[Federal Register Volume 68, Number 111 (Tuesday, June 10, 2003)]
[Notices]
[Pages 34660-34678]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-14277]


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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 Pub. L. 97-415, the U.S. Nuclear Regulatory Commission 
(the Commission or NRC staff) is publishing this regular biweekly 
notice. Pub. L. 97-415 revised section 189 of the Atomic Energy Act of 
1954, as amended (the Act), to require the Commission to publish notice 
of any amendments issued, or proposed to be issued, under a new 
provision of section 189 of the Act. This provision grants the 
Commission the authority to issue and make immediately effective any 
amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from, May 16, 2003, through May 29, 2003. The 
last biweekly notice was published on May 27, 2003 (68 FR 28843).

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

    The Commission has made a proposed determination that the following 
amendment requests involve no significant hazards consideration. Under 
the Commission's regulations in 10 CFR 50.92, this means that operation 
of the facility in accordance with the proposed amendment would not (1) 
involve a significant increase in the probability or consequences of an 
accident previously evaluated; or (2) create the possibility of a new 
or different kind of accident from any accident previously evaluated; 
or (3) involve a significant reduction in a margin of safety. The basis 
for this proposed determination for each amendment request is shown 
below.
    The Commission is seeking public comments on this proposed 
determination. Any comments received within 30 days after the date of 
publication of this notice will be considered in making any final 
determination.
    Normally, the Commission will not issue the amendment until the 
expiration of the 30-day notice period. However, should circumstances 
change during the notice period such that failure to act in a timely 
way would result, for example, in derating or shutdown of the facility, 
the Commission may issue the license amendment before the expiration of 
the 30-day notice period, provided that its final determination is that 
the amendment involves no significant hazards consideration. The final 
determination will consider all public and State comments received 
before action is taken. Should the Commission take this action, it will 
publish in the Federal Register a notice of issuance and provide for 
opportunity for a hearing after issuance. The Commission expects that 
the need to take this action will occur very infrequently.
    Written comments may be submitted by mail to the Chief, Rules and 
Directives Branch, Division of Administrative Services, Office of 
Administration, U.S. Nuclear Regulatory Commission, Washington, DC 
20555-0001, and should cite the publication date and page number of 
this Federal Register notice. Written comments may also be delivered to 
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville, 
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of 
written comments received may be examined at the Commission's Public 
Document Room (PDR), located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The 
filing of requests for a hearing and petitions for leave to intervene 
is discussed below.
    By July 10, 2003, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish

[[Page 34661]]

those facts or expert opinion. Petitioner must provide sufficient 
information to show that a genuine dispute exists with the applicant on 
a material issue of law or fact. Contentions shall be limited to 
matters within the scope of the amendment under consideration. The 
contention must be one which, if proven, would entitle the petitioner 
to relief. A petitioner who fails to file such a supplement which 
satisfies these requirements with respect to at least one contention 
will not be permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station, 
Unit 1, DeWitt County, Illinois

    Date of amendment request: January 29, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 3.6.5.1, ``Drywell,'' Surveillance 
Requirement 3.6.5.1.3 to delay the performance of the next drywell 
bypass leakage test to no later than November 23, 2008. The proposed 
amendment would also revise TS 5.5.13, ``Primary Containment Leakage 
Rate Testing Program,'' to remove an exception which is no longer 
applicable and to reflect a one-time deferral of the primary 
containment Type A test to no later than November 23, 2008.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the change involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed changes will revise TS 3.6.5.1, ``Drywell,'' 
Surveillance Requirement SR 3.6.5.1.3 to delay the performance of 
the next drywell bypass leakage rate test (DBLRT) to no later than 
November 23, 2008. This request will also revise CPS [Clinton Power 
Station] TS 5.5.13, ``Primary Containment Leakage Rate Testing 
Program,'' to reflect a one-time deferral of the primary containment 
Type A test to no later than November 23, 2008. The current Type A 
test interval of 10 years, based on past performance, would be 
extended on a one-time basis to 15 years from the last Type A test. 
In addition, AmerGen is proposing to delete from TS 5.5.13 the 
expired exception that allowed deferral of the leakage rate testing 
of the primary containment penetration 1MC-042 until the seventh 
refueling outage.
    The drywell houses the reactor pressure vessel, the reactor 
coolant recirculating loops, and branch connections of the Reactor 
Coolant System (RCS), which have isolation valves at the primary 
containment boundary. The function of the drywell is to maintain a 
pressure boundary that channels steam from a Loss of Coolant 
Accident (LOCA) to the suppression pool, where it is condensed. Air 
forced from the drywell is released into the primary containment 
through the suppression pool. The suppression pool is a concentric 
open container of water with a stainless steel liner that is located 
at the bottom of the primary containment. The suppression pool is 
designed to absorb the decay heat and sensible heat released during 
a reactor blowdown from safety/relief valve (SRV) discharges or from 
a LOCA.
    The function of the Mark III containment is to isolate and 
contain fission products released from the RCS following a design 
basis LOCA and to confine the postulated release of radioactive 
material to within limits. The test interval associated with the 
drywell bypass leakage and Type A testing is not a precursor of any 
accident previously evaluated. Therefore, extending these test 
intervals on a one-time basis from 10 years to 15 years does not 
result in an increase in the probability of occurrence of an 
accident. The successful performance history of the drywell bypass 
leakage and Type A testing provides assurance that the CPS drywell 
and primary containment will not exceed allowable leakage rate 
values specified in the TS and will continue to perform its design 
function following an accident. The risk assessment of the proposed 
changes has concluded that there is an insignificant increase in 
total population dose rate and an insignificant increase in the 
conditional containment failure probability.
    Therefore, the proposed changes do 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 changes for a one-time extension of the drywell 
bypass leakage and Type A tests and deletion of an expired local 
leak rate test exception for CPS, will not affect the control 
parameters governing unit

[[Page 34662]]

operation or the response of plant equipment to transient and 
accident conditions. The proposed changes do not introduce any new 
equipment or modes of system operation. No installed equipment will 
be operated in a new or different manner. As such, no new failure 
mechanisms are introduced.
    Therefore, the proposed changes do 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 a margin 
of safety?
    Response: No.
    CPS is a General Electric BWR/6 plant with a Mark III 
containment system. The Mark III containment design is a single-
barrier pressure containment and a multi-barrier fission containment 
system consisting of the drywell and primary containment. The 
drywell houses the reactor pressure vessel, the reactor coolant 
recirculating loops, and branch connections of the RCS, which have 
isolation valves at the primary containment boundary. The function 
of the drywell is to maintain a pressure boundary that channels 
steam from a LOCA to the suppression pool, where it is condensed. 
The suppression pool is an annular pool of demineralized water 
between the drywell and the outer primary containment boundary. This 
pool covers the horizontal vent openings in the drywell to maintain 
a water seal between the drywell interior and the remainder of the 
containment volume. The primary containment consists of a steel-
lined, reinforced concrete vessel, which surrounds the RCS and 
provides an essentially leak-tight barrier against an uncontrolled 
release of radioactive material to the environment. Additionally, 
this structure provides shielding from the fission products that may 
be present in the primary containment atmosphere following accident 
conditions. The primary containment is penetrated by access, piping 
and electrical penetrations.
    The integrity of the drywell is periodically verified by 
performance of the DBLRT. This test ensures that the measured 
drywell bypass leakage is bounded by the safety analysis 
assumptions. The drywell integrity is further verified by a number 
of additional tests, including drywell airlock door seal leakage 
tests, overall drywell airlock leakage tests and periodical visual 
inspections of exposed accessible interior and exterior drywell 
surfaces. Additional confidence that significant degradation in the 
drywell leaktightness has not developed is provided by the periodic 
qualitative assessment of drywell performance.
    The integrity of the primary containment penetrations and 
isolation valves is verified through Type B and Type C local leak 
rate tests (LLRTs) and the overall leak-tight integrity of the 
primary containment is verified by a Type A integrated leak rate 
test (ILRT) as required by 10 CFR 50, Appendix J, ``Primary Reactor 
Containment Leakage Testing for Water-Cooled Power Reactors.'' These 
tests are performed to verify the essentially leak-tight 
characteristics of the primary containment at the design basis 
accident pressure. The proposed changes for a one-time extension of 
the drywell bypass leakage and Type A tests and deletion of an 
expired local leak rate test exception for CPS, do not effect the 
method for drywell or containment testing or the test acceptance 
criteria.
    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: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendments request: April 25, 2003.
    Description of amendments request: The amendments would revise 
Specification 5.3.1 in Section 5.3, ``Unit Staff Qualifications,'' of 
the Technical Specifications, and add a new Specification 5.3.2. 
Specification 5.3.1 states the qualifications of the unit staff. The 
revision would state there is an exception for operator license 
applicants and the new specification would provide the requirements for 
these applicants. Only the qualifications of operator license 
applicants are being changed. Because a new specification would be 
added, the existing Specification 5.3.2 would also be renumbered 5.3.3.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The proposed Technical Specification (TS) change is an 
administrative change to clarify the current requirements for 
licensed operator qualifications and licensed operator training 
program. These changes conform to the current requirements of 10 CFR 
[Part] 55. The TS requirements for all other unit staff 
qualifications remain unchanged.
    Although licensed operator qualifications and training may have 
an indirect impact on accidents [involving operator action] 
previously evaluated, the NRC considered this impact during the 
rulemaking process, and by promulgation of the revised 10 CFR [Part] 
55 rule, concluded that this impact remains acceptable as long as 
the licensed operator training program is certified to be accredited 
and is based on a systems approach to training. Palo Verde's 
licensed operator training program is accredited by INPO [Institute 
of Nuclear Power Operations] and is based on a systems approach to 
training.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed TS change is an administrative change to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training program and to conform to the revised 10 
CFR [Part] 55. The TS requirements for all other unit staff 
qualifications remain unchanged.
    As noted above, although licensed operator qualifications and 
training may have an indirect impact on the possibility of a new or 
different kind of accident [involving operator action] from any 
accident previously evaluated, the NRC considered this impact during 
the rulemaking process, and by promulgation of the revised rule, 
concluded that this impact remains acceptable as long as the 
licensed operator training program is certified to be accredited and 
is based on a systems approach to training. [That is to say an 
accredited license operator training program that is based on a 
systems approach to training would not introduce a new or different 
kind of accident.] As previously noted, Palo Verde's licensed 
operator training program is accredited by INPO and is based on a 
systems approach to training.
    Additionally, the proposed TS change does not affect plant 
design, hardware, system operation, or procedures. Thus, the 
proposed amendment request does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed TS change is an administrative change to clarify 
the current requirements applicable to licensed operator 
qualifications and licensed operator-training program. This change 
is consistent with the requirements of 10 CFR [Part] 55. The TS 
qualification requirements for all other unit staff remain 
unchanged.
    Licensed operator qualifications and training can have an 
indirect impact on a margin of safety. However, the NRC considered 
this impact during the rulemaking process, and by promulgation of 
the revised 10 CFR [Part] 55 [rule] determined that this impact 
remains acceptable when licensees maintain a licensed operator 
training program that is accredited and based on a systems approach 
to training. As noted previously, Palo Verde's licensed operator 
training program is accredited by INPO and is based on a systems 
approach to training.

