[Federal Register Volume 69, Number 31 (Tuesday, February 17, 2004)]
[Notices]
[Pages 7517-7529]
From the Federal Register Online via the Government Printing Office [www.gpo.gov]
[FR Doc No: 04-3180]


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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 section 189a. (2) of the Atomic Energy Act of 1954, as 
amended (the Act), the U.S. Nuclear Regulatory Commission (the 
Commission or NRC staff) is publishing this regular biweekly notice. 
The Act requires the Commission publish notice of any amendments 
issued, or proposed to be issued and 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, January 22, 2004, through February 5, 2004. 
The last biweekly notice was published on February 3, 2004 (69 FR 
5200).

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. Within 60 days after the date of publication of this 
notice, 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.
    Normally, the Commission will not issue the amendment until the 
expiration of 60 days after the date of publication of this notice. The 
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment 
involves no significant hazards consideration. In addition, the 
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day 
comment period such that failure to act in a timely way would result, 
for example in derating or shutdown of the facility. Should the 
Commission take action prior to the expiration of either the comment 
period or the notice period, it will publish in the Federal Register a 
notice of issuance. Should the Commission make a final No Significant 
Hazards Consideration Determination, any hearing will take place 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 
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    Within 60 days after the date of publication of this notice, 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 2.309, 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 leave to intervene is filed by the above date, the Commission or a 
presiding officer designated by the Commission or by the Chief 
Administrative Judge of the Atomic Safety and Licensing Board Panel, 
will rule on the request and/or petition; and the Secretary or the 
Chief Administrative Judge of the Atomic Safety and Licensing Board 
will issue a notice of a hearing or an appropriate order.
    As required by 10 CFR 2.309, 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 general requirements: (1) The name, address and telephone 
number of the requestor or petitioner; (2) the nature of the 
requestor's/petitioner's right under the Act to be made a party to the 
proceeding; (3) the nature and extent of the requestor's/petitioner's 
property, financial, or other interest in the proceeding; and (4) the 
possible effect of any decision or order which may be entered in the 
proceeding on the

[[Page 7518]]

requestor's/petitioner's interest. The petition must also set forth the 
specific contentions which the petitioner/requestor seeks to have 
litigated at the proceeding.
    Each contention must consist of a specific statement of the issue 
of law or fact to be raised or controverted. In addition, the 
petitioner/requestor shall provide a brief explanation of the bases for 
the contention and a concise statement of the alleged facts or expert 
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The 
petitioner/requestor must also provide references to those specific 
sources and documents of which the petitioner is aware and on which the 
petitioner/requestor intends to rely to establish those facts or expert 
opinion. The petition must include 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/requestor to relief. A petitioner/
requestor 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.
    If a hearing is requested, and the Commission has not made a final 
determination on the issue of no significant hazards consideration, 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 by: (1) First class mail addressed to the Office of the 
Secretary of the Commission, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications 
Staff; (2) courier, express mail, and expedited delivery services: 
Office of the Secretary, Sixteenth Floor, One White Flint North, 11555 
Rockville Pike, Rockville, Maryland, 20852, Attention: Rulemaking and 
Adjudications Staff; (3) E-mail addressed to the Office of the 
Secretary, U.S. Nuclear Regulatory Commission, hearingdocket@nrc.gov; 
or (4) facsimile transmission addressed to the Office of the Secretary, 
U.S. Nuclear Regulatory Commission, Washington, DC, Attention: 
Rulemakings and Adjudications Staff at (301) 415-1101, verification 
number is (301) 415-1966. 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 it is requested that copies be transmitted either by 
means of facsimile transmission to 301-415-3725 or by email to 
OGCMailCenter@nrc.gov. A copy of the request for hearing and petition 
for leave to intervene should also be sent to the attorney for the 
licensee.
    Nontimely requests and/or petitions and contentions will not be 
entertained absent a determination by the Commission or the presiding 
officer of the Atomic Safety and Licensing Board that the petition, 
request and/or the contentions should be granted based on a balancing 
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
    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 email to 
pdr@nrc.gov.

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

    Date of amendment request: December 23, 2003.
    Description of amendment request: The licensee proposed to revise 
Section 4.5.D.2 of the Technical Specifications. This change would 
allow the licensee to leak test the Main Steam Isolation Valves (MSIV) 
at a lower pressure to eliminate the risk of lifting the disc of the 
inboard MSIV from its seat, producing inaccurate test data. The inboard 
MSIV would then have to be plugged before the leak test can be 
repeated. The current leak rate requirement is 0.05(0.75)La at Pa, 
where La is the maximum allowable leak rate, and Pa is the calculated 
peak containment pressure. This amendment would change this requirement 
to a leak rate of [le]11.9 standard cubic feet per hour (scfh) at a 
pressure [ge]20 psig. The leak rate of 11.9 scfh is a more conservative 
value based on control room habitability analysis, and 20 psig is based 
on the fact that post-accident pressure peaks in 2 to 3 seconds after 
an accident and would quickly drop below 20 psig. There is no physical 
changes to plant design associated with this amendment.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed amendment would change the pressure at which 
the leak rate of the MSIV is performed, while the leak rate test 
standard would be made more conservative than the current standard. No 
hardware design change is associated with the proposed amendment. 
Changing the MSIV leak test criterion would have no impact on the 
performance of the MSIVs. Thus, the proposed amendment would create no 
adverse effect on the functional performance of any plant structure, 
system, or component (SSC). All SSCs will continue to perform their 
design functions with no decrease in their capabilities to mitigate the 
consequences of previously analyzed postulated accidents. Accordingly, 
the proposed amendment would lead to no increase in the consequences of 
an accident previously evaluated, and no increase of the probability of 
an accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create

[[Page 7519]]

the possibility of a new or different kind of accident from any 
accident previously evaluated. The proposed amendment is not the result 
of a hardware design change, nor does it lead to the need for a 
hardware design change. There is no change in the methods the unit is 
operated. As a result, all SSCs will continue to perform as previously 
analyzed by the licensee, and previously evaluated and accepted by the 
NRC staff. Therefore, the proposed amendment will not create the 
possibility of a new or different kind of accident from any previously 
evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since the licensee did not propose to 
exceed or alter a design basis or safety limit, and did not propose to 
operate any component in a less conservative manner, the proposed 
amendment will not affect in any way the performance characteristics 
and intended functions of any SSC. Therefore, the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the proposed amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

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

    Date of amendments request: December 19, 2003, as supplemented 
January 14, 2004.
    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 December 19, 2003, as 
supplemented by letter dated January 14, 2004.
    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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: December 8, 2003.
    Description of amendment request: The amendment involves a one-time 
revision to the steam generator (SG) inservice inspection frequency 
requirements in Technical Specification 4.4.5.3a. to allow a 40-month 
inspection interval after the first inservice inspection following SG 
replacement, rather than after two consecutive inspections resulting in 
C-1 classification.
    Basis for proposed no significant hazards consideration 
determination:

[[Page 7520]]

As required by 10 CFR 50.91(a), the licensee has provided its analysis 
of the issue of no significant hazards consideration, which is 
presented below:

1. The proposed amendment does not involve a significant increase in 
the probability or consequences of an accident previously evaluated.

