[Federal Register Volume 69, Number 12 (Tuesday, January 20, 2004)]
[Notices]
[Pages 2735-2752]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-1104]


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NUCLEAR REGULATORY COMMISSION


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

I. Background

    Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory 
Commission (the Commission or NRC staff) is publishing this regular 
biweekly notice. Public Law 97-415 revised section 189 of the Atomic 
Energy Act of 1954, as amended (the Act), to require the Commission to 
publish notice of any amendments issued, or proposed to be issued, 
under a new provision of section 189 of the Act. This provision grants 
the Commission the authority to issue and make immediately effective 
any amendment to an operating license upon a determination by the 
Commission that such amendment involves no significant hazards 
consideration, notwithstanding the pendency before the Commission of a 
request for a hearing from any person.
    This biweekly notice includes all notices of amendments issued, or 
proposed to be issued from December 24, 2003, through January 8, 2004. 
The last biweekly notice was published on January 6, 2003 (69 FR 691).

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 February 19, 2004, the licensee may file a request for a hearing 
with respect to issuance of the amendment to the subject facility 
operating license and any person whose interest may be affected by this 
proceeding and who wishes to participate as a party in the proceeding 
must file a written request for a hearing and a petition for leave to 
intervene. Requests for a hearing and a petition for leave to intervene 
shall be filed in accordance with the Commission's ``Rules of Practice 
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested 
persons should consult a current copy of 10 CFR 2.714, which is 
available at the Commission's PDR, located at One White Flint North, 
Public File Area 01F21, 11555 Rockville Pike (first floor), Rockville, 
Maryland. Publicly available records will be accessible from the 
Agencywide Documents Access and Management System's (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing 
or petition for 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

[[Page 2736]]

Board up to 15 days prior to the first prehearing conference scheduled 
in the proceeding, but such an amended petition must satisfy the 
specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses.
    If a hearing is requested, the Commission will make a final 
determination on the issue of no significant hazards consideration. The 
final determination will serve to decide when the hearing is held.
    If the final determination is that the amendment request involves 
no significant hazards consideration, the Commission may issue the 
amendment and make it immediately effective, notwithstanding the 
request for a hearing. Any hearing held would take place after issuance 
of the amendment.
    If the final determination is that the amendment request involves a 
significant hazards consideration, any hearing held would take place 
before the issuance of any amendment.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemaking and 
Adjudications Staff, or may be delivered to the Commission's PDR, 
located at One White Flint North, Public File Area 01F21, 11555 
Rockville Pike (first floor), Rockville, Maryland, by the above date. 
Because of continuing disruptions in delivery of mail to United States 
Government offices, it is requested that petitions for leave to 
intervene and requests for hearing be transmitted to the Secretary of 
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for 
hearing and petition for leave to intervene should also be sent to the 
Office of the General Counsel, U.S. Nuclear Regulatory Commission, 
Washington, DC 20555-0001, and because of continuing disruptions in 
delivery of mail to United States Government offices, it is requested 
that copies be transmitted either by means of facsimile transmission to 
301-415-3725 or by e-mail to [email protected]. A copy of the 
request for hearing and petition for leave to intervene should also be 
sent to the attorney for the licensee.
    Nontimely filings of petitions for leave to intervene, amended 
petitions, supplemental petitions and/or requests for a hearing will 
not be entertained absent a determination by the Commission, the 
presiding officer or the Atomic Safety and Licensing Board that the 
petition and/or request should be granted based upon a balancing of 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
    For further details with respect to this action, see the 
application for amendment which is available for public inspection at 
the Commission's PDR, located at One White Flint North, Public File 
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. 
Publicly available records will be accessible from the Agencywide 
Documents Access and Management System's (ADAMS) Public Electronic 
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there 
are problems in accessing the documents located in ADAMS, contact the 
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to 
[email protected].

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

    Date of amendment request: May 1, 2003.
    Description of amendment request: The proposed amendment would 
revise the Clinton Power Station (CPS) Technical Specifications to (1) 
support an expansion of the core flow operating range, (2) implement an 
Oscillation Power Range Monitor (OPRM) Instrumentation system, and (3) 
implement the Detect and Suppress Solution--Confirmation Density 
approach to automatically detect and suppress neutronic/thermal-
hydraulic instabilities. These changes will support operation at 3,473 
megawatts thermal with core flow as low as 85 percent of rated core 
flow. The expanded operating range is identified as Maximum Extended 
Load Line Limit Analysis Plus (MELLLA+). The scope of evaluations 
required to support the expansion of the core flow operating range to 
MELLLA+ boundary is contained in the General Electric Licensing Topical 
Report (LTR) NEDC-33006P, ``Maximum Extended Load Line Limit Analysis 
Plus Licensing Topical Report.''
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. Does the proposed amendment involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    Response: No.
    The probability (frequency of occurrence) of a design basis 
accident (DBA) occurring is not affected by the operating range 
expansion, because the plant continues to comply with the regulatory 
and design basis criteria established for plant equipment. The 
MELLLA+ core operating range expansion does not require significant 
plant hardware modifications. The core operating range expansion 
involves changes to the operating power-to-flow map and a small 
number of setpoints and alarms. Because there is no change in the 
operating pressure, power, steam flow rate, or feedwater flow rate, 
there are no significant effects on the plant hardware outside of 
the Nuclear Steam Supply System (NSSS). The MELLLA+ operating range 
expansion does not cause additional requirements to be imposed on 
any of the safety, balance-of-plant, electrical, or auxiliary 
systems. No changes to the power generation and electrical 
distribution systems are required due to the introduction of 
MELLLA+. An evaluation of the probabilistic safety assessment 
concludes that the calculated increase in core damage frequencies 
due to the MELLLA+ operating range expansion are very small. Scram 
setpoints (equipment settings that initiate automatic plant 
shutdowns) are established such that there is no significant 
increase in scram frequency due to the MELLLA+ operating range 
expansion. No new challenges to safety-related equipment result from 
the MELLLA+ operating range