[[Page 34663]]

    The NRC has concluded, as stated in NUREG-1262, ``Answers to 
Questions at Public Meetings Regarding Implementation of Title 10, 
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that 
the standards and guidelines applied by INPO in their training 
accreditation program are equivalent to those put forth or endorsed 
by the NRC. As a result, maintaining an INPO accredited, systems 
approach based licensed operator training program is equivalent to 
maintaining [an] NRC approved licensed operator training program 
which conform[s] with applicable NRC Regulatory Guides or NRC 
endorsed industry standards. The margin of safety is maintained by 
virtue of maintaining an INPO accredited licensed operator training 
program.
    In addition, the NRC has published NRC Regulatory Issue Summary 
2001-01, ``Eligibility of Operator License Applicants,'' dated 
January 18, 2001, ``to familiarize addressees with the NRC's current 
guidelines for the qualification and training of reactor operator 
(RO) and senior operator (SO) license applicants.'' The document 
again acknowledges that the INPO National Academy for Nuclear 
Training (NANT) guidelines for education and experience, outline 
acceptable methods for implementing the NRC's regulations in this 
area.
    Therefore, there is no change in the analysis results and the 
proposed amendment request does not involve a significant reduction 
in a margin of safety.

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

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

    Date of amendment request: December 10, 2002.
    Description of amendment request: The proposed amendment would 
revise Technical Specification (TS) 4.6.E, ``Jet Pump Surveillance 
Requirements'' and its Bases. Specifically, Notes 1 and 2 would be 
added to the surveillance to provide clarity for performing the 
surveillance under the designated condition. The proposed change would 
also modify the applicability of the surveillance. Additionally, the 
condition for flow imbalance of the two recirculation loops would be 
changed from 15% to 10%. A reference in TS 4.11.C.1 to the bases for 
Specification 3.3.B.5 would also be changed to reference TS Table 
3.2.C.1, Note 5.
    Basis for proposed no significant hazards consideration 
determination: As required by title 10 of the Code of Federal 
Regulations (10 CFR), section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed Pilgrim TS 4.6.E imposes more restrictive 
surveillance requirements in accordance with the Standard Technical 
Specifications (STS) surveillance requirement 3.4.3.1 to ensure jet 
pump integrity during startup and run modes. The more restrictive 
conditions are: the recirculation loops have a flow imbalance of 
less than 10%, instead of the current 15%, when the pumps are 
operated at the same speed, and the occurrence of two of three 
conditions, instead of the simultaneous occurrence of all three 
conditions currently specified in TS 4.6.E for jet pump integrity.
    The proposed more restrictive surveillance requirements ensure 
safe operation of the plant during startup and run modes. The 
requirements are not accident precursors. The proposed change that 
corrects a reference in Surveillance 4.11.C.1 is an administrative 
change with no impact on safety. These changes do not create 
accident conditions or increase the probability of previously 
evaluated accidents. The proposed changes provide additional 
assurance that the assumptions (i.e., jet pump integrity) are met. 
Therefore, the probability or the consequences of an accident 
previously evaluated are 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 changes do not involve a change to the plant design 
or a new mode of equipment operation. As a result, the proposed 
changes do not affect parameters or conditions that could contribute 
to the initiation of any new or different kind of accident. 
Therefore, these proposed changes do not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the change involve a significant reduction in a margin 
of safety?
    Response: No.
    The proposed surveillance requirements increase the margin of 
safety by providing additional assurance of jet pump integrity. The 
proposed change to correctly reference the existing Specification is 
administrative in nature. Therefore, the proposed changes do not 
involve a significant reduction in the margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: J. M. Fulton, Esquire, Assistant General 
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth, 
Massachusetts 02360-5599.
    NRC Section Chief: James W. Clifford.

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

    Date of amendment request: March 19, 2003.
    Description of amendment request: The proposed amendment would 
delete Technical Specification (TS) 5.5.3, ``Post Accident Sampling,'' 
requirements to maintain a Post-Accident Sampling System (PASS). 
Licensees were generally required to implement PASS upgrades as 
described in NUREG-0737, ``Clarification of TMI [Three Mile Island] 
Action Plan Requirements,'' and Regulatory Guide 1.97, 
``Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess 
Plant and Environs Conditions During and Following an Accident.'' 
Implementation of these upgrades was an outcome of the lessons learned 
from the accident that occurred at TMI, Unit 2. Requirements related to 
PASS were imposed by an Order for many facilities and were added to, or 
included in, the TSs for nuclear power reactors currently licensed to 
operate. Lessons learned and improvements implemented over the last 20 
years have shown that the information obtained from PASS can be readily 
obtained through other means, or is of little use in the assessment and 
mitigation of accident conditions.
    The changes are based on NRC-approved Technical Specification Task 
Force (TSTF) Standard Technical Specification Change Traveler, TSTF-
413, ``Elimination of Requirements for a Post Accident Sampling System 
(PASS).'' The U.S. Nuclear Regulatory Commission (NRC) staff issued a 
notice of opportunity for comment in the Federal Register on December 
27, 2001 (66 FR 66949), on possible amendments concerning TSTF-413, 
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 March 20, 2002 (67 FR 13027). 
The licensee affirmed the

[[Page 34664]]

applicability of the following NSHC determination in its application 
dated March 19, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR) section 50.91(a), 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 PASS was originally designed to perform many sampling and 
analysis functions. These functions were designed and intended to be 
used in post accident situations and were put into place as a result 
of the TMI-2 accident. The specific intent of the PASS was to 
provide a system that has the capability to obtain and analyze 
samples of plant fluids containing potentially high levels of 
radioactivity, without exceeding plant personnel radiation exposure 
limits. Analytical results of these samples would be used largely 
for verification purposes in aiding the plant staff in assessing the 
extent of core damage and subsequent offsite radiological dose 
projections. The system was not intended to and does not serve a 
function for preventing accidents and its elimination would not 
affect the probability of accidents previously evaluated.
    In the 20 years since the TMI-2 accident and the consequential 
promulgation of post accident sampling requirements, operating 
experience has demonstrated that a PASS provides little actual 
benefit to post accident mitigation. Past experience has indicated 
that there exists in-plant instrumentation and methodologies 
available in lieu of a PASS for collecting and assimilating 
information needed to assess core damage following an accident. 
Furthermore, the implementation of Severe Accident Management 
Guidance (SAMG) emphasizes accident management strategies based on 
in-plant instruments. These strategies provide guidance to the plant 
staff for mitigation and recovery from a severe accident. Based on 
current severe accident management strategies and guidelines, it is 
determined that the PASS provides little benefit to the plant staff 
in coping with an accident.
    The regulatory requirements for the PASS can be eliminated 
without degrading the plant emergency response. The emergency 
response, in this sense, refers to the methodologies used in 
ascertaining the condition of the reactor core, mitigating the 
consequences of an accident, assessing and projecting offsite 
releases of radioactivity, and establishing protective action 
recommendations to be communicated to offsite authorities. The 
elimination of the PASS will not prevent an accident management 
strategy that meets the initial intent of the post-TMI-2 accident 
guidance through the use of the SAMGs, the emergency plan (EP), the 
emergency operating procedures (EOP), and site survey monitoring 
that support modification of emergency plan protective action 
recommendations (PARs).
    Therefore, the elimination of PASS requirements from Technical 
Specifications (TS) (and other elements of the licensing bases) does 
not involve a significant increase in the consequences of any 
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 elimination of PASS related requirements will not result in 
any failure mode not previously analyzed. The PASS was intended to 
allow for verification of the extent of reactor core damage and also 
to provide an input to offsite dose projection calculations. The 
PASS is not considered an accident precursor, nor does its existence 
or elimination have any adverse impact on the pre-accident state of 
the reactor core or post accident confinement of radioisotopes 
within the containment building.
    Therefore, this change does not create the possibility of a new 
or different kind of accident from any previously evaluated.

Criterion 3--The Proposed Change Does Not Involve a Significant 
Reduction in the Margin of Safety

    The elimination of the PASS, in light of existing plant 
equipment, instrumentation, procedures, and programs that provide 
effective mitigation of and recovery from reactor accidents, results 
in a neutral impact to the margin of safety. Methodologies that are 
not reliant on PASS are designed to provide rapid assessment of 
current reactor core conditions and the direction of degradation 
while effectively responding to the event in order to mitigate the 
consequences of the accident. The use of a PASS is redundant and 
does not provide quick recognition of core events or rapid response 
to events in progress. The intent of the requirements established as 
a result of the TMI-2 accident can be adequately met without 
reliance on a PASS.
    Therefore, this change does not involve a significant reduction 
in the margin of safety.

    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: James W. Clifford.

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 amendment request: May 8, 2003.
    Description of amendment request: The proposed amendment would 
change Technical Specification (TS) 3.3.6.1, ``Primary Containment and 
Drywell Isolation Instrumentation,'' to add a note allowing 
intermittent opening of penetration flow paths, under administrative 
control, that are isolated to comply with TS ACTIONS and to revise the 
operability requirement for the Reactor Core Isolation Cooling (RCIC) 
steam supply line low pressure isolation instrumentation to be 
consistent with the RCIC system operability requirements.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change to adopt TSTF [Technical Specification Task 
Force]-306 allows primary containment and drywell isolation valves 
to be unisolated under administrative controls when the associated 
isolation instrumentation is not operable. The isolation function is 
an accident mitigating function and is not an initiator of an 
accident previously evaluated. Administrative controls are required 
to be in effect when the valves are unisolated so that the 
penetration can be rapidly isolated when the need [for isolation] is 
indicated. Therefore the probability or consequences of previously 
evaluated accidents are not significantly increased.
    The proposed change also allows the RCIC turbine steam line low 
pressure containment isolation instrumentation to be inoperable 
during low startup operating pressures. These instruments primarily 
provide automatic isolation when steam line pressure is too low for 
RCIC turbine operation. The low pressure automatic isolation feature 
will only be unavailable during the time that the RCIC system is not 
required to be operable. Therefore the change does not adversely 
affect the ability of the RCIC system to perform its safety 
function.
    The RCIC steam line low pressure instruments also provide a 
diverse signal to indicate a possible system break. Even though the 
low pressure automatic isolation function will not be available for 
a short period during plant startup, the likelihood of a steam line 
break during the short period of time is low due to the low 
operating pressure. In addition, the safety function of providing 
containment integrity is maintained since there are other diverse 
leak detection instruments as well as other barriers or isolation 
capabilities that provide the isolation function.
    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.