    The proposed amendment revises the steam generator inspection 
frequency to allow a 40-month inspection frequency after the first 
inservice inspection following SG replacement, rather than after two 
consecutive inspections resulting in C-1 classification. The ``C-1'' 
category is defined in the Technical Specifications as having 
inspection results that indicate ``less than 5% of the total tubes 
inspected are degraded tubes and none of the inspected tubes are 
defective.''
    The 100% inspection of the open steam generator tubes performed 
during RFO [Refueling Outage]-11 represents a quantity of tubes 
inspected that is significantly greater than the amount required by 
the Technical Specifications over two successive inspective periods 
(i.e., 3% of the total number of tubes in all steam generators 
required in the first inspection following SG replacement and the 
same quantity of the tubes to be examined in the second inspection). 
The RFO-11 100% tube inspection did not indicate the tubes had 
experienced degradation from the cycle of operation.
    The assessment of the condition of the steam generator tubes 
indicated the structural condition of the tubing had not changed 
during the first cycle of operation following steam generator 
replacement and these results that indicated the tubes would still 
meet their structural criteria over the proposed inspection 
frequency. The steam generator tube inspection meets the current 
industry examination guidelines without performing inspections 
during the next refueling outage.
    The steam generator inspection frequency extension does not 
introduce a new failure mode or impact any other plant systems or 
components. The proposed change does not alter plant design. The HNP 
[Harris Nuclear Plant] steam generator tubes do not have an active 
damage mechanism which could lead to the potential of primary-to-
secondary steam generator leakage.
    Therefore, the proposed inspection frequency change does not 
involve a significant increase in the probability or consequences of 
an accident previously evaluated.

2. The proposed amendment does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.

    The proposed change to extend the steam generator tube 
inspection frequency does not impact the design or operation of the 
steam generators or any other plant structure, system or component. 
Extending the inspection frequency of the steam generator tubes does 
not introduce any new failure modes. The proposed change does not 
alter plant design basis, or alter any potential accident previously 
evaluated.
    The proposed change revises the steam generator inspection 
frequency to allow a 40-month inspection interval after the first 
inservice inspection following SG replacement, rather than after two 
consecutive inspections resulting in C-1 classification. The first 
steam generator inspection following replacement inspected 100% of 
the open tubing in all three steam generators. This inspection 
exceeded the existing technical specification inspection over the 
two consecutive inspections. This inspection indicated there was no 
service-induced degradation in the steam generator tubes. The HNP 
first cycle inspection results were comparable with other recent 
Westinghouse model replacement steam generators.
    Therefore, the proposed inspection frequency change does not 
create the possibility of a new or different kind of accident from 
any accident previously evaluated.

3. The proposed amendment does not involve a significant reduction 
in a margin of safety.

    The steam generator tubes are an integral part of the reactor 
coolant system pressure boundary. The tubes are expected to maintain 
primary system pressure and inventory. The tubes are a barrier to 
keep radioactive fission products in the reactor coolant system from 
transferring to the secondary system. The steam generator tubes 
transfer the heat from the primary system to the secondary system. 
The ability of the steam generator tubes to perform these functions 
depends on the integrity of the tubes.
    Steam generator tube integrity is a function of design, 
environment, and current physical condition. Extending the steam 
generator tube inspection frequency by one operating cycle will not 
alter the function or design of the steam generators. The steam 
generator tube inspections performed during the first outage 
following steam generator replacement demonstrated that the tubes do 
not have an active damage mechanism, and the scope of these 
inspections significantly exceeded the requirements of the Technical 
Specifications. These inspection results were comparable to similar 
inspection results for second generation Alloy 690 models of 
replacement steam generators installed at other plants, and 
subsequent inspections at those plants yielded results that support 
this extension request. The improved design of the replacement steam 
generators also provides reasonable assurance that significant tube 
degradation is not likely to occur over the proposed operating 
period.
    Therefore, the proposed inspection frequency change does not 
involve a significant reduction in a margin of safety.

Based on the above, Progress Energy Carolinas, Inc. [Carolina Power 
& Light Company] concludes that the proposed amendment presents no 
significant hazards consideration under the standards set forth in 
10 CFR 50.92(c), and accordingly, a finding of ``no significant 
hazards consideration'' is justified.

    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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Allen Howe.

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

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

    Date of amendment request: November 5, 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 November 5, 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.


[[Page 7521]]


    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: Ms. Lisa F. Vaughn , Legal Department 
(PB05E), Duke Energy Corporation, 422 South Church Street, Charlotte, 
North Carolina 28201-1006.
    NRC Section Chief: John A. Nakoski.

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

    Date of amendment request: August 19, 2003.
    Description of amendment request: The proposed amendment would 
modify Note 5 to Pilgrim Nuclear Power Station Technical Specification 
(TS) Table 3.2.C-1, to change the Rod Block Monitor (RBM) power-
dependent Low Power Set Point (LPSP) allowable value from [le] 29% to 
[le] 25.9%. The proposed change would make the RBM LPSP consistent with 
plant procedures and the Core Operating Limits Report (COLR) allowable 
value used in compliance with TS 5.6.5.
    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 Rod Block Monitor (RBM) power dependent Low Power 
Set Point (LPSP) of [le] 25.9% corrects the incorrect value of [le] 
29% in Note 5 of TS Table 3.2.C-1 and is more restrictive than the 
incorrect value. The proposed set point allowable value of [le] 
25.9% provides rod block protection over a wider range from [le] 
25.9% to 100%, instead of [le] 29% to 100%, thereby enforcing RBM 
protection against rod withdrawal error at a lower power level. 
Also, the proposed requirement is consistent with the core operating 
limits report and is in accordance with License Amendment 138.
    The proposed RBM LPSP value ensures safe operation of the plant 
during startup and run modes. This requirement is not an accident 
precursor. The proposed analytical value [le] 25.9% was derived from 
the Average Power Range Monitor, Rod Block Monitor and Technical 
Specification (ARTS) improvement program methodology that was 
approved by License Amendment 138 and complies with the analytical 
methods required by Technical Specification 5.6.5. The proposed 
change provides additional assurance that the core operating limits 
are followed for safe operation and assumptions for core operating 
limits are met.
    Therefore, the probability or consequences of an accident 
previously evaluated is not significantly increased.