[[Page 2737]]

expansion. As a result, there is no significant increase in the 
probability of an accident previously evaluated.
    The proposed changes specify limiting conditions for operation, 
required actions and surveillance requirements for the OPRM system, 
and allows operation in regions of the power-to-flow map currently 
restricted by the requirements of the Interim Corrective Actions 
(ICAs) and certain limiting conditions of operation of TS Section 
3.4.1. The restrictions of the ICAs and TS Section 3.4.1 were 
imposed to ensure adequate capability to detect and suppress 
conditions consistent with the onset of thermal-hydraulic 
oscillations that may develop into a thermal-hydraulic instability 
event. A thermal-hydraulic instability event has the potential to 
challenge the Minimum Critical Power Ratio (MCPR) safety limit. The 
OPRM system can automatically detect and suppress conditions 
necessary for thermal-hydraulic instability. The Backup Stability 
Protection (BSP), in lieu of the ICAs, will provide adequate 
protection should the OPRM equipment become temporarily inoperable. 
With the activation of the OPRM system, the restrictions of the ICAs 
and TS Section 3.4.1 will no longer be required.
    The probability of a thermal-hydraulic instability event is 
impacted by power to flow conditions such that only during operation 
inside specific regions of the power-to-flow map, in combination 
with power shape and inlet enthalpy conditions, can the occurrence 
of an instability event be postulated to occur. Operation in these 
regions may increase the probability that operation with conditions 
necessary for a thermal-hydraulic instability can occur.
    When the OPRM is operable, the OPRM can automatically detect the 
imminent onset of power oscillations and generate a trip signal. 
Actuation of a Reactor Protection System (RPS) trip will suppress 
conditions necessary for thermal-hydraulic instability and decrease 
the probability of a thermal-hydraulic instability event. In the 
event the trip capability of the OPRM is not maintained, the 
proposed changes limit the period of time before an alternate method 
to detect and suppress thermal-hydraulic oscillations is required. 
Since the duration of this period of time is limited, the increase 
in the probability of a thermal-hydraulic instability event is not 
significant. Therefore, the proposed changes do not result in a 
significant increase in the probability of an accident previously 
evaluated.
    The DSS-CD solution is designed to identify power oscillations 
upon inception and initiate control rod insertion (i.e., scram) to 
terminate the oscillations prior to any significant amplitude 
growth. The DSS-CD provides protection against violation of the 
Safety Limit Minimum Critical Power Ratio (SLMCPR) for anticipated 
oscillations. Compliance with Criterion 10, ``Reactor design.'', and 
Criterion 12, ``Suppression of reactor power oscillations.'', of 
10CFR50, Appendix A, ``General Design Criteria For Nuclear Power 
Plants,'' is accomplished via an automatic action. A developing 
instability event is suppressed by the DSS-CD system with 
substantial margin to the SLMCPR and no clad damage, with the event 
terminating in a scram and never developing into an accident. The 
DSS-CD system does not interact with equipment whose failure could 
cause an accident. Scram setpoints in the DSS-CD will be established 
so that analytical limits are met. The reliability of the DSS-CD 
will meet or exceed that of the existing system. No new challenges 
to safety-related equipment will result from the DSS-CD solution. 
Because an instability event would reliably terminate in an early 
scram without impact on other safety systems, there is no 
significant increase in the probability of an accident.
    The spectrum of hypothetical accidents and transients has been 
investigated, and are shown to meet the plant's currently licensed 
regulatory criteria. In the area of core design, for example, the 
fuel operating limits such as Maximum Average Planar Linear Heat 
Generation Rate (MAPLHGR) and SLMCPR continue to be met. The fuel 
reload analyses will show plant transients meet the criteria 
accepted by the NRC as specified in NEDO-24011, ``GESTAR II,'' 
(Reference 12). Challenges to fuel are evaluated, and shown to still 
meet the criteria of 10 CFR 50.46, ``Acceptance Criteria for 
Emergency Core Cooling Systems for Light-Water Nuclear Power 
Reactors.'', 10 CFR 50 Appendix K, ``ECCS Evaluation Models,'' and 
Regulatory Guide 1.70, ``Standard Format and Content of Safety 
Analysis Reports for Nuclear Power Plants,'' Section 6.3. Challenges 
to the containment have been evaluated, and the containment and its 
associated cooling systems meet Criterion 38, ``Containment heat 
removal.'', and Criterion 50, ``Containment design basis.'', of the 
general design criteria. Radiological release events have been 
evaluated, and are shown to be below the regulatory limits of 10 CFR 
100, ``Reactor Site Criteria''. Operation in the MELLLA+ region does 
not result in an increase in the consequences of an accident 
previously evaluated. Operation within the MELLLA+ region has been 
evaluated to ensure that the CPS response to accidents and 
transients remains within acceptable criteria. Thus, the proposed 
changes do not involve a significant increase in the consequences of 
an accident previously evaluated.
    An unmitigated thermal-hydraulic instability event is postulated 
to cause a violation of the MCPR safety limit. The proposed changes 
ensure mitigation of thermal-hydraulic instability events prior to 
challenging the MCPR safety limit if initiated from anticipated 
conditions by detection of the onset of oscillations and actuation 
of an RPS trip signal when the OPRM system is operable. The OPRM 
also provides the capability of an RPS trip being generated for 
thermal-hydraulic instability events initiated from unanticipated 
but postulated conditions. These mitigative capabilities of the OPRM 
system would become available as a result of the proposed changes 
and have the potential to reduce the consequences of unanticipated 
and postulated thermal-hydraulic instability events.
    As stated above, the DSS-CD solution meets the requirements of 
Criterion 10 and Criterion 12 of the GDC by automatically detecting 
and suppressing design basis thermal-hydraulic oscillations prior to 
exceeding the fuel SLMCPR. Proper operation of the DSS-CD system 
does not affect any fission product barrier or Engineered Safety 
Feature. Thus, the proposed change cannot change the consequences of 
any accident previously evaluated.
    Therefore, the proposed changes do not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Does the proposed amendment create the possibility of a new 
or different kind of accident from any previously evaluated?
    Response: No.
    Equipment that could be affected by MELLLA+ has been evaluated 
and no new operating mode, safety related equipment lineup, accident 
scenario, or equipment failure mode was identified. The full 
spectrum of accident considerations, defined in the CPS Updated 
Safety Analysis Report (USAR), has been evaluated, and no new or 
different kind of accident has been identified. The MELLLA+ 
operating range expansion uses existing technology and NRC approved 
safety analysis methodology, and applies them within the 
capabilities of already existing plant equipment in accordance with 
presently existing regulatory and industry criteria. The MELLLA+ 
operating range expansion will not create the possibility of a new 
or different kind of accident from any accident previously 
evaluated.
    The proposed changes specify limiting conditions for operations, 
required actions and surveillance requirements of the OPRM system 
and allows operation in regions of the power-to-flow map currently 
restricted by the requirements of the ICAs and TS Section 3.4.1. The 
OPRM system uses input signals shared with the Average Range Power 
Monitor (APRM) system and rod block functions to monitor core 
conditions and generate an RPS trip when required. Quality 
requirements for software design, testing, implementation and module 
self-testing of the OPRM system provide assurance that no new 
equipment malfunctions due to software errors are created. The 
design of the OPRM system also ensures that neither operation nor 
malfunction of the OPRM system will adversely impact the operation 
of the other systems and no accident or equipment malfunction of 
these other systems could cause the OPRM system to malfunction or 
cause a different kind of accident. No new failure modes of either 
the new OPRM equipment or of the existing APRM equipment have been 
introduced. Therefore, operation with the OPRM system does not 
create the possibility of a new or different kind of accident from 
any previously evaluated.
    The DSS-CD solution operates within the existing Option III OPRM 
hardware. Implementation of the DSS-CD will require a software/
hardware change to the existing Option III system. No new operating 
mode, safety-related equipment lineup, accident scenario, system 
interaction, or equipment failure mode was identified. Therefore, 
the DSS-CD solution will not adversely affect plant equipment. 
Because there are no significant hardware changes, there is no