[[Page 34665]]

    The proposed change does not involve any physical alteration of 
plant equipment and does not change the method by which any safety-
related system performs its function. The TS currently allow[s] 
containment and drywell isolation valves to be open under 
administrative controls after being closed to comply with TS ACTIONS 
for inoperable valves. Extending this allowance to the supporting 
instrumentation does not introduce any new method of isolation that 
has not already been evaluated.
    Allowing the RCIC turbine steam line low pressure isolation 
instrumentation to be inoperable during low startup operating 
pressures does not create the possibility of any new failure modes 
other than those previously evaluated. No new or different type of 
equipment will be installed. There are no new failure mechanisms or 
accident initiators introduced. The low pressure isolation is 
designed to terminate RCIC turbine operation at low steam pressures 
for equipment protection. However, this function is not required 
since the RCIC system is not required to be operable and the same 
function is accomplished by maintaining the turbine trip/throttle 
valve closed. The low pressure isolation function will continue to 
be required when the RCIC system is required to be operable.
    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 change to allow containment and drywell isolation valves to 
be unisolated under administrative control does not reduce any 
margins to safety since the proposed allowance for the supporting 
isolation instrumentation is no less restrictive than the allowance 
for the equipment it supports. When the valves are unisolated, the 
design basis function of containment isolation is maintained by 
administrative controls.
    The change to allow the RCIC turbine steam line low pressure 
isolation instrumentation to be inoperable during low startup 
operating pressures does not reduce any margins to safety. The 
current bounding analysis for a steam line break outside of 
containment remains bounding for a[n] RCIC steam break at lower 
pressures. In addition, the current high energy line break 
evaluations and subcompartment pressurization evaluations remain 
bounding for the low pressure condition. The design basis functions 
of containment isolation and containment integrity are maintained by 
the diverse leak detection instruments as well as other barriers or 
isolation capabilities that provide the isolation function.
    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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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 amendment request: May 12, 2003.
    Description of amendment request: The proposed amendment would 
change the Technical Specifications (TS) to remove the MODE 
restrictions for performance of Surveillance Requirement (SR) 3.8.4.7 
and SR 3.8.4.8 for the Division 3 direct current electrical power 
subsystem.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The power supplied by the battery is used only as a source of 
control and motive power for the HPCS [High Pressure Core Spray] 
system logic, HPCS diesel-generator set control and protection, and 
other Division 3 related controls. The loads supplied by this system 
are only loads associated with Division 3 of the Emergency Core 
Cooling Systems (ECCS).
    The battery testing period is within the period of time that the 
system is scheduled to be out of service for other planned 
maintenance. The battery test does not increase unavailability of 
the supported system or represent any change in risk above the 
current practice of planned system maintenance outages as currently 
allowed by the TS. Any risk associated with the testing of the 
Division 3 batteries will be enveloped by the risk management of the 
system outage.
    The out of service condition is controlled and evaluated for 
safety implications in accordance with 10 CFR 50.65. The HPCS system 
reliability and availability are monitored and evaluated in 
relationship to Maintenance Rule goals to ensure that total outage 
times do not degrade operational safety over time.
    Therefore, the proposed change will have no effect on the 
probability or consequences of any previously evaluated accident.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    This request involves the testing of the HPCS battery on-line 
while the system is already out of service. The testing will not add 
additional out of service time. Testing during this period has no 
influence on, nor does it contribute in any way to, the possibility 
of a new or different kind of accident or malfunction from those 
previously analyzed. The method of performing the test is not 
changed. No new accident modes are created by testing during the 
period when the system is already unavailable. Because the system is 
already out of service, no safety-related equipment or safety 
functions are altered as a result of this change.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The battery testing will be performed when the HPCS system is 
already out of service for maintenance. The out of service condition 
is controlled and evaluated for safety implications in accordance 
with 10 CFR 50.65. The batteries are not expected to be unavailable 
for more than 24 hours. This testing period is within the period of 
time that the system is scheduled to be out of service for other 
planned maintenance. Therefore, the battery test does not increase 
unavailability of the supported system or represent any change in 
risk above the current practice of planned system maintenance 
outages as currently allowed by the TS. Timing of this test has no 
effect on any fission product barrier.
    Therefore, the 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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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 amendment request: May 12, 2003.
    Description of amendment request: The proposed amendment would 
change the Technical Specification (TS) 3.3.6.1, ``Primary Containment 
and Drywell Isolation Instrumentation,'' to add a provision to the 
APPLICABILITY requirement specified in Table 3.3.6.1-1, to eliminate 
the requirement that the instrumentation for the Residual Heat Removal 
(RHR) System Isolation

[[Page 34666]]

Function on Reactor Vessel Water Level-Low, Level 3, be OPERABLE during 
certain conditions in MODE 5. Specifically, the proposed change would 
remove the requirement when the upper containment reactor cavity is at 
the High Water Level condition specified in TS 3.5.2, ``Emergency Core 
Cooling Systems (ECCS) Shutdown.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the 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 revises the applicability requirement for 
the Residual Heat Removal (RHR) System Isolation function of the 
Primary Containment and Drywell Isolation Instrumentation during 
MODE 5. The change removes the requirement that the instrumentation 
be operable during certain conditions during refueling outages. The 
function is intended to mitigate reactor vessel draindown events. 
Although draindown events during refueling operations are not 
specifically evaluated in the Updated Final Safety Analysis Report 
(UFSAR), these events were evaluated in support of licensing actions 
for the Alternate Decay Heat Removal System (ADHRS). The probability 
that a draindown event will be initiated is unrelated to operability 
requirement for this instrumentation or the associated isolation 
valves. The evaluation supporting this change determined that 
mitigating actions can be taken to terminate all postulated 
draindown events prior to fuel uncovery. As a result, the 
probability of draindown events causing fuel uncovery and the 
potential for radiological releases has not significantly increased. 
The operation or failure of the shutdown cooling suction isolation 
does not contribute to the occurrence of an accident. No active or 
passive failure mechanisms that could lead to an accident are 
affected by the proposed change.
    The consequences of a vessel drainage event are not 
significantly increased by the proposed change. Entergy [Entergy 
Operations, Inc.] has evaluated various draindown and pumpdown 
events through the shutdown cooling flow path and determined that 
adequate time is available for operations personnel to identify and 
take action to mitigate such events such that adequate core cooling 
is maintained and a radiological release does not occur.
    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.
    Entergy has evaluated various draindown events through the 
shutdown cooling flow path and determined that adequate time is 
available for operations personnel to identify and take action to 
mitigate any events such that adequate core cooling is maintained. 
With the containment refueling cavity flooded, sufficient inventory 
is available to allow operator action to terminate the inventory 
loss prior to reaching a low water level in the reactor. Installed 
equipment is not operated in a new or different manner, no new or 
different system interactions are created, and no new processes are 
introduced. No new failures have been created by the proposed 
changes.
    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 changes do not introduce any new setpoints at which 
protective or mitigative actions are initiated. No current setpoints 
are altered by this change. The design and functioning of the 
containment and drywell isolation function is also unchanged. The 
change simply modifies the applicability of the Technical 
Specifications (TS) by removing the requirement that the RHR system 
isolation on low reactor vessel level be operable with the upper 
containment cavity flooded in MODE 5. During MODE 5, the RHR system 
isolation mitigates postulated draindown events through the RHR 
system. Entergy has evaluated various draindown events through this 
flow path and determined that adequate time is available for 
operations personnel to identify and take action to mitigate such 
events such that adequate core cooling is 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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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 amendment request: May 12, 2003.
    Description of amendment request: The proposed amendment would 
change 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 (15 years) for 
performance of the next ILRT and DWBT.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed amendment to TS 5.5.12 adds a one-time extension to 
the current interval for Type A testing (i.e., the ILRT) and the 
DWBT. The current interval of ten years, based on past performance, 
would be extended on a one-time basis to 15-years from the date of 
the last test. The proposed extension to the Type A test cannot 
increase the probability of an accident since there are no design or 
operating changes involved and the test is not an accident 
initiator. The proposed extension of the test interval does not 
involve a significant increase in the consequences since research 
documented in NUREG-1493, ``Performance Based Containment Leak Rate 
Test Program,'' has found that, generically, fewer than 3% of the 
potential containment leak paths are not identified by Type B and C 
testing. A risk evaluation of the interval extension for GGNS [Grand 
Gulf Nuclear Station, Unit 1] is consistent with these results. In 
addition, the testing and containment inspections also provide a 
high degree of assurance that the containment will not degrade in a 
manner detectable only by a Type A test. Inspections required by the 
Maintenance Rule (10 CFR 50.65) and by the American Society of 
Mechanical Engineers Boiler and Pressure Vessel Code are performed 
to identify containment degradation that could affect leak 
tightness.
    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 extension to the interval for the Type A test does 
not involve any design or operational changes that could lead to a 
new or different kind of accident from any accidents previously 
evaluated. The tests are not being modified, but are only being 
performed after a longer interval. The proposed change does not 
involve a physical alteration of the plant (no new or different type 
of equipment will be installed) or a change in the methods governing 
normal plant operation.
    Therefore, the proposed change does not create the possibility 
of a new or

[[Page 34667]]