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 does not involve a change to the plant 
design or a new mode of equipment operation and enforces previously 
evaluated conditions. As a result, the proposed changes do not 
affect parameters or conditions that could contribute to the 
initiation of any new of different kind of accident. Therefore, this 
proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

3. Does the change involve a significant reduction in a margin of 
safety?

    Response: No.
    The proposed change increases the margin of safety by providing 
additional assurance that the RBM downscale trip is not bypassed for 
reactor power [ge] 25.9% of rated thermal power and is based on 
previously evaluated methodologies. 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell J. Roberts (Acting).

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

    Date of amendment request: December 8, 2003.
    Description of amendment request: The proposed amendment would 
delete a portion of the Pilgrim Nuclear Power Station (Pilgrim) 
Technical Specification (TS) 4.6.A.2, ``Primary System Boundary--
Thermal and Pressurization Limitations,'' and the associated TS Table 
4.6-3, ``Reactor Vessel Material Surveillance Program Withdrawal 
Schedule.'' The amendment would replace the existing Reactor Vessel 
Material Surveillance Program with the Boiling Water Reactor Vessel and 
Internal Project (BWRVIP) Integrated Surveillance Program (ISP) and 
Supplemental Surveillance Program

[[Page 7522]]

(SSP). The BWRVIP ISP/SSP would be incorporated into the Pilgrim 
Updated Final Safety Analysis Report.
    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 licensing basis continue to assure 
that applicable regulatory requirements are met and the same 
assurance of reactor pressure vessel integrity continues to be 
provided. The proposed changes to the TS[s] and licensing basis 
follow the [U.S. Nuclear Regulatory Commission] NRC Safety 
Evaluation approving the implementation of the ISP. The proposed 
changes ensure that the reactor pressure vessel will continue to be 
operated within the design, operational, and testing limits.
    The proposed changes do not modify the reactor coolant pressure 
boundary, (i.e., there are no changes in operating pressure, 
materials, or seismic loading). The proposed changes do not 
adversely affect the integrity of the reactor coolant pressure 
boundary such that its function in the control of radiological 
consequences is affected.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. Does the change create the possibility of a new or different kind 
of accident from any accident previously evaluated?

    Response: No.
    The proposed change does not involve a modification to the 
design of plant structures, systems, or components. Thus, no new 
modes of operation are introduced by the proposed change. The 
proposed change will not create any failure mode not bounded by 
previously evaluated accidents. 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 proposed implementation of ISP has been previously approved 
by the NRC and found to provide an acceptable alternative to plant-
specific reactor vessel material surveillance programs. Operation of 
Pilgrim within the program ensures that the reactor vessel materials 
will continue to behave in a non-brittle manner, thereby preserving 
the original safety design bases. 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell J. Roberts, Acting.

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

    Date of amendment request: January 16, 2004.
    Description of amendment request: The proposed amendment would 
approve an engineering evaluation performed in accordance with Pilgrim 
Nuclear Power Station Technical Specification (TS) 3.6.D.3 to justify 
continued power operation with safety relief valve (SRV) -3A and SRV-3D 
discharge pipe temperatures exceeding 212 degrees Fahrenheit 
(F) for greater than 24 hours as required by TS 3.6.D.4.
    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.
    Indication of elevated Safety Relief Valve (SRV) discharge pipes 
temperature is attributed to leakage past the SRVs. Excessive 
leakage, corresponding to temperatures greater than 255 
F, has the potential to affect SRV operability by 
affecting the SRV setpoint or response time. Continued operation 
with the discharge pipes of the SRVs indicating temperatures less 
than 255 F ensures that the leakage past the SRVs is 
maintained below the threshold for a leakage rate that would 
potentially have an effect on SRV setpoint or response time.
    Administrative controls are in place to ensure that margin to 
the 255 F value is maintained to assure reliable 
operation and to reduce the potential for damage to the pilot seat 
and disc. The SRVs continue to perform their intended design/safety 
function with no adverse effect because the leakage past the SRVs is 
maintained below the threshold for a leakage rate that could 
potentially have an adverse impact on the ability of the SRVs to 
perform their design functions. The impact of the leakage on other 
systems is small and all systems continue to be able to perform 
their intended design functions. Current accident analyses remain 
bounding and there is no significant increase in the consequences of 
any accident previously evaluated. In addition, as a result of the 
leakage, normal plant operating parameters are not affected and 
consequently there is no increased risk in a plant transient.
    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.
    Continued plant operation with elevated discharge pipe 
temperatures for SRV-3A & 3D within the bounds of the established 
administrative controls ensures that the leakage past the SRVs is 
maintained below the threshold for a leakage rate that would 
potentially have an effect on SRV setpoint or response time. This 
ensures that the SRVs will perform their intended design/safety 
function. The leakage does not adversely impact the ability of any 
system to perform its design function. The methods governing plant 
operation and testing remain consistent with current safety analysis 
assumptions. 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.
    Continued operation with the discharge pipes of SRV-3A & 3D 
indicating temperatures in excess of 212 F does not 
adversely affect existing plant safety margins or the reliability of 
the equipment assumed to operate in the safety analysis. The leakage 
does not result in excess SRV setpoint drift or response time 
changes. The imposed administrative controls on plant operation 
provide assurance that there will be no adverse effect on the 
ability of the SRVs to perform their intended design/safety 
function. There are no changes being made to safety analysis 
assumptions, safety limits or safety system settings that would 
adversely affect plant safety. 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: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: Darrell J. Roberts, Acting.

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

    Date of amendment request: November 25, 2003.