[[Page 2738]]

change in the possibility or consequences of a failure. The worst-
case failure of the equipment is a failure to initiate mitigating 
action (i.e., scram), but no failure can cause an accident of a new 
or different kind than any previously evaluated.
    As such the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Does the proposed amendment involve a significant reduction 
in a margin of safety?
    Response: No.
    The calculated loads on all affected structures, systems and 
components have been shown to remain within design allowables for 
all design basis event categories. No NRC acceptance criteria are 
exceeded. The margins of safety currently included in the design of 
the plant are not affected by the MELLLA+ operating range expansion. 
Because the plant configuration and response to transients and 
hypothetical accidents do not result in exceeding the presently 
approved NRC acceptance limits, operation in the MELLLA+ region does 
not involve a significant reduction in a margin of safety.
    The OPRM system monitors small groups of LPRM signals for 
indication of local variations of core power consistent with 
thermal-hydraulic oscillations and generates an RPS trip when 
conditions consistent with the onset of oscillations are detected. 
An unmitigated thermal-hydraulic instability event has the potential 
to result in a challenge to the MCPR safety limit. The OPRM system 
provides the capability to automatically detect and suppress 
conditions which might result in a thermal-hydraulic instability 
event and thereby maintains the margin of safety by providing 
automatic protection for the MCPR safety limit while reducing the 
burden on the control room operators significantly. The BSP, in lieu 
of the ICAs, will provide adequate protection should the OPRM 
equipment become temporarily inoperable. Operation with the OPRM 
system does not involve a significant reduction in a margin of 
safety.
    The DSS-CD solution is designed to identify the power 
oscillations upon inception and initiate control rod insertion to 
terminate (i.e., scram) the oscillations prior to any significant 
amplitude growth. The DSS-CD solution algorithm will maintain or 
increase the margin to the SLMCPR for anticipated instability 
events. The safety analyses in NEDC-33075P demonstrate the margin to 
the SLMCPR for postulated bounding stability events. In addition, 
the current Option III algorithms are retained to provide defense-
in-depth protection for unanticipated reactor instability events. As 
a result, there is no impact on the MCPR Safety Limit identified for 
an instability event.
    Therefore, operation of CPS in accordance with the proposed 
changes will not involve a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Edward J. Cullen, Deputy General Counsel 
Exelon BSC--Legal, 2301 Market Street, Philadelphia, PA 19101.
    NRC Section Chief: Anthony J. Mendiola.

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

    Date of amendment request: December 23, 2003.
    Description of amendment request: The licensee proposed to revise 
Section 3.4.A and 3.5.A.2 of the Technical Specifications to clarify 
requirements for inoperable components and allow meeting the water 
availability requirements during periods of core spray system 
inoperability (e.g., when the plant is shutdown) in an alternate 
manner. Specifically, this would allow the required water volume for 
core spray system operability be located in the torus, condensate 
storage tank, or a combination of both, in order to provide operational 
flexibility in water management and outage work scheduling. 
Additionally, the licensee proposed to improve consistency of 
verification requirements within the specifications and provide more 
definitive bases for the specifications. No physical changes to the 
plant are involved, and the requirements in the current specifications 
will be maintained.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration. The NRC staff has reviewed the licensee's analysis 
against the three standards of 10 CFR 50.92(c). The NRC staff's 
analysis is presented below:
    The first standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
increase in the probability or consequences of an accident previously 
evaluated. The proposed changes will be made in a manner such that the 
current requirements are maintained for the core spray system. The 
source of core spray water was not considered as a precursor of any 
previously analyzed and evaluated accident. No hardware design change 
is involved with the proposed amendment. Thus, the proposed amendment 
would create no adverse effect on the functional performance of any 
plant structure, system, or component (SSC). All SSCs will continue to 
perform their design functions with no decrease in their capabilities 
to mitigate the previously analyzed consequences of postulated 
accidents. Accordingly, the revised specifications will lead to no 
increase in the consequences of an accident previously evaluated, and 
no increase of the probability of an accident previously evaluated.
    The second standard requires that operation of the unit in 
accordance with the proposed amendment will not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated. The proposed amendment is not the result of a hardware 
design change, nor does it lead to the need for a hardware design 
change. There is no change in the methods the unit is operated. As a 
result, all SSCs will continue to perform as previously analyzed by the 
licensee, and previously evaluated and accepted by the NRC staff. 
Therefore, the proposed amendment will not create the possibility of a 
new or different kind of accident from any previously evaluated.
    The third standard requires that operation of the unit in 
accordance with the proposed amendment will not involve a significant 
reduction in a margin of safety. Since the licensee did not propose to 
exceed or alter a design basis or safety limit, and did not propose to 
operate any component in a less conservative manner, the proposed 
amendment will not affect in any way the performance characteristics 
and intended functions of any SSC. Therefore, the proposed amendment 
does not involve a significant reduction in a margin of safety.
    Based on the NRC staff's analysis, it appears that the three 
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff 
proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: Kevin P. Gallen, Morgan, Lewis & Bockius, 
LLP, 1800 M Street, NW., Washington, DC 20036-5869.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: December 23, 2003.
    Brief description of amendments: The licensee proposed to revise 
various parts of the Technical Specifications (TSs) to allow entry into 
a mode or other specified condition in the applicability of a 
specification while in a condition statement and the associated 
required actions of the TSs, provided the licensee