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 generic study of the increase in the Type A test interval, 
NUREG-1493, concluded there is an imperceptible increase in the 
plant risk associated with extending the test interval out to twenty 
years. The evaluations done in support of this change confirm that 
(conclusion). Further, the extended test interval would have a 
minimal effect on this risk since Type B and C testing detects 97% 
of potential leakage paths. For the requested change in the GGNS 
ILRT/DWBT interval, it was determined that the risk contribution of 
leakage will increase 0.99%. This change is considered very small 
and does not represent a significant reduction in the margin of 
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: Nicholas S. Reynolds, Esquire, Winston and 
Strawn, 1400 L Street, NW., 12th Floor, Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 18, 2003.
    Description of amendment request: The proposed amendments would 
revise Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will 
modify TS Table 3.3.6.1-1, ``Primary Containment Isolation 
Instrumentation,'' to add the requirement to perform a Channel Check in 
accordance with Surveillance Requirement (SR) 3.3.6.1.1 to thirteen 
listed instrument functions. The proposed change is the result of the 
replacement of existing plant equipment with equipment that has the 
capability of permitting the performance of a Channel Check with the 
plant in MODE 1, 2, and 3. The proposed change is consistent with the 
wording specified in NUREG-1434, ``Standard Technical Specifications 
General Electric Plants, BWR/6,'' Revision 2, dated June 2001.
    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 probability or consequences of an accident previously evaluated.
    The proposed change to Technical Specifications (TS) Table 
3.3.6.1-1, ``Primary Containment Isolation Instrumentation'' will 
incorporate into the LaSalle County Station (LSCS) TS, wording 
specified in NUREG-1434, ``Standard Technical Specifications General 
Electric Plants, BWR/6,'' Revision 2, dated June 2001. The proposed 
change will modify TS Table 3.3.6.1-1 to add the requirement to 
perform a Channel Check in accordance with Surveillance Requirement 
(SR) 3.3.6.1.1 to thirteen listed instrument functions. The 
performance of TS surveillance testing is not a precursor to any 
accident previously evaluated. A Channel Check is a monitoring 
activity that does not represent an accident initiator. Thus, the 
proposed change does not have any effect on the probability of an 
accident previously evaluated.
    The function of instrumentation listed on TS Table 3.3.6.1-1, in 
combination with other accident mitigation features, is to limit 
fission product release during and following postulated Design Basis 
Accidents (DBAs) to within limits. The surveillance testing 
specified in TS Table 3.3.6.1-1 will provide assurance that the 
instrumentation will perform as designed. Thus, the radiological 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The failure modes of the new instrumentation 
do not give rise to a new or different kind of accident. The 
proposed change does not introduce any new modes of system operation 
or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    The leak detection system at LaSalle County Station uses ambient 
or differential temperature increases to detect small primary 
coolant boundary leaks in the Main Steam Line Tunnel and in various 
rooms of the Reactor Core Isolation Cooling (RCIC) System and the 
Reactor Water Cleanup (RWCU) System. The existing thermocouple 
monitors did not have the capability to allow a Channel Check to be 
performed without undue risk of initiating an inadvertent system 
isolation in MODE 1, 2 and 3. Thus, the LSCS TS took exception to 
the guidance contained in NUREG-1434 and did not specify on TS Table 
3.3.6.1-1 that a SR 3.3.6.1.1 Channel Check be performed on the 
above listed thirteen instrument functions.
    The new thermocouple monitors have continuously reading digital 
displays that permit the performance of a Channel Check with the 
Unit in MODE 1, 2 and 3 without risk of inadvertent system 
isolations. The new thermocouple digital displays have been 
installed on Unit 2 during the January/February 2003 refuel outage 
and are scheduled to be installed in Unit 1 during the upcoming 
January 2004 refuel outage. LSCS after the return to service of Unit 
2 in March of 2003, verified that the thermocouple digital displays 
do permit a Channel Check to be successfully performed on the above 
listed thirteen instrument functions. Therefore, LSCS is requesting 
that TS Table 3.3.6.1-1 is modified to specify that a SR 3.3.6.1.1 
Channel Check be performed in MODE 1, 2 and 3, consistent with the 
guidance contained in NUREG-1434, Rev. 2.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the above, Exelon Generation 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: April 18, 2003.
    Description of amendment request: The proposed amendments would 
revise Appendix A, Technical Specifications (TS), of Facility Operating 
License Nos. NPF-11 and NPF-18. Specifically, the proposed change will 
modify TS Surveillance Requirement (SR) 3.6.1.3.8 to identify that the 
specified testing requirement is applicable to reactor instrumentation 
lines. The proposed change is consistent with the SR wording specified 
in

[[Page 34668]]

NUREG-1433, ``Standard Technical Specifications General Electric 
Plants, BWR/4,'' Revision 2, dated June 2001.
    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 probability or consequences of an accident previously evaluated.
    The proposed change to the Technical Specifications (TS) 
Surveillance Requirement (SR) 3.6.1.3.8 will incorporate into the 
SR, wording specified in NUREG-1433, ``Standard Technical 
Specifications General Electric Plants, BWR/4,'' Revision 2, dated 
June 2001. The proposed change will specify that the testing 
required by SR 3.6.1.3.8 is applicable to reactor instrumentation 
line excess flow check valves (EFCVs). The performance of TS 
surveillance testing is not a precursor to any accident previously 
evaluated. Thus, the proposed change does not have any affect on the 
probability of an accident previously evaluated.
    The function of reactor instrumentation line EFCVs, in 
combination with other accident mitigation features, is to limit 
fission product release. The surveillance testing specified in SR 
3.6.1.3.8 will provide assurance that the reactor instrumentation 
line EFCVs will perform as designed. Thus, the radiological 
consequences of any accident previously evaluated are not increased.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed change does not affect the control parameters 
governing unit operation or the response of plant equipment to 
transient conditions. The proposed change does not introduce any new 
equipment, modes of system operation or failure mechanisms.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any previously 
evaluated.
    3. The proposed changes do not involve a significant reduction 
in a margin of safety.
    NUREG-1433, Rev. 2, provided licensees with the latest NRC 
recommended content and format for TS. The NUREG-1433 SR for testing 
EFCVs, SR 3.6.1.3.10, specifies that this testing is associated with 
reactor instrumentation line EFCVs. The Bases to SR 3.6.1.3.10 in 
NUREG-1433, Rev. 2, provides a reference to NEDO-32977-A, ``Excess 
Flow Check Valve Testing Relaxation,'' dated June 2000. NEDO-32977-A 
was approved for use by licensees in a NRC letter dated March 14, 
2000. NEDO-32977-A states the following on the scope of TS testing 
associated with EFCVs:

    EFCVs in instrument lines which connect to the reactor coolant 
pressure boundary (RCPB) are normally tested during refueling 
outages to meet Technical Specification requirements. Instrument 
lines that connect to the containment atmosphere, such as those 
which measure drywell pressure, or monitor the containment 
atmosphere or suppression pool water level, are considered 
extensions of primary containment. A failure of one of these 
instrument lines during normal operation would not result in the 
closure of the associated EFCV, since normal operating containment 
pressure is not sufficient to operate the valve. Such EFCVs will 
only close with a downstream line break concurrent with a Loss of 
Coolant Accident (LOCA). Since these conditions are beyond the plant 
design basis, EFCV closure is not needed and containment atmospheric 
instrument line EFCVs need not be tested.
    The proposed change will incorporate the wording from NUREG-1433 
into LaSalle County Station SR 3.6.1.3.8 to limit the scope of TS 
required testing to EFCVs that are directly connected to the RCPB.
    Therefore, the proposed changes do not involve a significant 
reduction in a margin of safety.
    Based upon the above, Exelon Generation 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 
requested amendments involve no significant hazards consideration.
    Attorney for licensee: Mr. Edward J. Cullen, Deputy General 
Counsel, Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: May 14, 2003.
    Description of amendment request: The proposed amendment would 
modify the Technical Specifications by allowing entry into Mode 3 
operation (shutdown with reactor coolant system temperature equal to or 
greater than 280 degrees Fahrenheit) during the current outage only 
with neither high pressure injection (HPI) pump capable of taking 
suction from the low pressure injection system trains when aligned for 
containment sump recirculation. The HPI system will otherwise be 
operable.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided their 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 allows the plant to operate in Mode 3 in 
support of RCS [reactor coolant system] leakage inspection 
activities conducted during the ongoing Thirteenth Refueling Outage, 
utilizing a limited exception to Limiting Condition for Operation 
(LCO) 3.5.2. This LCO applies in plant operational Modes 1 (Power 
Operation), 2 (Startup), and 3 (Hot Standby). Under the proposed 
exception, for entry into Mode 3, both HPI trains would be required 
to be operable except for the capability of maintaining suction from 
the containment emergency sump during the recirculation phase.
    The ability of the HPI pumps to draw suction from the 
containment emergency sump (via the LPI [low pressure injection] 
pumps) is a design feature credited by the Davis-Besse Nuclear Power 
Station Updated Safety Analysis Report (USAR) for mitigation of 
various types of loss-of-coolant accidents (LOCAs). Due to the 
potential susceptibility to damage from debris contained in the 
pumped fluid, the existing HPI pumps may not be capable of 
maintaining suction from the containment emergency sump without an 
increased probability for malfunction. However, the current plant 
conditions are unique in that decay heat generation rate in the 
reactor core is extremely low due to the fact that the plant has not 
operated in more than 14 months and 76 unirradiated fuel assemblies 
have been loaded into the core, replacing irradiated fuel 
assemblies.
    A LOCA evaluation has been performed considering the current 
reactor core decay heat generation rate. The evaluation shows that 
in the unlikely event that a LOCA did occur while operating in Mode 
3 under the proposed exception, the accident can be mitigated 
without crediting HPI flow during the recirculation phase, while 
crediting additional operator actions not presently credited in the 
USAR. In addition, a risk evaluation has been performed and shows 
that the increase in core damage frequency, accounting for human 
error probability for the additional operator actions, is very 
small. Also, in the unlikely event that a LOCA did occur while 
operating in Mode 3 under the proposed exception, radiological 
consequences would be very small compared to the accident analyses 
results of record, given the fission product decay over the extended 
plant shutdown. Therefore, the proposed change would 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.