[[Page 7523]]

    Description of amendment request: Exelon Generation Company, LLC, 
the licensee, is proposing a change to the Limerick Generating Station 
(LGS), Units 1 and 2, Technical Specifications (TSs) contained in 
Appendix A to Operating Licenses NPF-39 and NPF-85, respectively. The 
proposed changes involve relocating the Reactor Coolant System (RCS) 
chemistry Limiting Conditions for Operation (LCO) from the Technical 
Specifications (TSs) to the Technical Requirements Manual (TRM). 
Additionally, proposed changes to TS RCS specific activity requirements 
involve removing various items and modifying the surveillance frequency 
of the isotopic analysis for Dose Equivalent I-131 from at least once 
per 31 days to once per 7 days.
    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 relocation of the reactor coolant 
system chemistry requirements from Technical Specifications (TS) to 
the Technical Requirements Manual (TRM) is administrative in nature 
and does not involve the modification of any plant equipment or 
affect basic plant operation. Conductivity, chloride and pH limits 
are not assumed to be an initiator of any analyzed event, nor are 
these limits assumed in the mitigation of consequences of accidents.
    The proposed elimination from TS of the reactor coolant system 
specific activity requirements involving E-bar, gross beta, and 
gross gamma does not involve the modification of any plant equipment 
or affect basic plant operation. Specific activity is not assumed to 
be an accident initiator, and the specific activity requirements 
remaining in TS provide reasonable assurance that the reactor 
coolant specific activity is maintained at a sufficiently low level 
to preclude offsite doses from exceeding a small fraction of the 
limits of 10 CFR part 100 in the event of an accident.
    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 proposed changes to relocate the reactor 
coolant system chemistry requirements from TS to the TRM, and to 
eliminate the reactor coolant system specific activity requirements 
involving E-bar, gross beta, and gross gamma, do not involve any 
physical alteration of plant equipment and do not change the method 
by which any safety-related system performs its function. As such, 
no new or different types of equipment will be installed, and the 
basic operation of installed equipment is unchanged. The methods 
governing plant operation and testing remain consistent with current 
safety analysis assumptions.
    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. The proposed change to the reactor coolant system 
chemistry requirements involves the relocation of current TS 
requirements to the TRM based on regulatory guidance and previously 
approved changes for other stations. The proposed change is 
administrative in nature, does not negate any existing requirement, 
and does not adversely affect existing plant safety margins or the 
reliability of the equipment assumed to operate in the safety 
analysis. As such, there are no changes being made to safety 
analysis assumptions, safety limits or safety system settings that 
would adversely affect plant safety as a result of the proposed 
change. Margins of safety are unaffected by requirements that are 
retained, but relocated from the TS to the TRM.
    The proposed change also involves the elimination from TS of the 
reactor coolant system specific activity requirements involving E-
bar, gross beta, and gross gamma. The specific activity requirements 
remaining in TS provide reasonable assurance that the reactor 
coolant specific activity is maintained at a sufficiently low level 
to preclude offsite doses from exceeding a small fraction of the 
limits of 10 CFR Part 100 in the event of an accident. As a result, 
the proposed change does not adversely affect existing plant safety 
margins.
    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: Mr. Edward Cullen, Vice President & General 
Counsel, Exelon Generation Company, LLC, 2301 Market Street, 
Philadelphia, PA 19101.
    NRC Section Chief: Darrell Roberts, Acting.

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

    Date of amendments request: December 19, 2003, as supplemented 
January 14, 2004.
    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 December 19, 2003, as 
supplemented by letter dated January 14, 2004.
    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

[[Page 7524]]

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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box 
1551, Raleigh, North Carolina 27602.
    NRC Section Chief: Allen G. Howe.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-220, Nine Mile 
Point Nuclear Station Unit No. 1 (NMP1), Oswego County, New York

    Date of amendment request: January 9, 2004.
    Description of amendment request: The licensee proposed to revise 
the Technical Specifications (TSs) and the Updated Final Safety 
Analysis Report (UFSAR) by replacing the current plant-specific reactor 
pressure vessel (RPV) material surveillance program with the Boiling 
Water Reactor Vessel and Internals Project (BWRVIP) Integrated 
Surveillance Program (ISP). Specifically, the proposed amendment would 
(1) delete the current reactor vessel material specimen surveillance 
schedule in Section 3/4.2.2, ``Minimum Reactor Vessel Temperature for 
Pressurization;'' (2) delete the special reporting requirement 
regarding material surveillance specimen examination in Section 
6.6.6.a; and (3) approve changes in the UFSAR to reflect the licensee's 
participation in the ISP and use of a methodology for determining 
neutron fluences.
    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 implement an ISP that has been evaluated by 
the NRC as meeting the requirements of paragraph III.C of Appendix H 
to 10 CFR [Part] 50; remove a TS surveillance requirement that 
prescribes a plant-specific withdrawal schedule for RPV surveillance 
specimens; and delete an unnecessary reporting requirement relating 
to RPV surveillance specimen examination. The proposed changes 
provide the same assurance of RPV integrity as has always been 
provided. Implementation of an ISP is not a precursor or initiator 
of any accident previously evaluated. No physical changes to the 
plant will result from the proposed changes. The proposed changes 
will not cause the RPV or interfacing systems to be operated outside 
of any design or testing limits, and will not alter any assumptions 
or initial conditions previously used in evaluating the radiological 
consequences of accidents.
    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 revise the NMP1 licensing bases to reflect 
participation in the BWRVIP ISP. The ISP was approved by the NRC 
staff as an acceptable material surveillance program that complies 
with 10 CFR [Part] 50, Appendix H. No physical changes to the plant 
will result from the proposed changes. The proposed changes do not 
affect the design or operation of any system, structure, or 
component. As an alternate monitoring program, the ISP cannot create 
a new failure mode involving the possibility of a new or different 
kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

3. Does the proposed change involve a significant reduction in a 
margin of safety?

    Response: No.
    The proposed changes have no impact on the margin of safety of 
any TS. There is no impact on safety limits or limiting safety 
system settings. The changes do not affect any plant safety 
parameters or setpoints. No physical or operational changes to the 
plant will result from the proposed changes.
    The RPV material surveillance program requirements contained in 
10 CFR [Part] 50, Appendix H provide assurance that adequate margins 
of safety exist during any condition of normal operation, including 
anticipated operational occurrences and system hydrostatic tests, to 
which the reactor coolant pressure boundary may be subjected over 
its service lifetime. The BWRVIP ISP has been approved by the NRC 
staff as an acceptable material surveillance program that complies 
with 10 CFR [Part] 50, Appendix H. The ISP will provide the material 
surveillance data that will assure that the safety margins required 
by the NRC regulations are maintained.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