[[Page 2739]]

performs a risk assessment and manages risk consistent with the program 
in place for complying with the requirements of 10 CFR 50.65(a)(4). 
Specifically, TS 3.0, ``Limiting Conditions for Operation (General),'' 
as well as other portions of the TSs (i.e., Sections 3.4, 3.7, and 3.8) 
referencing TS 3.0, will be revised.
    This change was proposed by the industry's Technical Specification 
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a 
notice of opportunity for comment in the Federal Register on August 2, 
2002 (67 FR 50475), on possible amendments concerning TSTF-359, 
including a model safety evaluation and model no significant hazards 
consideration (NSHC) determination, using the consolidated line item 
improvement process. The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on April 4, 2003 (68 FR 16579). In 
its application for amendment, the licensee affirmed the applicability 
of the following NSHC determination.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee presented 
an analysis of NSHC by endorsing the model NSHC published in 68 FR 
16579 (reproduced below):

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The proposed change allows entry into a mode or other specified 
condition in the applicability of a TS, while in a TS condition 
statement and the associated required actions of the TS. Being in a 
TS condition and the associated required actions is not an initiator 
of any accident previously evaluated. Therefore, the probability of 
an accident previously evaluated is not significantly increased. The 
consequences of an accident while relying on required actions as 
allowed by proposed LCO 3.0.4, are no different than the 
consequences of an accident while entering and relying on the 
required actions while starting in a condition of applicability of 
the TS. Therefore, the consequences of an accident previously 
evaluated are not significantly affected by this change. The 
addition of a requirement to assess and manage the risk introduced 
by this change will further minimize possible concerns. Therefore, 
this change does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.

Criterion 2--The Proposed Change Does Not Create the Possibility of a 
New or Different Kind of Accident From Any Previously Evaluated

    The proposed change does not involve a physical alteration of 
the plant (no new or different type of equipment will be installed). 
Entering into a mode or other specified condition in the 
applicability of a TS, while in a TS condition statement and the 
associated required actions of the TS, will not introduce new 
failure modes or effects and will not, in the absence of other 
unrelated failures, lead to an accident whose consequences exceed 
the consequences of accidents previously evaluated. The addition of 
a requirement to assess and manage the risk introduced by this 
change will further minimize possible concerns. Thus, this change 
does not create the possibility of a new or different kind of 
accident from an accident previously evaluated.

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

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

    The NRC staff proposes to determine that the amendment request 
involves no significant hazards consideration.
    Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston & 
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Richard J. Laufer.

Calvert Cliffs Nuclear Power Plant, Inc., Docket No. 50-317, Calvert 
Cliffs Nuclear Plower Plant, Unit No. 2, Calvert County, Maryland

    Date of amendment request: September 30, 2003.
    Description of amendment request: The proposed amendment would 
increase the maximum enrichment limit of the fuel assemblies that can 
be stored in the Unit 2 spent fuel pool by taking credit for soluble 
boron, burnup and configuration control in maintaining acceptable 
margins of subcriticality.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Would not involve a significant increase in the probability 
or consequences of an accident previously evaluated.
    The proposed change will increase the maximum enrichment limit 
of the fuel assemblies that can be stored in the Unit 2 spent fuel 
pool (SFP) by taking credit for soluble boron, burnup and 
configuration control in maintaining acceptable margins of 
subcriticality. The proposed change will modify Technical 
Specification 4.3.1 ``Criticality,'' add Technical Specification 
3.7.16, ``Spent Fuel Pool Boron Concentration'' and add Technical 
Specification 3.7.17 ``Spent Fuel Pool Storage.'' The postulated 
accidents for the SFP are basically four types; (1) dropped fuel 
assembly on top of the storage rack, (2) a misloading accident, (3) 
an abnormal location of a fuel assembly, and (4) loss-of-normal 
cooling to the SFP.
    There is no increase in the probability of a fuel assembly drop 
accident in the SFP when considering the higher enriched fuel or the 
presence of soluble boron in the SFP water. Dropping a fuel assembly 
on top of the SFP storage racks is not credible at Calvert Cliffs 
due to the design of the spent fuel handling machine and the height 
of the SFP storage racks. The handling of fuel assemblies has always 
been performed in borated water and will not change as a result of 
crediting soluble boron in the SFP criticality analysis. The 
proposed change does not change the general design or 
characteristics of the fuel assemblies. Therefore, the proposed 
change does not increase the probability of a fuel assembly drop 
accident.
    There is no increase in the probability of the accidental 
misloading of irradiated fuel assemblies into the SFP storage racks 
when considering the higher enriched fuel or the presence of soluble 
boron in the SFP water for criticality control. Fuel assembly 
placement will continue to be controlled pursuant to approved fuel 
handling procedures.
    Due to the design of the SFP storage racks, an abnormal 
placement of a fuel assembly into the SFP storage racks is not 
possible. Also, the design of the SFP prevents an inadvertent 
placement of a fuel assembly between the outer most storage cell and 
the pool wall. The proposed change does not make any change to the 
design of SFP. Therefore, there is no increase in the probability of 
abnormal placement of a fuel assembly into the SFP storage racks.
    The proposed change will not result in any changes to the SFP 
cooling system, and the fuel assembly design and characteristics are 
not changed by an increase in fuel enrichment. Therefore, there is 
no increase in the probability of a loss of SFP cooling. Also, since 
a high concentration of soluble boron has always been maintained in 
the SFP water, there is no increase in the probability of the loss 
of normal cooling to the SFP water considering the presence of 
soluble boron in the pool water for criticality control.

[[Page 2740]]