[[Page 34669]]

    There are no new or different accident initiators introduced by 
the proposed change to allow the plant to operate n Mode 3 under a 
limited exception, with the HPI pumps not capable of maintaining 
suction from the containment emergency sump (via the LPI pumps) 
during the recirculation phase of a LOCA. 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 change allows the plant to operate in Mode 3 under 
a limited exception, with the HPI pumps not capable of maintaining 
suction from the containment emergency sump (via the LPI pumps) 
during the recirculation phase of a LOCA. Although the ability of 
the HPI pumps to draw suction from the containment emergency sump 
(via the LPI pumps) is a design feature credited by the Davis-Besse 
Nuclear Power Station USAR for mitigation of various types of LOCAs, 
an evaluation shows that given the extremely low decay heat 
generation rate in the reactor core under current plant conditions, 
and crediting additional operator actions, in the unlikely event 
that a LOCA did occur while operating in Mode 3 under the proposed 
exception, the accident can be mitigated without crediting HPI flow 
during the recirculation phase. In addition, a risk evaluation has 
been performed and shows that the increase in core damage frequency, 
accounting for human error probability for the additional operator 
actions, would be expected to be very small. Also, in the unlikely 
event that a LOCA did occur while operating in Mode 3 under the 
proposed exception, radiological consequences would be very small 
compared to the accident analyses results of record, given the 
fission product decay over the extended plant shutdown. Accordingly, 
given that accident severity or consequences will not be 
significantly increased under the proposed change, a significant 
reduction in a margin of safety is not involved.

    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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: May 19, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) by removing the designation of 
safety grade as a description of the flow indication for the motor 
driven feedwater pump system. The licensee inadvertently requested that 
the flow indication be designated as safety grade in an amendment 
request that was approved as license Amendment No. 193.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided their 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 corrects a post modification and repair 
Surveillance Requirement for the Motor Driven Feedwater Pump System. 
This surveillance is not an initiator to any accident previously 
evaluated. Consequently, the probability of an accident previously 
evaluated is not significantly increased. The Technical 
Specifications continue to require the MDFP System to be operable 
and capable of performing its design function. As a result, the 
consequences of any accident previously evaluated are not 
significantly affected. 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 correction does not involve a physical alteration 
of the plant (no new or different type of equipment will be 
installed) or a change in the methods governing normal plant 
operation. Thus, this change does not create the possibility of a 
new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed correction does not result in a significant 
reduction in the margin of safety. The corrected Surveillance 
Requirement continues to ensure that the Motor Driven Feedwater Pump 
System can perform its required function. Thus, appropriate 
equipment continues to be tested in a manner that provides 
confidence that the equipment can perform its assumed function. 
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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse 
Nuclear Power Station, Unit 1, Ottawa County, Ohio

    Date of amendment request: May 21, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications (TS) by relocating to the 
licensee's Technical Requirements Manual the TS surveillance 
requirement pertaining to flow balance testing of the emergency core 
cooling system (ECCS) high pressure injection and low pressure 
injection subsystems following system modifications that alter 
subsystem flow characteristics. Also, the proposed amendment would add 
an ECCS pump operability requirement to the TS consistent with NUREG-
1430, Standard Technical Specifications-Babcock and Wilcox Plants, 
Revision 2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided their 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 surveillance requirement relocation and replacement 
does not alter the design, operation, or testing of any structure 
system or component. No previously analyzed accident scenario is 
changed. Initiating conditions and assumptions remain as previously 
analyzed. 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 surveillance requirement relocation and replacement 
does not alter the design, operation, or testing of any structure 
system or component. No new or different accident initiators are 
created as a result of the proposed changes. Therefore, the proposed 
changes do not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.

[[Page 34670]]

    The proposed surveillance requirement relocation and replacement 
does not reduce or adversely affect the capabilities of the ECCS. 
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: Mary E. O'Reilly, Attorney, FirstEnergy 
Corporation, 76 South Main Street, Akron, OH 44308.
    NRC Section Chief: Anthony J. Mendiola.

Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear 
Power Plant, Kewaunee County, Wisconsin; Docket No. 50-255, Palisades 
Plant, Van Buren County, Michigan; and Docket Nos. 50-266 and 50-301, 
Point Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc 
County, Wisconsin

    Date of amendment request: April 30, 2003.
    Description of amendment request: The proposed amendments would 
revise the Kewaunee Nuclear Power Plant Technical Specification (TS) 
Section 6.3, ``Plant Staff Qualifications,'' Palisades Plant TS Section 
5.3, ``Plant Staff Qualifications,'' and Point Beach Nuclear Plant TS 
5.3, ``Facility Staff Qualifications,'' to specify an exception to the 
current TS minimum qualifications. This exception requires licensed 
operators to meet the education and experience eligibility requirements 
of the National Academy for Nuclear Training (NANT) (ACAD 00-003), 
``Guidelines for Initial Training and Qualification of Licensed 
Operators,'' dated January 2000.
    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 Technical Specification (TS) amendments are 
administrative changes to clarify the current requirements for 
licensed operator qualifications and licensed operator training 
program. With these amendments, the TS continue to meet the current 
requirements of 10 CFR 55.
    Although licensed operator qualifications and training may have 
an indirect impact on accidents previously evaluated, the Nuclear 
Regulatory Commission (NRC) considered this impact during the 
rulemaking process, and by issuance of the revised 10 CFR 55 rule, 
concluded that this impact remains acceptable, as long as the 
licensed operator training programs are certified to be accredited 
and are based on a systems approach to training. NMC licensed 
operator training programs are accredited by the National Nuclear 
Accrediting Board (NNAB) and are based on a systems approach to 
training. The proposed TS amendments take credit for the NNAB 
accreditation of the licensed operator training programs. The TS 
requirements for all other facility staff qualifications remain 
unchanged.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed TS amendments are administrative changes to clarify 
the current requirements for licensed operator qualifications and 
licensed operator training programs and to conform to the revised 10 
CFR 55.
    As discussed above, although licensed operator qualifications 
and training may have an indirect impact on the possibility of a new 
or different kind of accident from any accident previously 
evaluated, the NRC considered this impact during the rulemaking 
process, and by issuance of the revised rule, concluded that this 
impact remains acceptable, as long as licensed operator training 
programs are certified to be accredited and based on a systems 
approach to training. As previously noted, NMC licensed operator 
training programs are accredited by NNAB and are based on a systems 
approach to training. The proposed TS amendments take credit for the 
NNAB accreditation of the licensed operator training programs. The 
TS requirements for all other facility staff qualifications remain 
unchanged.
    Additionally, the proposed TS amendments do not affect plant 
design, hardware, system operation, or procedures. Therefore, the 
proposed changes do not create the possibility of a new or different 
kind of accident from any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed TS amendments are administrative changes to clarify 
the current requirements applicable to licensed operator 
qualifications and licensed operator training programs. With these 
changes the TS continue to be consistent with the requirements of 10 
CFR 55. The TS qualification requirements for all other facility 
staff remain unchanged.
    Licensed operator qualifications and training can have an 
indirect impact on a margin of safety. However, the NRC considered 
this impact during the rulemaking process, and by issuance of the 
revised 10 CFR 55, determined that this impact remains acceptable, 
when licensees maintain a licensed operator training program that is 
accredited and based on a systems approach to training. As noted 
previously, NMC licensed operator training programs are accredited 
by NNAB and are based on a systems approach to training.
    The NRC has concluded, as stated in NUREG-1262, ``Answers to 
Questions at Public Meetings Regarding Implementation of Title 10, 
Code of Federal Regulations, Part 55 on Operators' Licenses,'' that 
the standards and guidelines applied by the Institute for Nuclear 
Power Operations in their training accreditation program are 
equivalent to those put forth or endorsed by the NRC. As a result, 
maintaining NNAB accredited, systems approach based, licensed 
operator training programs is equivalent to maintaining NRC approved 
licensed operator training programs, which conform to applicable NRC 
Regulatory Guides or NRC endorsed industry standards. The margin of 
safety is maintained by virtue of maintaining the NNAB accredited 
licensed operator training programs.
    In addition, the NRC published NRC Regulatory Issue Summary 
2001-01, ``Eligibility of Operator License Applicants,'' dated 
January 18, 2001, ``to familiarize addressees with the NRC's current 
guidelines for the qualification and training of reactor operator 
(RO) and senior operator (SO) license applicants.'' This document 
acknowledges that the National Academy for Nuclear Training 
guidelines for education and experience outline acceptable methods 
for implementing the NRC's regulations in this area.
    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: Bradley D. Jackson, Esq., Foley and Lardner, 
P.O. Box 1497, Madison, WI 53701-1497.
    NRC Section Chief: L. Raghavan.

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

    Date of amendment request: May 22, 2003.
    Description of amendment request: The proposed amendment would 
revise the Kewaunee Nuclear Power Plant (KNPP) operating license and 
Technical Specifications (TSs) to increase the licensed rated power by 
6.0 percent from 1673 megawatts thermal (MWt) to 1772 MWt.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the

[[Page 34671]]

licensee has provided its analysis of the issue of no significant 
hazards consideration, which is presented below:

Proposed Power Level Changes

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The stretch uprate evaluations performed included performance of 
accident analyses at uprated power parameters using approved 
methodologies. Results of these analyses continue to meet the event 
acceptance criteria. An evaluation of components and systems, 
including interface and control systems, that could be affected by 
the change in power level, were performed for the stretch power 
uprate. Components and systems will continue to function as designed 
and performance requirements for these systems will continue to be 
met. Additionally, the proposed change in power level was not found 
to initiate any accident, and therefore, does not increase the 
probability of an accident.
    Dose consequences were evaluated using the uprated power 
parameters. Acceptance criteria continue to be met. Therefore, the 
change also does not increase the consequences of an accident.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The change has no adverse effect on any safety related system and 
does not change the performance or integrity of any safety related 
system. Additionally, no new safety related equipment is being added 
or changed as a result of this proposed change in power. Therefore, 
the possibility of a new or different kind of accident is not 
created.
    3. Involve a significant reduction in the margin of safety.
    All analyses supporting the proposed uprated power condition 
continue to meet the appropriate acceptance criteria. Therefore, the 
proposed change does not involve a significant reduction in the 
margin of safety.
    Therefore, there are no significant hazards associated with the 
changes in rated power level.

Proposed Safety Limit Change

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The proposed change is an industry accepted safety limit 
applicable to the KNPP transition to Westinghouse fuel. Therefore, 
the change does not increase the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new accident scenarios, failure mechanisms or limiting single 
failures are introduced as a result of the proposed change in fuel 
centerline temperature. The change has no adverse effect on the fuel 
or the performance or integrity of the fuel. Therefore, the 
possibility of a new or different kind of accident is not created.
    3. Involve a significant reduction in the margin of safety.
    The proposed safety limit change is backed by technical 
evaluations performed by Westinghouse and experimental data. The 
limit is shown to be met as part of reload safety evaluations 
performed on a cycle specific basis. All applicable analyses 
supporting the proposed uprated power condition continue to meet the 
appropriate acceptance criteria. Therefore, the proposed change does 
not involve a significant reduction in the margin of safety.
    Therefore, there are no significant hazards associated with the 
change in the safety limit.