Nine Mile Point Nuclear Station, LLC, Docket No. 50-410, Nine Mile 
Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York

    Date of amendment request: January 9, 2004.
    Description of amendment request: The licensee proposed to revise 
the licensing basis documented in the Updated Safety Analysis Report 
(USAR) by replacing the current plant-specific reactor pressure vessel 
(RPV) material surveillance program with the Boiling Water Reactor 
Vessel and Internals Project (BWRVIP) Integrated Surveillance Program 
(ISP). Specifically, the proposed amendment would approve revising the 
USAR to reflect the licensee's participation in the ISP and

[[Page 7525]]

use of a methodology for determining neutron fluences.
    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 change implements an ISP that has been evaluated by 
the NRC [Nuclear Regulatory Commission] as meeting the requirements 
of paragraph III.C of Appendix H to 10 CFR [Part] 50. The proposed 
change provides the same assurance of RPV integrity as has always 
been provided. Implementation of an ISP is not a precursor or 
initiator of any accident previously evaluated. No physical changes 
to the plant will result from the proposed change. The proposed 
change will not cause the RPV or interfacing systems to be operated 
outside of any design or testing limits, and will not alter any 
assumptions or initial conditions previously used in evaluating the 
radiological consequences of accidents.
    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 revises the NMP2 licensing bases to reflect 
participation in the BWRVIP ISP. The ISP was approved by the NRC 
staff as an acceptable material surveillance program that complies 
with 10 CFR [Part] 50, Appendix H. No physical changes to the plant 
will result from the proposed change. The proposed change does not 
affect the design or operation of any system, structure, or 
component. As an alternate monitoring program, the ISP cannot create 
a new failure mode involving the possibility of a new or different 
kind of accident.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

3. Does the proposed change involve a significant reduction in a 
margin of safety?

    Response: No.
    The proposed change has no impact on the margin of safety of any 
TS [Technical Specification]. There is no impact on safety limits or 
limiting safety system settings. The change does not affect any 
plant safety parameters or setpoints. No physical or operational 
changes to the plant will result from the proposed change.
    The RPV material surveillance program requirements contained in 
10 CFR [Part] 50, Appendix H provide assurance that adequate margins 
of safety exist during any condition of normal operation, including 
anticipated operational occurrences and system hydrostatic tests, to 
which the reactor coolant pressure boundary may be subjected over 
its service lifetime. The BWRVIP ISP has been approved by the NRC 
staff as an acceptable material surveillance program that complies 
with 10 CFR [Part] 50, Appendix H. The ISP will provide the material 
surveillance data that will assure that the safety margins required 
by the NRC regulations are maintained.
    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: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: January 16, 2004.
    Description of amendment request: The proposed amendment is to 
revise Technical Specifications (TS) for the Kewaunee Nuclear Power 
Plant (KNPP). The proposed change would revise (1) the containment 
closure TS to allow the equipment hatch to be open during refueling 
operations and/or during movement of irradiated fuel assemblies within 
containment, (2) the containment tests TS to require verification of 
the ability to close the equipment hatch periodically during refueling 
operations, and (3) the control room post-accident recirculation system 
TS to include requirements for operability during fuel handling 
operations in which the fuel that is being moved has been irradiated 
less than 30 days ago.
    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.
    The proposed change would allow the containment equipment hatch 
to remain open during irradiated fuel movement in containment. This 
penetration is not an initiator of any accident. The probability of 
a fuel handling accident (FHA) in the containment is unaffected by 
the position of the equipment hatch. Adoption of this change 
requires analyses, approved by the [Nuclear Regulatory Commission] 
NRC staff, demonstrating that the dose consequences of a FHA with 
the equipment hatch open are acceptable. Therefore, the proposed 
change does 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.
    The proposed change does not involve the addition or 
modification of any plant equipment. Also, the proposed change would 
not alter the design, configuration, or method of operation of the 
plant beyond the standard functional capabilities of the equipment. 
The proposed change involves a change to the Technical 
Specifications (TS) that would allow the equipment hatch to remain 
open during irradiated fuel movement within the containment. Having 
the equipment hatch open does not create the possibility of a new 
accident. 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.
    Analysis demonstrates that the resultant doses associated with a 
fuel handling accident are well within the appropriate acceptance 
limits. This change removes a defense-in-depth barrier that the 
analysis did not credit but provides additional restrictions on 
fission product release. Thus, this proposed change has the 
potential for an increased dose at the site boundary due to a FHA; 
however, the analysis demonstrates that the resultant doses are well 
within the appropriate acceptance limits. Without the containment 
structure, analysis demonstrates that the dose consequences are 
still approximately 20% of the allowable value for the control room 
dose and less than 2% of the allowable value for offsite dose. Thus, 
the margin of safety has not been significantly reduced. 
Administrative provisions that facilitate closing the equipment 
hatch following an evacuation of the containment further reduces the 
offsite doses in the event of a FHA and provides additional margin 
to the calculated offsite doses. 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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

[[Page 7526]]

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

    Date of amendment request: December 1, 2003.
    Description of amendment request: The proposed amendment will 
modify Fort Calhoun Station Technical Specification (TS) 2.7, 
``Electrical Systems,'' TS Table 3-5, ``Minimum Frequencies for 
Equipment Tests,'' and TS 5.0, ``Administrative Controls.'' This 
proposed amendment modifies the requirements for diesel generator (DG) 
fuel oil for consistency with the Improved Standard Technical 
Specifications (ISTS) and adds requirements for DG lubricating oil and 
DG starting air. The proposed changes will assure that the required 
quality and quantity of DG fuel oil is maintained and also will assure 
that sufficient DG lubricating oil and DG starting air is maintained.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

1. The proposed change does not involve a significant increase in 
the probability or consequences of an accident previously evaluated. 
The proposed change will revise Technical Specification (TS) 2.7, 
``Electrical Systems,'' TS Table 3-5, ``Minimum Frequencies for 
Equipment Tests,'' and TS 5.0, ``Administrative Controls.'' This 
proposed amendment modifies the requirements for Diesel Generator 
(DG) Fuel Oil for consistency with the Improved Standard Technical 
Specifications (ISTS) and adds requirements for DG Lubricating Oil, 
and DG Starting Air. The Surveillance interval of Diesel Fuel Supply 
Surveillance [Table 3-5, Item 9 (changed to 9a)] is being changed 
from daily to monthly. The 31 day Surveillance interval is adequate 
to ensure that a sufficient supply of fuel oil is available, since 
low level alarms are provided and unit operators would be aware of 
any large uses of fuel oil during this period. Therefore, this 
change does not significantly increase the probability of a 
previously analyzed accident. Further, an increase of the 
Surveillance interval will not affect the capability of the 
component or system to perform its function. Therefore, this change 
does not significantly increase the consequences of a previously 
analyzed accident. All other changes are more restrictive changes. 
The changes will ensure that proper Limiting Conditions for 
Operation are entered for equipment or functional inoperability. 
There are no physical alterations being made to the DGs or related 
systems.