    There is no increase in the consequences of an accidental drop, 
accidental misloading, or abnormal placement of a maximum enriched 
fuel assembly into the SFP storage racks, because the criticality 
analysis demonstrates that the pool will remain subcritical 
following either event. The Technical Specification limit for SFP 
boron concentration will ensure that an adequate SFP boron 
concentration will be maintained.
    There is no increase in the consequences of a loss-of-normal SFP 
cooling because the Technical Specification boron concentration 
provides significant negative reactivity. Loss of the SFP water via 
boiling will not result in a loss of soluble boron, since the 
soluble boron is not volatile. Therefore, loss of SFP cooling 
system, without makeup flow, is not a mechanism for boron dilution. 
Even in the unlikely event that soluble boron in the SFP is 
completely diluted via unborated makeup flow, a pool completely 
filled with maximum enriched unburned assemblies will remain 
subcritical by a design margin that meets the requirements of 10 CFR 
50.68.
    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 will increase the maximum enrichment limit 
of the fuel assemblies that can be stored in the Unit 2 SFP by 
taking credit for soluble boron, burnup and configuration control in 
maintaining acceptable margins of subcriticality. Increasing the 
maximum enrichment limit does not create a new type of criticality 
accident.
    Soluble boron has been maintained in the SFP water and is 
currently required by procedures. Therefore, crediting soluble boron 
in the SFP criticality analysis will have no effect on normal pool 
operation and maintenance. Crediting soluble boron will only result 
in increased sampling to verify the boron concentration in 
accordance with the proposed Technical Specification Surveillance 
Requirement. This increased sampling will not create the possibility 
of a new or different kind of accident.
    A dilution of the SFP soluble boron has always been a 
possibility. However, the boron dilution event previously had no 
consequences, since boron was not previously credited in the 
accident analysis. The initiating events that were considered for 
having the potential to cause dilution of the boron in the SFP to a 
level below that credited in the criticality analyses fall into 
three categories: dilution by flooding, dilution by loss-of-coolant 
induced makeup, and dilution by loss-of-cooling system induced 
makeup. The SFP dilution analysis demonstrates that a dilution event 
that could increase k-effective in the SFP to greater than 0.95 is 
not a credible event. It is not credible that dilution could occur 
for the required length of time without operator notice, since this 
event would activate the high level alarm and initiate Auxiliary 
Building flooding. In addition, in excess of 1,043,000 gallons of 
unborated water must be added to the SFP to reach the minimum 
soluble boron concentration. This is more water volume than is 
contained in both pretreated water storage tanks and also more water 
volume than is contained in the demineralized water storage tank and 
both condensate storage tanks combined. Even in the unlikely event 
that soluble boron in the SFP is completely diluted, the SFP will 
remain subcritical by a design margin that meets the requirements of 
10 CFR 50.68.
    Burned assemblies have been stored in the SFP for many cycles. 
Therefore, crediting burnup in the SFP criticality analysis will 
have no effect on normal pool operation and maintenance. Fuel 
assembly placement, although more complex, will continue to be 
controlled pursuant to approved fuel handling procedures and in 
accordance with Technical Specification spent fuel rack storage 
configuration limitations.
    The proposed change will not result in any other change in the 
plant configuration or equipment design. Therefore, the proposed 
change does not create the possibility of a new or different kind of 
accident from any previously evaluated.
    3. The proposed change does not involve a significant reduction 
in a margin of safety.
    The Technical Specification changes proposed by this license 
amendment request will provide an adequate safety margin to ensure 
that the stored fuel assembly array of maximum enriched fuel will 
always remain subcritical. Those limits are based on a plant 
specific criticality analysis performed for the Calvert Cliffs Unit 
2 SFP, that include technically supported margins.
    Soluble boron is used to provide subcritical margin such that 
the SFP k-effective is maintained less than or equal to 0.95. Since 
k-effective is less than or equal to 0.95, the current margin of 
safety is maintained. In addition, while the criticality analysis 
utilized credit for soluble boron, the fuel in the SFP rack will 
remain subcritical with no soluble boron with a 95 percent 
probability at a 95 percent confidence level as required by 10 CFR 
50.68. This substantial reduction in the SFP soluble boron 
concentration was evaluated and shown not to be credible.
    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 
amendments request involves no significant hazards consideration.
    Attorney for licensee: James M. Petro, Jr., Esquire, Counsel, 
Constellation Energy Group, Inc., 750 East Pratt Street, 5th floor, 
Baltimore, MD 21202.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: October 16, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specification (TS) 3.4.9 to change the minium 
pressurizer (PZR) heater capacity from 126 to 400 kW to correct a non-
conservative TS associated with a PZR design basis deficiency.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated:
    No. The proposed changes revise the minimum PZR [pressurizer] 
heater capacity required and capable of being powered from an 
emergency power supply source. UFSAR [Updated Final Safety Analysis 
Report] do not take credit for PZR heater operation; however, an 
implicit initial condition assumption of the safety analyses is that 
RCS [Reactor Coolant System] is operating at normal pressure. 
Assurance of this assumption is enhanced due to these proposed 
changes. Consequently, the proposed changes do not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any kind of accident previously evaluated:
    No. These changes correct a non-conservative value from the TS 
[technical specification] and are necessary to assure RCS pressure 
control and adequate natural circulation cooling. The available 
heater capacity being powered from an emergency power supply is 
approximately 1000 kW for the most restrictive unit which exceeds 
the proposed 400 kW minimum capacity required by TS. The proposed 
changes help ensure that the RCS is operating at normal pressure 
which is an implicit initial assumption used in several UFSAR 
described safety analyses. Consequently, these changes do not create 
the possibility of a new or different kind of accident from any kind 
of accident previously evaluated.
    3. Involve a significant reduction in a margin of safety:
    No. The proposed change does not adversely affect any plant 
safety limits, set points, or design parameters. The change also 
does not adversely affect the fuel, fuel cladding, RCS, or 
containment integrity. Therefore, the proposed changes do not 
involve a reduction in a margin of safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are

[[Page 2741]]

satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Anne W. Cottington, Winston and Strawn, 1200 
17th Street, NW., Washington, DC 20005.
    NRC Section Chief: John A. Nakoski.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of amendment request: December 5, 2003.
    Description of amendment request: The proposed amendment would 
revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) values in 
Technical Specification 1.1.A.1 to incorporate the results of the 
cycle-specific core reload analysis for Vermont Yankee Nuclear Power 
Station Cycle 24 operation.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration which is presented below:

    1. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant increase in the probability or consequences of an 
accident previously evaluated.
    The basis of the Safety Limit Minimum Critical Power Ratio 
(SLMCPR) is to ensure no mechanis0tic fuel damage is calculated to 
occur if the limit is not violated. The new SLMCPR values preserve 
the existing margin to transition boiling and probability of fuel 
damage is not increased. The derivation of the revised SLMCPR for 
Vermont Yankee for incorporation into the Technical Specifications, 
and its use to determine plant and cycle-specific thermal limits, 
have been performed using NRC [U.S. Nuclear Regulatory Commission] 
approved methods. These plant-specific calculations are performed 
each operating cycle and if necessary, will require future changes 
to these values based upon revised core designs. The revised SLMCPR 
values do not change the method of operating the plant and have no 
effect on the probability of an accident initiating event or 
transient.
    Based on the above, Vermont Yankee has concluded that the 
proposed change will not result in a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not create the 
possibility of a new or different kind of accident from any accident 
previously evaluated.
    The proposed changes result only from a specific analysis for 
the Vermont Yankee core reload design. These changes do not involve 
any new or different methods for operating the facility. No new 
initiating events or transients result from these changes.
    Based on the above, Vermont Yankee has concluded that the 
proposed change will not create the possibility of a new or 
different kind of accident from those previously evaluated.
    3. The operation of Vermont Yankee Nuclear Power Station in 
accordance with the proposed amendment, will not involve a 
significant reduction in a margin of safety.
    The new SLMCPR is calculated using NRC approved methods with 
plant and cycle specific parameters for the current core design. The 
SLMCPR value remains high enough to ensure that greater than 99.9% 
of all fuel rods in the core will avoid transition boiling if the 
limit is not violated, thereby preserving the fuel cladding 
integrity. The operating MCPR limit is set appropriately above the 
safety limit value to ensure adequate margin when the cycle specific 
transients are evaluated. Accordingly, the margin of safety is 
maintained with the revised values.
    As a result, Vermont Yankee has determined that the proposed 
change will not result in a significant reduction in a margin of 
safety.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and 
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
    NRC Section Chief: Darrell J. Roberts, Acting.