Engineered Safety Feature (ESF) Setting Change

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The stretch power uprate evaluations performed included 
performance of accident analyses. Results of the accident analyses 
have verified that the acceptance criteria continue to be met. 
Neither the change in the analytical limit nor the change in the TS 
setting limit changes how the system functions. Systems will 
continue to function as designed and system performance criteria 
will continue to be met. Dose consequences have also been evaluated 
at uprate conditions and doses remain within the appropriate 
acceptance criteria. Therefore, the change does not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The change has no adverse effect on any safety related system and 
does not change the performance or integrity of any safety related 
system. Additionally, no new safety related equipment is being added 
or changed as a result of the proposed change in the high-high steam 
flow TS setting limit. Therefore, the possibility of a new or 
different kind of accident is not created.
    3. Involve a significant reduction in the margin of safety.
    The results of the accident analyses demonstrate the acceptance 
criteria continue to be met. Systems will continue to function as 
designed and system performance criteria continue to be met. 
Therefore, the proposed change does not involve a significant 
reduction in the margin of safety.
    Therefore, there are no significant hazards associated with the 
change in the high-high steam flow TS setting limit.

Proposed Containment Cooling Systems Change

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    Removal of the LCO [limiting condition for operation] is 
conservative in that it eliminates relaxation of a design 
requirement for system redundancy. Deletion of the less conservative 
condition is more conservative by definition. Maintaining the system 
in a more conservative condition cannot create new challenges to 
components and systems that could adversely affect their ability to 
mitigate accident consequences or diminish the integrity of any 
fission product barrier. Therefore, the deletion of the LCO does not 
increase the probability or consequences of an accident previously 
evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    Maintaining the system in a more conservative condition does not 
adversely affect any fission product barrier, nor does it alter the 
safety function of safety related systems, structures, and 
components depended upon for accident prevention or mitigation. 
Equipment important to safety will continue to function at its 
design capacity. No new equipment is being added, replaced, or taken 
away by the deletion of the LCO. Therefore, the possibility of a new 
or different kind of accident is not created.
    3. Involve a significant reduction in the margin of safety.
    Safety analysis acceptance criteria continue to be satisfied for 
containment heat removal with deletion of this LCO. Therefore, the 
proposed change does not involve a significant reduction in the 
margin of safety.
    Therefore, there are no significant hazards associated with the 
containment cooling systems change.

Proposed Condensate Storage Tank (CST) Changes

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The stretch power uprate project evaluations performed included 
a review of the SBO [station blackout] event. Results of the 
evaluation verified that with the increase in the CST [condensate 
storage tanks] inventory, the evaluation criteria continue to be 
met. Systems will continue to function as designed and system 
performance criteria will continue to be met. Additionally, dose 
consequences have been evaluated for the power uprate and results 
remain within the appropriate acceptance criteria. Therefore, the 
changes to CST inventory do not increase the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed changes. 
The changes have no adverse effect on any safety related system and 
do not change the performance or integrity of any safety related 
system. Additionally, no new safety related equipment is being added 
or changed as a result of the proposed changes in inventory. 
Therefore, the possibility of a new or different kind of accident is 
not created.
    3. Involve a significant reduction in the margin of safety.
    The results of the SBO event review have verified that the 
analysis criteria continue to be met. Systems will continue to 
function as designed and system performance criteria

[[Page 34672]]

continue to be met. Therefore, the proposed changes do not involve a 
significant reduction in the margin of safety.
    Therefore, there are no significant hazards associated with the 
changes in CST inventory.

Proposed Auxiliary Feedwater System Changes

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The LONF accident analyses have demonstrated that the TS 
required AFW [auxiliary feedwater] trains at the minimum assumed 
flow capability provide sufficient heat removal capacity to mitigate 
the LONF accident such that acceptance criteria are satisfied. 
Single failure criteria are still met, and no physical system 
changes have been made. Dose consequences have been evaluated for 
the power uprate and the results remain within the appropriate 
acceptance criteria. Therefore, the changes to the auxiliary 
feedwater system technical specifications do not increase the 
probability or consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    No new accident scenarios, failure mechanisms, or limiting 
single failures are introduced as a result of the proposed change. 
The change has no adverse effect on any safety related system and 
does not change the performance or integrity of any safety related 
system. Additionally, no new safety related equipment is being added 
or changed as a result of these proposed changes to technical 
specifications. Therefore, the possibility of a new or different 
kind of accident is not created.
    3. Involve a significant reduction in the margin of safety.
    The LONF analysis supporting the proposed changes to technical 
specifications meets the appropriate acceptance criteria. Therefore, 
the proposed changes do not involve a significant reduction in the 
margin of safety.
    Therefore, there are no significant hazards associated with the 
auxiliary feedwater system technical specification changes.

Proposed Editorial and Administrative Changes

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated.
    The editorial and administrative changes do not affect the 
analysis performed in support of the stretch power uprate. 
Therefore, the changes do not increase the probability or 
consequences of an accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated.
    The editorial and administrative changes do not affect the 
analysis performed in support of the stretch power uprate. No new 
accident scenarios, failure mechanisms or limiting single failures 
are introduced as a result of the proposed editorial and 
administrative changes. Therefore, the possibility of a new or 
different kind of accident is not created.
    3. Involve a significant reduction in the margin of safety.
    The editorial and administrative changes do not affect the 
analysis performed in support of the stretch power uprate. 
Therefore, the proposed changes do not involve a significant 
reduction in the margin of safety.
    Therefore, there are no significant hazards associated with the 
editorial changes.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: John H. O'Neill, Jr., Esq., Shaw Pittman, 
Potts & Trowbridge, 2300 N. Street, NW, Washington, DC 20037-1128.
    NRC Section Chief: L. Raghavan.

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 amendment request: April 10, 2003.
    Description of amendment request: The amendments would modify the 
Salem Nuclear Generating Station (Salem), Unit Nos. 1 and 2, Technical 
Specifications (TSs) Table 3.3-1 ``Condition and Setpoint'' description 
for permissive P-7 to reflect the new location of pressure 
transmitters.
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration. The NRC 
staff has reviewed the licensee's analysis against the standards of 10 
CFR 50.92(c). The NRC staff's review is presented below:
    1. Does the proposed change involve a significant increase in 
probability or consequences of an accident previously evaluated?
    The proposed change to replace the words ``impulse chamber'' 
with ``steam line input'' in the descriptive text associated with 
the P-7 function of the Reactor Trip System does not involve any 
physical or design change to the P-7 function. The proposed change 
renames the turbine inlet pressure to reflect the change in turbine 
design and the new location where the pressure is sensed. Because 
the P-7 function is not affected by the proposed amendment request, 
the changes to the Salem TSs are effectively editorial in nature. 
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?
    The intent of the proposed change is to revise the description 
of the P-7 permissive as a result of changes to the design of the 
turbine. The P-7 permissive function is based on a relationship 
between first stage turbine inlet pressure and rated thermal power 
(RTP). Although the pressure sensed at the new location will be 
slightly higher, the instrument and controls logic, and all design 
basis functions that rely on the P-7 function, will remain the same. 
Therefore, the proposed change does not create the possibility of a 
new or different kind of accident than any previously evaluated.
    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    As previously stated, the proposed change is editorial in nature 
and maintains the design basis functions associated with the P-7 
permissive interlock. This is accomplished because the turbine 
pressure input to the P-7 function will continue to exhibit a 
consistent and accurate relationship to RTP following plant 
modifications. Therefore, because there will be no changes to the 
input assumptions associated with Salem's accident analysis, the 
proposed change does not involve a significant reduction in the 
margin of safety.

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

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

    Date of amendment request: April 11, 2003.
    Description of amendment request: The amendment would modify 
Surveillance Requirements and Bases regarding response time testing of 
the Engineered Safeguards System Actuation System (ESFAS) and the 
Reactor Trip System (RTS).
    Basis for proposed no significant hazards consideration 
determination: As required by Title 10 of the Code of Federal 
Regulations (10 CFR), section 50.91(a), the licensee has provided its 
analysis of the issue of no significant hazards consideration, which is 
presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    This change to the Technical Specifications does not result in a 
condition where the design, material, and construction

[[Page 34673]]

standards that were applicable prior to the change are altered. The 
same RTS and ESFAS instrumentation is being used; the time response 
allocations/modeling assumptions in the Chapter 15 analyses are 
still the same; only the method of verifying time response is 
changed. The proposed change will not modify any system interface 
and could not increase the likelihood of an accident since these 
events are independent of this change. The proposed activity will 
not change, degrade or prevent actions or alter any assumptions 
previously made in evaluating the radiological consequences of an 
accident described in the SAR [safety analysis report]. Therefore, 
the proposed amendment does not result in any increase in the 
probability or consequences of an accident previously evaluated.
    The proposed change to remove the footnote from Unit 1 
Surveillance Requirement 4.3.2.1.3 is an administrative change and 
does not result in any 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.
    This change does not alter the performance of the pressure and 
differential pressure transmitters and switches used in the plant 
protection systems. All sensors will still have response time 
verified by test before placing the sensor in operational service 
and after any maintenance that could affect response time. Changing 
the method of periodically verifying instrument response for certain 
sensors (assuring equipment operability) from time response testing 
to calibration and channel checks will not create any new accident 
initiators or scenarios. Periodic surveillance of these instruments 
will detect significant degradation in the sensor response 
characteristic. Implementation of 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 remove the footnote from Unit 1 
Surveillance Requirement 4.3.2.1.3 is an administrative change and 
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.
    This change does not affect the total system response time 
assumed in the safety analysis. The periodic system response time 
verification method for selected pressure and differential pressure 
sensors is modified to allow use of actual test data or engineering 
data. The method of verification still provides assurance that the 
total system response is within that defined in the safety analysis, 
since calibration tests will detect any degradation which might 
significantly affect sensor response time. Based on the above, it is 
concluded that the proposed license amendment does not result in a 
reduction in margin with respect to plant safety.
    The proposed change to remove the footnote from Unit 1 
Surveillance Requirement 4.3.2.1.3 is an administrative change and 
does 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: Jeffrie J. Keenan, Esquire, Nuclear Business 
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
    NRC Section Chief: James W. Clifford.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendment request: March 31, 2003.
    Description of amendment request: The proposed change would replace 
``Central Power and Light Company (CPL)'' with ``AEP Texas Central 
Company'' throughout the Operating License of each unit.
    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 administrative license amendment only changes the 
name of one of the owners of STP in the Operating Licenses. This is 
not an initiator for accidents nor does this action affect the 
consequences of an accident. 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 administrative license amendment only changes the 
name of one of the owners of STP in the Operating Licenses. This is 
not an initiator for accidents. Therefore, the 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.
    Margin of safety is associated with confidence in the ability of 
the fission product barriers (i.e., fuel and fuel cladding, reactor 
coolant pressure boundary, and containment structure) to limit the 
level of radiation dose to the public. The proposed administrative 
license amendment only changes the name of one of the owners of STP 
in the Operating Licenses. The proposed action does not affect 
margin of safety at all. Therefore, the proposed change does not 
involve a significant reduction in a margin of safety.
    Based on the above, STPNOC concludes that the proposed amendment 
involves no significant hazards consideration under the standards 
set forth in 10 CFR 50.92 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 standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
request for amendments involves no significant hazards consideration.
    Attorney for licensee: A. H. Gutterman, Esq., Morgan, Lewis & 
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