    With regards to TSTF-254, Rev. 2, the proposed change does not 
require any physical change to any plant systems, structures, or 
components nor does it require any change in systems or plant 
operations. The proposed change does not require any change in 
safety analysis methods or results. The water content of the DG fuel 
oil system is not considered an accident initiator. The change to 
reduce the fuel oil sampling frequency for water content from 31 
days to 92 days does not present a significant impact to DG 
operability or significantly degrade DG performance and, therefore, 
does not present a significant detrimental impact on structures, 
systems, or components that support accident recovery.
    With regards to TSTF-374, Rev. 0, the proposed changes relocate 
the specific ASTM Standard references from the Administrative 
Controls Section of TS to a licensee-controlled document. Since any 
change to the licensee-controlled document will be evaluated 
pursuant to the requirements of 10 CFR 50.59, ``Changes, tests and 
experiments,'' no increase in the probability or consequences of an 
accident previously evaluated is involved. In addition, the ``clear 
and bright'' test used to establish the acceptability of new fuel 
oil for use prior to addition to storage tanks has been expanded to 
allow a water and sediment content test to be performed to establish 
the acceptability of new fuel oil. The proposed changes revise Bases 
for TS 3.2 to reference the current specific ASTM Standards. The 
Bases for TS 3.2 are revised to indicate that the API gravity is 
tested in accordance with ASTM D287.
    Relocating the specific ASTM Standard references from the TS to 
a licensee-controlled document, allowing a water and sediment 
content test to be performed to establish the acceptability of new 
fuel oil, and revising the TS Bases will not affect nor degrade the 
ability of the DGs to perform their specified safety function. Fuel 
oil quality will continue to meet ASTM requirements.
    The proposed changes do not adversely affect accident initiators 
or precursors nor alter the design assumptions, conditions, and 
configuration of the facility or the manner in which the plant is 
operated and maintained. The proposed changes do not alter or 
prevent the ability of structures, systems, and components (SSCs) 
from performing their intended function to mitigate the consequences 
of an initiating event within the assumed acceptance limits. The 
proposed changes do not affect the source term, containment 
isolation, or radiological release assumptions used in evaluating 
the radiological consequences of an accident previously evaluated. 
Further, the proposed changes do not increase the types and amounts 
of radioactive effluent that may be released offsite, nor 
significantly increase individual or cumulative occupational/public 
radiation exposures.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.

2. The proposed change does not create the possibility of a new or 
different kind of accident from any accident previously evaluated.

    The proposed changes will not result in any physical alterations 
to the DGs, any plant configuration, systems, equipment, or 
operational characteristics. There will be no changes in operating 
modes, or safety limits, or instrument limits. With the proposed 
changes in place, Technical Specifications will retain requirements 
for the DGs.
    With regards to TSTF-254, Rev. 2, the accident analyses do not 
consider the water content of the EDG fuel oil systems. Failure of a 
DG to start and load upon accident initiation is considered in the 
accident analyses, but is not affected by the proposed change to the 
fuel oil sampling Surveillance intervals. The existing analyses 
remain unchanged and the proposed TS change does not affect any 
accident initiators that would create a new accident.
    With regards to TSTF-374, Rev. 0, the proposed changes relocate 
the specific ASTM Standard references from the Administrative 
Controls Section of the TS to a licensee-controlled document. In 
addition, the ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
proposed changes [ ] also revise the Bases of TS 3.2 to reference 
the current specific ASTM Standards. The Bases for TS 3.2 is revised 
to indicate that the API gravity is tested in accordance with ASTM 
D287.
    Therefore, the proposed changes do not create the possibility of 
a new or different kind of accident from any previously evaluated.

3. The proposed change does not involve a significant reduction in a 
margin of safety.

    The proposed changes clarify the regulatory requirements for the 
DGs. The Completion Times and Frequencies established are within 
those invoked by the present Technical Specifications or equal to 
those previously reviewed and approved for use by the NRC. The 
proposed changes will not alter any physical or operational 
characteristics of the DGs and associated systems and equipment.
    With regards to TSTF-254, Rev. 2, the proposed change does not 
require any change in accident analysis methods or results. The 
safety margin as established in the current license basis remains 
unchanged. Reducing the Surveillance interval for DG fuel oil 
sampling does not, in itself, result in a measurable impact on the 
operability of the DGs. The water content of the DG fuel oil systems 
will continue to be assessed and corrective action taken should any 
condition adverse to DG operability be detected.
    With regards to TSTF-374, Rev. 0, [t]he proposed changes 
relocate the specific ASTM Standard references from the 
Administrative Controls Section of [the] TS to a licensee-controlled 
document. Instituting the proposed changes will continue to ensure 
the use of current applicable ASTM Standards to evaluate the quality 
of both new and stored fuel oil designated for use in the emergency 
DGs. The detail associated with the specific ASTM Standard 
references is not required to be in the TS to provide adequate 
protection of the public health and safety, since the TS still 
retain the requirement for compliance with the applicable ASTM 
Standard. Changes

[[Page 7527]]

to the licensee-controlled document are performed in accordance with 
the provisions of 10 CFR 50.59. Should it be determined that future 
changes involve a potential reduction in a margin of safety, NRC 
review and approval would be necessary prior to implementation of 
the changes. This approach provides an effective level of regulatory 
control and provides for a more appropriate change control process.
    The ``clear and bright'' test used to establish the 
acceptability of new fuel oil for use prior to addition to storage 
tanks has been expanded to allow a water and sediment content test 
to be performed to establish the acceptability of new fuel oil. The 
proposed changes revise the Bases for TS 3.2 to reference the 
current specific ASTM Standards. The Bases for TS 3.2 is revised to 
indicate that the API gravity is tested in accordance with ASTM 
D287. The level of safety of facility operation is unaffected by the 
proposed changes since there is no change in the intent of the TS 
requirements of assuring fuel oil is of the appropriate quality for 
emergency DG use. The proposed changes provide the flexibility 
needed to maintain state-of-the-art technology in fuel oil sampling 
and analysis methodology.
    Therefore, the proposed changes do not involve a 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: James R. Curtiss, Esq., Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Stephen Dembek.