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

    Date of amendment request: December 19, 2003.
    Description of Amendment Request: The proposed amendment deletes 
requirements from the Technical Specifications (TS) to maintain 
hydrogen recombiners and hydrogen monitors.
    The NRC staff issued a notice of opportunity for comment in the 
Federal Register on August 2, 2002 (67 FR 50374), on possible 
amendments to eliminate the hydrogen recombiners from TS, including a 
model safety evaluation and model no significant hazards consideration 
(NSHC) determination, using the Consolidated Line Item Improvement 
Process (CLIIP). The NRC staff subsequently issued a notice of 
availability of the models for referencing in license amendment 
applications in the Federal Register on September 25, 2003 (68 FR 
55416). The licensee affirmed the applicability of the model NSHC 
determination in its application dated December 19, 2003.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), an analysis of the issue 
of no significant hazards consideration is presented below:

Criterion 1--The Proposed Change Does Not Involve a Significant 
Increase in the Probability or Consequences of an Accident Previously 
Evaluated

    The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates 
requirements for hydrogen control systems to mitigate such a 
release. The installation of hydrogen recombiners and/or vent and 
purge systems required by 10 CFR 50.44(b)(3) was intended to address 
the limited quantity and rate of hydrogen generation that was 
postulated from a design-basis LOCA. The Commission has found that 
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage. In addition, these systems were 
ineffective at mitigating hydrogen releases from risk-significant 
accident sequences that could threaten containment integrity.
    With the elimination of the design-basis LOCA hydrogen release, 
hydrogen monitors are no longer required to mitigate design-basis 
accidents and, therefore, the hydrogen monitors do not meet the 
definition of a safety-related component as defined in 10 CFR 50.2. 
RG [Regulatory Guide] 1.97 Category 1, is intended for key variables 
that most directly indicate the accomplishment of a safety function 
for design-basis accident events. The hydrogen monitors no longer 
meet the definition of Category 1 in RG 1.97. As part of the 
rulemaking to revise 10 CFR 50.44, the Commission found that 
Category 3, as defined in RG 1.97, is an appropriate categorization 
for the hydrogen monitors because the monitors are required to 
diagnose the course of beyond design-basis accidents.
    The regulatory requirements for the hydrogen monitors can be 
relaxed 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. 
Classification of the hydrogen monitors as Category 3, and removal 
of the hydrogen monitors from TS will not prevent an accident 
management strategy 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 the hydrogen recombiner 
requirements and relaxation of

[[Page 2742]]

the hydrogen monitor requirements, including removal of these 
requirements from TS, does not involve a significant increase in the 
probability or 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 the hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, will not result in any failure mode 
not previously analyzed. The hydrogen recombiner and hydrogen 
monitor equipment was intended to mitigate a design-basis hydrogen 
release. The hydrogen recombiner and hydrogen monitor equipment are 
not considered accident precursors, nor does their existence or 
elimination have any adverse impact on the pre-accident state of the 
reactor core or post accident confinement of radionuclides 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 hydrogen recombiner requirements and 
relaxation of the hydrogen monitor requirements, including removal 
of these requirements from TS, 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.
    The installation of hydrogen recombiners and/or vent and purge 
systems required by 10 CFR 50.44(b)(3) was intended to address the 
limited quantity and rate of hydrogen generation that was postulated 
from a design-basis LOCA. The Commission has found that this 
hydrogen release is not risk-significant because the design-basis 
LOCA hydrogen release does not contribute to the conditional 
probability of a large release up to approximately 24 hours after 
the onset of core damage.
    Category 3--hydrogen monitors are adequate 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 intent of the 
requirements established as a result of the TMI, Unit 2 accident can 
be adequately met without reliance on safety-related hydrogen 
monitors.
    Therefore, this change does not involve a significant reduction 
in the margin of safety. Removal of hydrogen monitoring from TS will 
not result in a significant reduction in their functionality, 
reliability, and availability.

    Based upon the reasoning presented above, the requested change does 
not involve a significant hazards consideration. Therefore, the NRC 
staff proposes to determine that the amendment request involves no 
significant hazards consideration.
    Attorney for licensee: N. S. Reynolds, Esquire, Winston & Strawn, 
1400 L Street, NW., Washington, DC 20005-3502.
    NRC Section Chief: Robert A. Gramm.

Exelon Generation Company, LLC, Docket No. STN 50-454, Byron Station, 
Unit 1, Ogle County, Illinois

