Kewaunee technical specifications to change the Nuclear Regulatory 
Commission reporting requirements for the discovery of defective or 
degraded steam generator tubes so that the requirements are aligned 
with 10 CFR 50.72 and 10 CFR 50.73.
    Date of issuance: November 20, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment No.: 171.
    Facility Operating License No. DPR-43: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2807).
    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 20, 2003.
    No significant hazards consideration comments received: No.

Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County, 
Alabama Southern Nuclear Operating Company, Inc., et al., Docket Nos. 
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, 
Burke County, Georgia

    Date of amendments request: September 2, 2003.
    Brief Description of amendments: The amendments extend from 1 hour 
to 24 hours the completion time for Condition B of Technical 
Specification 3.5.1, which defines requirements for the restoration of 
an emergency core cooling system accumulator when it has been declared 
inoperable for a reason other than boron concentration.

[[Page 68675]]

    Date of issuance: November 18, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 162, 155, 129, & 107.
    Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise 
the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59220).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated November 18, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant, 
Unit 1, Rhea County, Tennessee

    Date of application for amendment: December 19, 2002.
    Brief description of amendment: The amendment consists of changes 
to Technical Specification (TS) 5.0, ``Administrative Controls,'' to 
incorporate three approved TS Task Force (TSTF) changes: TSTF-258, 
Revision 4, ``Changes to Section 5.0, Administrative Controls''; TSTF-
299, Revision 0, ``Administrative Controls Program 5.5.2.b Test 
Interval and Exception''; and TSTF-308, Revision 1, ``Determination of 
Cumulative and Projected Dose Contributions in the Radioactive Effluent 
Controls Program.'' In addition, two editorial changes are incorporated 
to update personnel titles and clarify required staffing levels.
    Date of issuance: November 13, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days.
    Amendment No.: 49.
    Facility Operating License No. NPF-90: Amendment revised the TSs.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15764).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated November 13, 2003.
    No significant hazards consideration comments received: No.

    For the Nuclear Regulatory Commission

    Dated at Rockville, Maryland, this 1st day of December 2003.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-30246 Filed 12-8-03; 8:45 am]
BILLING CODE 7590-01-P