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

    Date of amendment request: December 12, 2003.
    Description of amendment request: Proposed changes to the Technical 
Specifications (TSs) of the Control Room Emergency Filtration System 
(CREPS) would no longer require it to be OPERABLE in COLD SHUTDOWN. 
However, CREPS would have to be operable during operations with 
potential for draining the reactor vessel. The TSs for the Control Room 
Ventilation Radiation Monitor would be revised so that OPERABILITY 
would no longer be required during refueling. However, OPERABILITY 
would be required for operations with potential for draining the 
reactor vessel.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:
1. Does the change involve a significant increase in the probability 
or consequences of an accident previously analyzed?

    Response: No.
    The proposed changes to Table 3.3.7.1-1, Radiation Monitoring 
Instrumentation, and Table 4.3.7.1-1, Radiation Monitoring 
Instrumentation Surveillance Requirements, adds ``recently'' to 
modify irradiated fuel in the ``*'' footnote to provide consistency 
with TSTF-51, Rev. 2. Proposed changes to eliminate Operational 
Condition 5 from Tables 3.3.7.1-1 and 4.3.7.1-1, Control Room 
Ventilation Radiation Monitor, Operational Condition 4 from Control 
Room Emergency Filtration (CREF) System and adding operations with 
the potential for draining the reactor vessel (OPDRV) to Tables 
3.3.7.1-1 and 4.3.7.1-1 footnote ``*'' and the CREF System are 
consistent with NUREG-1433 Vol. 1, Rev. 2, Standard Technical 
Specifications, General Electric Plants.
    The proposed changes associated with the fuel handling accident 
(FHA) do not involve a change to structures, components, or systems 
that would affect the probability of an accident previously 
evaluated in the Hope Creek Updated Final Safety Analysis Report 
(UFSAR). The FHA for the Hope Creek Generating Station (HCGS) is 
defined as a drop of a fuel assembly over irradiated assemblies in 
the reactor core 24 hours after reactor shutdown. Alternative Source 
Term (AST) is used to evaluate the dose consequences of a postulated 
accident. The FHA has been analyzed without credit for Secondary 
Containment, Filtration Recirculation and Ventilation System (FRVS), 
and CREF system. The resultant radiological consequences are within 
the acceptance criteria set forth in 10 CFR 50.67 and Regulatory 
Guide (RG) 1.183. This amendment does not alter the methodology or 
equipment used in fuel handling operations. The equipment hatch, 
personnel air locks, other containment penetrations, or any 
component thereof is not an accident initiator. Actual fuel handling 
operations are not affected by the proposed changes.
    Consequently the probability of a previously analyzed FHA is not 
affected by the proposed amendment. No other accident initiator is 
affected by the proposed changes.
    Therefore, this proposed amendment does not involve a 
significant increase in the probability of occurrence or 
radiological consequences of an accident previously analyzed.

2. Does the change create the possibility of a new or different kind 
of accident from any accident previously analyzed?

    Response: No.
    The proposed changes to Table 3.3.7.1-1, Radiation Monitoring 
Instrumentation, and Table 4.3.7.1-1, Radiation Monitoring 
Instrumentation Surveillance Requirements, adds recently'' to modify 
irradiated fuel in the ``*'' footnote provides consistency with 
TSTF-51, Rev. 2. Proposed changes to eliminate Operational Condition 
5 from Tables 3.3.7.1-1 and 4.3.7.1-1, Control Room Ventilation 
Radiation Monitor, Operational Condition 4 from CREF System and 
adding OPDRV to Table 3.3.7.1-1 and 4.3.7.1-1 footnote ``*'' and the 
CREF System are consistent with NUREG-1433 Vol. 1, Rev. 2, Standard 
Technical Specifications, General Electric Plants.
    The proposed amendment will not create the possibility for a new 
or different type of accident from any accident previously evaluated 
because changes to the allowable activity in the primary and 
secondary systems do not result in changes to the design or 
operation of these systems. The evaluation of the effects of the 
proposed changes indicates that all design standard and applicable 
safety criteria limits are met. Equipment important to safety will 
continue to operate as designed. Component integrity is not 
challenged. The changes do not result in any event previously deemed 
incredible being made credible. The changes do not result in more 
adverse conditions or result in any increase in the challenges to 
safety systems. The systems affected by the changes are used to 
mitigate the consequences of an accident that has already occurred. 
The proposed TS changes do not significantly affect the mitigative 
function of these systems.
    Therefore, the proposed changes would not create the possibility 
of a new or different kind of accident from any previously 
evaluated.

3. Does the change involve a significant reduction in the margin of 
safety?

    Response: No.
    The proposed changes to Table 3.3.7.1-1, Radiation Monitoring 
Instrumentation, and Table 4.3.7.1-1, Radiation Monitoring 
Instrumentation Surveillance Requirements, adds ``recently'' to 
modify irradiated fuel in the ``*'' footnote provides consistency 
with TSTF-51, Rev. 2. Proposed changes to eliminate Operational 
Condition 5 from Tables 3.3.7.1-1 and 4.3.7.1-1 Control Room 
Ventilation Radiation Monitor, Operational Condition 4 from CREF 
System and adding OPDRV to Table 3.3.7.1-1 and 4.3.7.1-1 footnote 
``*'' and the CREF System are consistent with NUREG-1433 Vol. 1, 
Rev. 2, Standard Technical Specifications, General Electric Plants.
    The proposed changes revise the TS to establish operational 
conditions where specific activities represent situations during 
which significant radioactive releases can be postulated. These 
operational conditions are consistent with the design basis analysis 
and are established such that the radiological consequences remain 
at or below the regulatory guidelines. Safety margins and analytical 
conservatisms are retained to ensure that the analysis adequately 
bounds all postulated event scenarios. The proposed TS continue[s] 
to ensure that the total effective dose equivalent (TEDE) for the 
control room (CR), the exclusion area boundary (EAB), and low 
population zone (LPZ) boundaries are below the corresponding 
acceptance criteria specified in I0 CFR 50.67 and RG 1.183.
    Therefore, these changes do not involve a significant reduction 
in margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three

[[Page 7528]]

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: Darrell Roberts, Acting.