    Date of amendment request: December 5, 2003.
    Description of amendment request: The proposed amendment would 
allow irradiation of two lead test assemblies (LTAs) and two 
``standard'' Westinghouse 17x17 VANTAGE+ZIRLOTM assemblies 
beyond the current fuel rod-average licensing basis burnup value of 
60,000 MWD/MTU up to 65,000 MWD/MTU during the current operating cycle 
(B1C13). Irradiation of these four assemblies is intended to confirm 
the acceptable use of the ZIRLOTM alloys to a discharge 
burnup level exceeding the current licensing basis.
    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.
    Fuel rod defects or failures are not considered as initiators 
for any previously analyzed accident; therefore the requested 
license amendment will have no effect on the probability of any 
previously evaluated accident. In addition, NRC-approved 
methodologies and technical reports have been used in the B1C13 
specific reload safety evaluation to confirm that the fuel rod 
design limits will be met; therefore, increasing the burnup limit of 
the specified fuel assemblies to the requested value will not 
increase the consequences of any previously analyzed accident.
    The regular ZIRLOTM and ZIRLOTM (LT-1) 
high burnup fuel rods will continue to satisfy the specified 
acceptable fuel design limits (SAFDLs) specified in NRC-approved 
Westinghouse topical reports. The clad integrity of the 
ZIRLOTM and ZIRLOTM (LT-1) high burnup rods 
will be maintained as the subject fuel assemblies will be placed in 
less than limiting core locations and will continue to meet the 
safety parameter requirements. The acceptability of using the 
ZIRLOTM and ZIRLOTM (LT-1) high burnup rods 
has been evaluated and confirmed in the B1C13 Reload Safety 
Evaluation supported by the Westinghouse LTA Report, ``Byron Unit 1 
Cycle 13 LTA Report,'' dated August 2003.
    It has been shown in WCAP-12610-P-A, that even though there are 
variations in core inventories of isotopes due to extended burnup up 
to 75,000 MWD/MTU, there are no significant increases of isotopes 
that are major contributors to accident doses. It is worthy to note 
that, at higher burnups, there is actually a reduction in certain 
isotopes that are major dose contributors under accident situations 
(e.g., Kr-88). With only a limited number of ZIRLOTM and 
ZIRLOTM (LT-1) high burnup rods in the entire core, any 
variation of isotopes will be extremely small. Thus, the radiation 
dose limitations of 10 CFR 100, ``Reactor Site Criteria,'' will not 
be exceeded.
    Based on the above discussion, it is concluded that 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 to increase the current fuel rod-average 
burnup limit does not involve the use or installation of new 
equipment and all currently installed equipment will not be operated 
in a new or different manner. No new or different system 
interactions are created and no new processes are introduced. The 
proposed change will not introduce any new failure mechanisms, 
malfunctions, or accident initiators not already considered in the 
design and licensing bases.
    Based on this evaluation, 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 change to increase the current fuel rod-average 
burnup limit of 60,000 MWD/MTU up to 65,000 MWD/MTU during B1C13 
will cause the following fuel rod design criteria to become more 
limiting: Fuel rod growth, clad fatigue, rod internal pressure and 
cladding corrosion. However, the regular ZIRLOTM and 
ZIRLOTM (LT-1) high burnup fuel rods will continue to 
satisfy the SAFDLs specified in NRC-approved Westinghouse topical 
reports as noted above. The clad integrity of the ZIRLOTM 
and ZIRLOTM (LT-1) high burnup rods and the appropriate 
margin to safety will be maintained as the subject fuel assemblies 
will be placed in less than limiting core locations and will 
continue to meet the safety parameter requirements. The 
acceptability of using the ZIRLOTM and ZIRLOTM 
(LT-1) high burnup rods has been evaluated and confirmed in the 
B1C13 Reload Safety Evaluation supported by the Westinghouse LTA 
Report, ``Byron Unit 1 Cycle 13 LTA Report,'' dated August 2003.
    Based on the above evaluation, 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 
requested amendments involve no significant hazards consideration.

[[Page 2743]]

    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 No. 50-265, Quad Cities Nuclear 
Power Station, Unit 2, Rock Island County, Illinois Date of amendment 
request:

    Date of amendment request: November 14, 2003, as supplemented by 
letter dated December 23, 2003.
    Description of amendment request: The proposed amendment would 
revise the values and wording of the technical specifications safety 
limit minimum critical power ratio (SLMCPR).
    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?
    The probability of an evaluated accident is derived from the 
probabilities of the individual precursors to that accident. The 
consequences of an evaluated accident are determined by the 
operability of plant systems designed to mitigate those 
consequences. Limits have been established consistent with NRC 
approved methods to ensure that fuel performance during normal, 
transient, and accident conditions is acceptable. The proposed 
change conservatively establishes the SLMCPR for Quad Cities Nuclear 
Power Station (QCNPS), Unit 2, Cycle 18 such that the fuel is 
protected during normal operation and during any plant transients or 
anticipated operational occurrences (AOOs).
    Changing the SLMCPR does not increase the probability of an 
evaluated accident. The change does not require any physical plant 
modifications, physically affect any plant components, or entail 
changes in plant operation. Therefore, no individual precursors of 
an accident are affected.
    The proposed change revises the SLMCPR to protect the fuel 
during normal operation as well as during any transients or 
anticipated operational occurrences. Operational limits will be 
established based on the proposed SLMCPR to ensure that the SLMCPR 
is not violated during all modes of operation. This will ensure that 
the fuel design safety criterion (i.e., that at least 99.9% of the 
fuel rods do not experience transition boiling during normal 
operation and anticipated operational occurrences) is met. Since the 
proposed change does not affect operability of plant systems 
designed to mitigate any consequences of accidents, the consequences 
of an accident previously evaluated are not expected to increase.
    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?
    Creation of the possibility of a new or different kind of 
accident would require creating one or more new precursors of that 
accident. New accident precursors may be created by modifications of 
plant configuration, including changes in allowable modes of 
operation. The proposed change does not involve any plant 
configuration modifications or changes to allowable modes of 
operation. The proposed change to the SLMCPR assures that safety 
criteria are maintained for QCNPS, Unit 2, Cycle 18.
    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?
    The SLMCPR provides a margin of safety by ensuring that at least 
99.9% of the fuel rods do not experience transition boiling during 
normal operation and AOOs if the MCPR limit is not violated. The 
proposed change will ensure the appropriate level of fuel 
protection. Additionally, operational limits will be established 
based on the proposed SLMCPR to ensure that the SLMCPR is not 
violated during all modes of operation. This will ensure that the 
fuel design safety criteria (i.e., that no more than 0.1% of the 
rods are expected to be in boiling transition if the MCPR limit is 
not violated) are met.
    Therefore, the proposed change does not involve a significant 
reduction in a margin of safety.

    The supplemental letter contained clarifying information and did 
not change the initial no significant hazards consideration 
determination and did not expand the scope of the original Federal 
Register notice.
    The 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.

Florida Power and Light Company, Docket No. 50-389, St. Lucie Plant, 
Unit No. 2, St. Lucie County, Florida

    Date of amendment request: December 2, 2003.
    Description of amendment request: The proposed amendment would 
revise the Technical Specifications to allow a reduction in the minimum 
reactor coolant system flow, corresponding to an increase in the steam 
generator tube plugging limit from 15 percent to 30 percent.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    (1) Operation of the facility in accordance with the proposed 
amendment would 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 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: M.S. Ross, Attorney, Florida Power & Light, 
PO Box 14000, Juno Beach, Florida 33408-0420.
    NRC Section Chief: Allen G. Howe.