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

    Date of amendment request: April 30, 2003.
    Description of amendment request: The amendment would modify 
several surveillance requirements (SRs) in Technical Specifications 
(TSs) 3.8.1 and 3.8.4 on alternating current and direct current 
sources, respectively, for plant operation. The revised SRs would have 
notes deleted or modified to allow the SRs to be performed, or 
partially performed, in reactor modes that are currently not allowed by 
the TSs. The current SRs are not allowed to be performed in Modes 1 and 
2. Several of the current SRs also cannot be performed in Modes 3 and 
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. The proposed change does not involve a significant increase 
in the probability or consequences of an accident previously 
evaluated.
    The design of plant equipment is not being modified by the 
proposed changes. In addition, the DGs [diesel generators] and their 
associated emergency loads are accident mitigating features. As 
such, testing of the DGs themselves is not associated with any 
potential accident-initiating mechanism. Therefore, there will be no 
significant impact on any accident probabilities by the approval of 
the requested changes.
    The changes include an increase in the online time that a DG 
under test will be paralleled to the grid (for SRs 3.8.1.10 and

[[Page 34674]]

3.8.1.14). As such, the ability of the tested DG to respond to a 
design basis accident [(DBA)] could be adversely impacted by the 
proposed changes. However, the impacts are not considered 
significant based, in part, on the ability of the remaining DG to 
mitigate a DBA or provide safe shutdown. With regard to SR 3.8.1.10 
and SR 3.8.1.14, experience shows that testing per these SRs 
typically does not perturb the electrical distribution system. In 
addition, operating experience and qualitative evaluation of the 
probability of the DG or bus loads being adversely affected 
concurrent with or due to a significant grid disturbance, while the 
DG is being tested, support the conclusion that the proposed changes 
do not involve any significant increase in the likelihood of a 
safety-related bus blackout or damage to plant loads.
    The SR changes that are consistent with TSTF [Technical 
Specification Task Force]-283 have been approved by the NRC for 
submittal by licensees. The on-line tests allowed by the TSTF are 
only to be performed for the purpose of establishing OPERABILITY [of 
the DG being tested]. Performance of these SRs during restricted 
MODES will require an assessment to assure plant safety is 
maintained or enhanced.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. The proposed change does not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The capability to synchronize a DG to the offsite source (via 
the associated plant bus) and test the DG in such a configuration is 
a design feature of the DGs, including the test mode override in 
response to a safety injection signal. Paralleling the DG for longer 
periods of time during plant operation may slightly increase the 
probability of incurring an adverse effect from the offsite source, 
but this increase in probability is judged to be still quite small 
and such a possibility is not a new or previously unrecognized 
consideration.
    The proposed changes would not require any new or different 
accidents to be postulated since no changes are being made to the 
plant that would introduce any new accident causal mechanisms. This 
license amendment request does not impact any plant systems that are 
potential accident initiators; nor does it have any significantly 
adverse impact on any accident mitigating systems.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The proposed changes do not involve a significant reduction in 
the margin of safety. The margin of safety is related to the 
confidence in the ability of the fission product barriers to perform 
their design [safety] functions during and following an accident 
situation. These barriers include the fuel cladding, the reactor 
coolant system, and the containment system. The proposed changes do 
not directly affect these barriers, nor do they involve any 
significantly adverse impact on the DGs which serve to support these 
barriers in the event of an accident concurrent with a loss of 
offsite power. The proposed changes to the testing requirements for 
the plant DGs do not affect the OPERABILITY requirements for the 
DGs, as verification of such OPERABILITY will continue to be 
performed as required (except during different allowed MODES [of 
operation]). These changes have an insignificant impact on DG 
availability, as continued verification of OPERABILITY supports the 
capability of the DGs to perform their required [safety] function of 
providing emergency power to plant equipment that supports or 
constitutes the fission product barriers. Only one DG is to be 
tested at a time, so that the remaining DG will be available to 
safely shut down the plant if required. Consequently, performance of 
the fission product barriers will not be impacted by implementation 
of the proposed amendment.
    In addition, the proposed changes involve no changes to [safety] 
setpoints or limits established or assumed by the accident analyses. 
On this and the above basis, no safety margins will be impacted.
    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: Jay Silberg, Esq., Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
    NRC Section Chief: Stephen Dembek.


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

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

Exelon Generation Company, LLC, Docket Nos. 50-254 and 50-265, Quad 
Cities Nuclear Power Station, Units 1 and 2, Rock Island County, 
Illinois

    Date of amendment request: May 1, 2003, as supplemented by letter 
dated May 2, 2003.
    Brief description of amendment request: The proposed amendments 
would modify Technical Specification Surveillance Requirements to 
provide an alternative means of testing the Unit 1 main steam 
electromatic relief valves, including those that provide the automatic 
depressurization and the low set relief functions, and provide an 
alternative means for testing the Units 1 and 2 dual function Target 
Rock safety/relief valves.
    Date of publication of individual notice in Federal Register: May 
13, 2003 (68 FR 25645).
    Expiration date of individual notice: May 27, 2003.

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

[[Page 34675]]

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

Detroit Edison Company, Docket No. 50-16, Enrico Fermi Atomic Power 
Plant, Unit 1, (Fermi 1) Monroe County, Michigan

    Date of amendment request: January 28, 2003 (Reference NRC-03-
0011).
    Brief description of amendment: This amendment revises the Fermi 1, 
Technical Specifications by removing the requirements for Water 
Intrusion alarms, associated surveillances, and liquid waste tank level 
check surveillance. The sections containing Reactor Building and Fuel 
and Repair Building drains descriptions are removed in their entirety, 
clarification is added for evolutions when tritium sampling is not 
required. This amendment also removes previously deleted items and re-
numbers/letters remaining sections, and makes several editorial 
corrections.
    Date of issuance: May 16, 2003.
    Effective date: On the date of issuance of this amendment and must 
be fully implemented no later than 60-calendar days from the date of 
issuance.
    Amendment No.: 20.
    Facility Operating License No. DPR-9: Amendment revised the 
Technical Specifications by: (1) Deleting Sections A.1, 2, 4, 8, C.1, 
D, E.1, H.3.b, I.5, I.7b, I.9.d, which were previously deleted and the 
word ``Deleted'' used as a place marker to alleviate the need to 
renumber or re-letter the remaining sections. Also, the remaining 
sections were renumbered or re-lettered as appropriate. (2) Deleting 
Sections C.2 and E.2 which cover the Reactor Building and Fuel and 
Repair Building Drains. These requirements are no longer necessary in 
this phase of Fermi 1 decommissioning. (3) In Section F, the following 
words were added, ``Monitoring or sampling for tritium will not be 
required if the sample results have determined that tritium is not 
present during a given evolution.'' This wording was added to clarify 
during which evolutions resulting in radioactive gaseous effluents the 
effluents would be monitored or sampled and analyzed for tritium. (4) 
Sections H.1 and H.2, which covered water intrusion monitoring system 
alarms, including surveillances, allowed out-of-service time, 
compensatory measures and alarm readouts for alarms associated with 
water intrusion, were deleted. (5) In Section H.3 the surveillance 
requirement for radiation for the sump pump serving the reactor 
building annulus will not be required once the pump is made inactive 
and the surveillance requirement for radiation of the steam cleaning 
room access plug is deleted. In Section H.4 the requirement for a 
monthly level check of the liquid waste tanks was deleted. (6) Table H-
1, which only lists water intrusion alarms, was deleted. (7) Editorial 
changes included in this amendment are in Section I.2, the word 
``employes'' was changed to ``employees'; in Section I.2.b the word 
``He'' was changed to ``The Health Physicist'; in Section I.7 the word 
``his'' was removed from the following sentence, ``The Custodian may 
temporarily change a procedure by Written Order following his 
determination that the change does not constitute a significant 
increase in the hazards associated with the operation.'' In Section 
I.9.h the word ``usual'' will be changed to ``unusual.''
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18271). The NRC's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 16, 2003.
    No significant hazards consideration comments: None received.

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

    Date of application for amendment: May 23, 2002.
    Brief description of amendment: The amendment deletes License 
Condition 2.C.(19) of the Operating License which pertains to 
historical actions that have been met. The amendment also deletes 
Section 2.F of the Operating License which requires reporting 
violations of the requirements in Section 2.C of the Operating License. 
The reporting requirements in Section 2.F are either adequately 
addressed by the requirements of 10 CFR 50.72 and 10 CFR 50.73, or are 
not needed because more restrictive requirements are contained in the 
specific License Condition.
    In its May 23, 2003, application, the licensee also proposed to 
delete License Conditions 2.C.(20) and 2.C.(21) which pertain to 
historical actions that have been met. The Nuclear Regulatory 
Commission staff's evaluation of the proposed deletion of License 
Conditions 2.C.(20) and 2.C.(21) will be addressed under separate 
cover.
    Date of issuance: May 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days.
    Amendment No.: 155.
    Facility Operating License No. NPF-43: Amendment revises the 
Operating License.
    Date of initial notice in Federal Register: June 25, 2002 (67 FR 
42817).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated May 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of amendment request: May 13, 2002.
    Brief description of amendment: The amendment revised selected 
radiological-related technical specifications of the Millstone Unit 1 
Permanently Defueled Technical Specifications. These changes are a 
result of the revision to part 20 of title 10 of the Code of Federal 
Regulations.
    Date of issuance: May 15, 2003.
    Effective date: May 15, 2003, and shall be implemented within 120 
days from the date of issuance.
    Amendment No.: 112.
    Facility Operating License No. DPR-21: The amendment revised the 
Permanently Defueled Technical Specifications.
    Date of initial notice in Federal Register: July 23, 2002 (67 FR 
48215). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 15, 2003.
    No significant hazards consideration comments received: No.