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

    Date of amendment request: December 24, 2003.
    Description of amendment request: The amendment request changes the 
Technical Specifications (TSs) to allow the use of GE14 fuel in reload 
cycle 13. Specifically, the proposed changes modify the TSs to reflect 
the use of General Electric (GE) core reload analysis methodologies. 
The proposed changes would revise the limiting conditions for operation 
for the recirculation loops to modify and add action statements to 
provide further thermal limit control during single-loop operation to 
be consistent with GE methodology specified in the core operating 
limits report. The proposed changes also modify the TS definitions and 
TS requirements for average planar linear heat generation rate 
consistent with NUREG-1433, ``Standard Technical Specifications (STS) 
General Electric Plants, BWR/4,'' Revision 2. Additionally, TS Section 
6.9.1.9 would be revised to correct an error in a previous amendment 
that inadvertently removed a reference. The NRC-approved reference 
would be restored to TS 6.9.1.9 in the format prescribed in NUREG-1433, 
Revision 2.
    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 revised information and references relative to the fuel 
vendor's calculation methodologies throughout the Technical 
Specifications are considered to be administrative in nature because 
they reflect the NRC approved methodologies to be used by PSEG 
Nuclear LLC and the fuel vendor to develop operating and safety 
limits for the fuel and core designs. The changes to the 
Recirculation System Action statements ensure the appropriate 
adjustments are made to core operating limits for single loop 
operation, and the Core Operating Limits Report (COLR) will still be 
developed in accordance with NRC approved methods. These proposed 
changes do not alter the method of operating the plant and have no 
effect on the probability of an accident initiating event or 
transient.
    There are no significant increases in the radiological 
consequences of an accident previously evaluated. The basis of the 
COLR and the PSEG Nuclear LLC and fuel vendor calculation 
methodologies is to ensure that no mechanistic fuel damage is 
calculated to occur if the limits on plant operation are not 
violated. The COLR will continue to preserve the existing margin to 
fuel damage and the probability of fuel damage is not increased.
    Therefore, the proposed change does not involve an 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.
    These changes do not involve any new method for operating the 
facility, any changes to setpoints, or any new facility 
modifications for the reload core operation. No new initiating 
events or transients result from these changes.
    The revised information and references relative to the fuel 
vendor's calculation methodologies throughout the Technical 
Specifications are considered to be administrative in nature because 
they reflect the NRC approved methodologies to be used by PSEG 
Nuclear LLC and the fuel vendor to develop operating and safety 
limits for the fuel and core designs. The changes to the 
Recirculation System Action statements ensure the appropriate 
adjustments are made to core operating limits for single loop 
operation, and the COLR will still be developed in accordance with 
NRC-approved methods.
    Therefore, the proposed Technical Specification 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 revised information and references relative to the fuel 
vendor's calculation methodologies throughout the Technical 
Specifications are considered to be administrative in nature because 
they reflect the NRC approved methodologies to be used by PSEG 
Nuclear LLC and the fuel vendor to develop operating and safety 
limits for the fuel and core designs. The changes to the 
Recirculation System Action statements ensure the appropriate 
adjustments are made to core operating limits for single loop 
operation, and the COLR will still be developed in accordance with 
NRC approved methods. The proposed changes will continue to ensure 
that the plant is operated within specified acceptable fuel design 
limits. Therefore, the proposed Technical Specifications 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: Darrell Roberts, Acting.

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

[[Page 7529]]

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 email to pdr@nrc.gov.

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

    Date of application for amendment: May 12, 2003, as supplemented by 
letter dated October 29, 2003.
    Brief description of amendment: The amendment changes 
administrative Technical Specification (TS) 5.5.12 regarding 
containment integrated leakage rate testing (ILRT) and TS 3.6.5.1.1 
regarding drywell bypass leak rate testing (DWBT). The change would 
allow for a one-time extension of the interval from 10 to 15 years for 
performance of the next ILRT and DWBT.
    Date of issuance: January 28, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 164.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34666).
    The October 29, 2003, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 28, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: March 13, 2003.
    Brief description of amendment: The amendment deletes Technical 
Specification (TS) 6.8.4.c, ``Post Accident Sampling,'' and thereby 
eliminates the requirements to have and maintain the post accident 
sampling system at the Hope Creek Generating Station.
    Date of issuance: January 29, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 180 days.
    Amendment No.: 149.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28856).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: June 17, 2003.
    Brief description of amendment: The amendment corrects 
typographical errors in the Technical Specification (TS) Index and 
deletes TS 4.6.2.1.b.2.b, verification that thermal power is less than 
or equal to 1% of rated thermal power at least once per hour when the 
suppression chamber temperature exceeds 95 F. The 
proposed TS change is consistent with the standard TSs for General 
Electric Plants, Boiling-Water Reactor/4 (NUREG-1433, Revision 2).
    Date of issuance: January 30, 2004.
    Effective date: As of the date of issuance, to be implemented 
within 60 days.
    Amendment No.: 150.
    Facility Operating License No. NPF-57: This amendment revised the 
TSs.
    Date of initial notice in Federal Register: July 8, 2003 (68 FR 
40717).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 30, 2004.
    No significant hazards consideration comments received: No.

PSEG Nuclear, LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear 
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey

    Date of application for amendments: June 6, 2003.
    Brief description of amendments: The amendments modify the Salem 
Nuclear Generating Station, Unit Nos. 1 and 2, Technical Specifications 
(TSs) by: (1) Adding a footnote to TS 3/4.11.2.5 to clarify the 
applicability of the Limiting Condition for Operation while the system 
is removed from service for maintenance; (2) revising Surveillance 
Requirement 4.11.2.5 to delete the reference to hydrogen concentration; 
and (3) revising the corresponding TS Bases.
    Date of issuance: January 29, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment Nos.: 261 and 243.
    Facility Operating License Nos. DPR-70 and DPR-75: The amendments 
revised the TSs.
    Date of initial notice in Federal Register: August 5, 2003 (68 FR 
46246).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 29, 2004.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: July 18, 2003.
    Brief description of amendments: The amendments modified Technical 
Specification (TS) requirements for mode change limitations to adopt 
Industry/TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility 
in Mode Restraints.''
    Date of issuance: January 23, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days from the date of issuance.
    Amendment Nos.: 109 and 109.
    Facility Operating License Nos. NPF-87 and NPF-89: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59222).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 23, 2004.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 9th day of February 2004.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 04-3180 Filed 2-13-04; 8:45 am]
BILLING CODE 7590-01-P