Dominion Nuclear Connecticut, Inc., Docket Nos. 50-336 and 50-423, 
Millstone Power Station, Unit Nos. 2 and 3, New London County, 
Connecticut

    Date of application for amendments: May 13, 2002.
    Brief description of amendments: The amendments revise the 
Millstone Power Station, Unit No. 2 (MP2) and Unit No. 3 (MP3) 
Technical Specifications (TSs) changing selected MP2 and MP3 
radiological-related TSs. These changes are due to the revision to part 
20 of title 10 of the Code of Federal Regulations.
    Date of issuance: May 15, 2003.

[[Page 34676]]

    Effective date: As of the date of issuance and shall be implemented 
within 120 days from the date of issuance.
    Amendment Nos.: 276 and 215.
    Facility Operating License Nos. DPR-65 and NPF-49: These amendments 
revised the TSs.
    Date of initial notice in Federal Register: July 9, 2002 (67 FR 
45562). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 15, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application of amendments: February 17, 2003.
    Brief description of amendments: The amendments revised the 
Technical Specifications Surveillance Requirement 3.10.1.9 to increase 
the loading requirements for the Standby Shutdown Facility Diesel 
Generator from = 3000 kW to = 3280 kW.
    Date of Issuance: May 19, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 90 days from the date of issuance.
    Amendment Nos.: 331, 331, and 332.
    Renewed Facility Operating License Nos. DPR-38, DPR-47, and DPR-55: 
Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: April 1, 2003 (68 FR 
15759). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 19, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Operations, Inc., Docket No. 50-333, James A. 
FitzPatrick Nuclear Power Plant, Oswego County, New York

    Date of application for amendment: February 27, 2003.
    Brief description of amendment: The amendment deletes Technical 
Specification 5.5.3, ``Post Accident Sampling,'' and thereby eliminates 
the requirements to have and maintain the post accident sampling system 
for the James A. FitzPatrick Nuclear Power Plant.
    Date of issuance: May 16, 2003.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 278.
    Facility Operating License No. DPR-59: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18276). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: December 12, 2002, as 
supplemented on April 3, 2003 and May 2, 2003.
    Brief description of amendment: The amendment revises the Facility 
Operating License and the Technical Specifications (TSs) to increase 
the licensed core thermal power level to 3114.4 megawatts (MWt), which 
is a 1.4% increase above the currently authorized power level of 3071.4 
MWt. The power uprate is based on the improvement in the core power 
uncertainty allowance originally required for the emergency core 
cooling system (ECCS) evaluations performed in accordance with Appendix 
K, ``ECCS Evaluation Models,'' to Part 50 of Title 10 of the Code of 
Federal Regulations. Specifically, the reduced uncertainty is obtained 
by using a more accurate measurement of feedwater flow. In addition, 
changes were made to TS Sections 1.1, 2.1, 2.3, 3.1, 3.4, 6.9, and the 
applicable TS Bases to account for the change in power level.
    Date of issuance: May 22, 2003.
    Effective date: May 22, 2003.
    Amendment No.: 237.
    Facility Operating License No. DPR-26: Amendment revised the 
Technical Specifications and License.
    Date of initial notice in Federal Register: January 7, 2003 (68 FR 
00801). The April 3 and May 3, 2003, letters provided clarifying 
information that did not enlarge the scope of the original Federal 
Register notice or change the initial proposed no significant hazards 
consideration determination. The Commission's related evaluation of the 
amendment is contained in a Safety Evaluation dated May 22, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: November 22, 2002, as 
supplemented by letter dated March 13, 2003.
    Brief description of amendment: The amendment allows for a one-time 
change to revise the steam generator in-service inspection frequency 
requirements in Technical Specification 4.4.5.3.a to allow a 40-month 
inspection interval after one inspection, rather than after two 
consecutive inspections, based on the results falling into the C-1 
classification.
    Date of issuance: May 28, 2003.
    Effective date: As of the date of issuance to be implemented within 
60 days from the date of issuance.
    Amendment No.: 247.
    Facility Operating License No. NPF-6: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: December 24, 2002 (67 
FR 78520). The March 13, 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 May 28, 2003.
    No significant hazards consideration comments received: No.

Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455, 
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois; Docket Nos. 
STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1 and 2, Will 
County, Illinois

    Date of application for amendments: December 12, 2002.
    Brief description of amendments: The amendments would add a new 
Surveillance Requirement to Technical Specification Section 3.7.5, 
``Auxillary Feedwater (AF) System,'' which requires operation of the 
diesel-driven AF pump on a monthly frequency (i.e., once every 31 days) 
for greater than or equal to 15 minutes.
    Date of issuance: May 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 132/127.
    Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77: 
The amendments revised the Technical Specifications 3.7.5.
    Date of initial notice in Federal Register: February 18, 2003 (68 
FR 7817). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 28, 2003.
    No significant hazards consideration comments received: No.

[[Page 34677]]

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

    Date of application for amendments: November 27, 2002.
    Brief description of amendments: These amendments deleted TS 5.5.3, 
``Post Accident Sampling,'' and thereby eliminated the requirements to 
have and maintain the post accident sampling system for Peach Bottom 
Atomic Power Station, Units 2 and 3.
    Date of issuance: May 22, 2003.
    Effective date: As of the date of issuance, to be implemented 
within 180 days.
    Amendments Nos.: 248 and 251.
    Renewed Facility Operating License Nos. DPR-44 and DPR-56: The 
amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: January 21, 2003 (68 FR 
2802). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 22, 2003.
    No significant hazards consideration comments received: No.

Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald 
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan

    Date of application for amendments: August 30, 2002 as supplemented 
by letters dated February 27, April 7, April 29, and May 2, 2003.
    Brief description of amendments: The amendments revise the reactor 
trip system and engineered safety features actuation system 
surveillance requirements, increasing selected surveillance intervals 
for analog channels, logic cabinets, and reactor trip breakers. 
Additionally, the amendments revise the reactor trip system and 
engineered safety features actuation system surveillance requirements, 
increasing the completion time and bypass time for the reactor trip 
breakers.
    Date of issuance: May 23, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days.
    Amendment Nos.: 277 and 260.
    Facility Operating License Nos. DPR-58 and DPR-74: Amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 15, 2002 (67 FR 
63695). The supplemental letters contained clarifying information and 
did not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice. The Commission's related evaluation of the amendments 
is contained in a Safety Evaluation dated May 23, 2003.
    No significant hazards consideration comments received: No.

Nuclear Management Company, LLC, Docket No. 50-331, Duane Arnold Energy 
Center, Linn County, Iowa

    Date of application for amendment: March 11, 2003.
    Brief description of amendment: The amendment changes the operating 
license by adding a paragraph authorizing the licensee to revise the 
updated final safety analysis report by deleting the notation that the 
Nuclear Regulatory Commission does not endorse the reactor building 
crane as single-failure-proof.
    Date of issuance: May 16, 2003.
    Effective date: As of the date of issuance and shall be implemented 
no later than the update of the final safety analysis report to be 
submitted in accordance with 10 CFR 50.71(e).
    Amendment No.: 251.
    Facility Operating License No. DPR-49: The amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: April 15, 2003 (68 FR 
18278). The Commission's related evaluation of the amendment is 
contained in a Safety Evaluation dated May 16, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendments: July 10, 2002, as supplemented 
May 9, 2003.
    Brief description of amendments: The amendments change the Sequoyah 
Nuclear Plant (SQN) Technical Specifications (TSs) by modifying the 
requirements applicable when actions or other requirements direct 
suspension of activities that involve a positive reactivity change for 
the SQN TSs. The proposed change will remove the requirement to not 
make positive reactivity changes during certain conditions. The changes 
will permit limited positive reactivity changes that are necessitated 
by plant operations. These changes will limit the amount of reactivity 
changes to those that will continue to assure appropriate reactivity 
limits are met, either shutdown margin or refueling boron 
concentration, as appropriate.
    Date of issuance: May 22, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 285 and 274.
    Facility Operating License Nos. DPR-77 and DPR-79: Amendments 
revise the Technical Specifications.
    Date of initial notice in Federal Register: August 6, 2002 (67 FR 
50961). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 22, 2003.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket No. 50-327, Sequoyah Nuclear Plant, 
Units 1 and 2, Hamilton County, Tennessee

    Date of application for amendment: November 15, 2002, as 
supplemented February 28, 2003, March 14, 2003, and April 25, 2003.
    Brief description of amendment: The Amendments revise the Technical 
Specification (TS) 3.7.1.3, ``Condensate Storage Water,'' Limiting 
Condition for Operation by increasing the required minimum amount of 
stored water from 190,000 gallons to 240,000 gallons. This change is 
being made to support the replacement steam generator requirements.
    Date of issuance: May 27, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 45 days of issuance.
    Amendment Nos.: 286 and 275.
    Facility Operating License No. DPR-77 and DPR-79: Amendments revise 
the TSs.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5682). The supplemental letters provided clarifying information that 
was within the scope of the initial notice and did not change the 
initial proposed no significant hazards consideration determination. 
The Commission's related evaluation of the amendment is contained in a 
Safety Evaluation dated May 27, 2003.
    No significant hazards consideration comments received: No.

STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South 
Texas Project, Units 1 and 2, Matagorda County, Texas

    Date of amendments request: August 19, 2002.
    Brief description of amendments: The amendment revised Technical 
Specification Section 3/4.3.2, ``Engineered Safety Features Actuation 
System Instrumentation,'' to extend the interval between slave relay 
tests.

[[Page 34678]]

    Date of issuance: May 19, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance.
    Amendment Nos.: Unit 1-152 ; Unit 2-140.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 1, 2002 (67 FR 
61685). The Commission's related evaluation of the amendments is 
contained in a Safety Evaluation dated May 19, 2003.
    No significant hazards consideration comments received: No.

    Dated at Rockville, Maryland, this 2nd day of June, 2003.

    For the Nuclear Regulatory Commission.
John A. Zwolinski,
Director, Division of Licensing Project Management, Office of Nuclear 
Reactor Regulation.
[FR Doc. 03-14277 Filed 6-9-03; 8:45 am]
BILLING CODE 7590-01-P