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

    Date of amendment request: October 8, 2003.
    Description of amendment request: The proposed amendment is to 
revise Technical Specifications (TS) 4.2.b.3.a, ``Inspection 
Frequency,'' for the Kewaunee Nuclear Power Plant (KNPP). The proposed 
one-time change would revise the steam generator (SG) inspection 
interval requirements in TS for KNPP to allow a 40-month inspection 
interval after one SG inspection.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Involve a significant increase in the probability or 
consequences of an accident previously evaluated?
    The proposed one-time change revises the Steam Generator (SG) 
inspection interval requirements in Technical Specifications (TS) 
4.2.b.3.a, following the Kewaunee Nuclear Plant, spring 2003 
refueling outage, to allow a 40-month inspection frequency after one 
inspection, rather than after two consecutive inspections results 
that are within the C-1 category.
    The proposed on-time extension of the SG tube in-service 
inspection interval does not involve changing any structure, system, 
or component, or affect reactor operations. It is not an initiator 
of an accident and does not

[[Page 2744]]

change any existing safety analysis previously analyzed in the 
Kewaunee Updated Safety Analysis Report (USAR). As such, the 
proposed changes do not involve a significant increase in the 
probability of an accident previously evaluated.
    Since the proposed change does not alter the plant design, there 
is no direct increase in SG leakage. Industry experience indicates 
that the probability of increased SG tube degradation would be very 
low. Additionally, steps described below will further minimize the 
risk associated with this extension. For example, the scope of 
inspections performed during the last KNPP refueling outage (i.e., 
the first refueling outage following Steam generator replacement 
(SGR) exceeded the TS requirements for the first two refuleing 
outages after SGR. That is, more tubes were inspected than were 
required by TS (i.e., 100 percent inspection was performed). 
Currently, KNPP does not have an active SG damage mechanism, and 
will meet the current industry examination guidelines without 
performing additional SG inspections until the spring 2006 refueling 
outage. Additionally, as part of our SG Tube Surveillance Program, 
both a Condition Monitoring Assessment and an Operational Assessment 
are performed after each inspection and compared to the Nuclear 
Energy Institute (NEI) 97-06, ``Steam Generator Program 
Guidelines,'' performance criteria. The results of the Condition 
Monitoring Assessment demonstrated that all performance criteria 
were met during the KNPP spring 2003 refueling outage, and the 
results of the Operational Assessment show that all performance 
criteria will be met over the proposed operating period. Considering 
these actions, along with improved SG design and reliability of 
Westinghouse replacement SGs, extending the SG tube inspection 
frequency does not involve a significant increase in the 
consequences of an accident previously evaluated.
    Therefore, the proposed change does not involve a significant 
increase in the probability or consequences of an accident 
previously evaluated.
    2. Create the possibility of a new or different kind of accident 
from any accident previously evaluated?
    The proposed change revises the SG inspection frequency 
requirements in TS 4.2.b.3.a, to allow a 40-month inspection 
interval after one inspection, rather than after two consecutive 
inspections with inspection results within the C-1 category.
    The proposed change will not alter any plant design basis or 
postulated accident resulting from potential SG tube degradation. 
The scope of inspections (i.e., 100 percent) performed during the 
last KNPP refueling outage (i.e., the first refueling outage 
following SG replacement) significantly exceeded the TS requirements 
for the scope of the first two refueling outages after SG 
replacement.
    Primary to secondary leakage that may be experienced during all 
plant conditions is expected to remain within current accident 
analysis assumptions. The proposed change does not affect the design 
of the SGs, the method of SG operation, or reactor coolant chemistry 
controls. No new equipment is being introduced, and installed 
equipment is not being operated in a new or different manner. The 
proposed change involves a one-time extension to the SG tube in-
service inspection frequency, and therefore will not give rise to 
new failure modes. In addition, the proposed change does not impact 
any other plant system or components.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.
    3. Involve a significant reduction in a margin of safety?
    The SG tubes are an integral part of the Reactor coolant System 
(RCS) pressure boundary that are relied upon to maintain the RCS 
pressure and inventory. The SG tubes isolate the radioactive fission 
products in the reactor coolant from the secondary system. The 
safety function of the SG is maintained by ensuring integrity of the 
SG tubes. In addition, the SG tubes comprise the heat transfer 
surface between the primary and secondary systems such that residual 
heat can be removed from the primary system.
    SG tube integrity is a function of the design, environment, and 
current physical condition. Extending the SG tube inservice 
inspection frequency by one operating cycle will not alter the 
function or design of the SG. SG inspections conducted during the 
first refueling outage following SG replacement demonstrated that 
the SGs do not have an active damage mechanism, and the scope of 
those inspections significantly exceeded those required by the TS. 
These inspection results were comparable to similar inspection 
results for similar replacement SGs installed at other plants, and 
subsequent inspections at those plants yielded results that support 
this extension request. The improved design of the replacement SGs 
also provides reasonable assurance that significant tube degradation 
is not likely to occur over the proposed operating period.
    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. 50-282, Prairie Island 
Nuclear Generating Plant, Unit 1, Goodhue County, Minnesota

    Date of amendment request: August 27, 2003, as supplemented 
December 16, 2003.
    Description of amendment request: The proposed amendment would 
revise Technical Specification 5.5.14, ``Containment Leakage Rate 
Testing Program,'' to allow Unit 1 to be excepted from the requirements 
of Regulatory Guide 1.163, for post-modification integrated leakage 
rate testing associated with steam generator replacement.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    Response: No.
    The proposed change would provide the Prairie Island Nuclear 
Generating Plant an exception from performing a required containment 
integrated leak rate test following the replacement of the steam 
generators in Unit 1.
    Integrated leak rate tests are performed to assure the leak-
tightness of the primary containment boundary system, and as such 
they are not accident initiators. Therefore, not performing an 
integrated leak rate test will not affect the probability of an 
accident previously evaluated.
    The intent of post-modification integrated leak rate testing 
requirements is to assure the leak-tight integrity of the area 
affected by the modification. For the Unit 1 steam generator 
replacement modification, this intent will be satisfied by 
performing the American Society of Mechanical Engineers code 
required inspections and tests. Since the leak-tightness integrity 
of the primary containment boundary affected by replacement of the 
steam generators will be assured, there is no change in the primary 
containment boundary's ability to confine radioactive materials 
during an accident.
    Therefore adding a Technical Specification requirement that 
provides an exception for Unit 1 from the steam generator 
replacement post-modification integrated leak rate testing 
requirements does not involve a significant increase in the 
probability or consequences of an accident previously evaluated.
    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    Response: No.
    The proposed change would provide the Prairie Island Nuclear 
Generating Plant an exception from performing a required containment 
integrated leak rate test following the replacement of the steam 
generators in Unit 1.
    Providing an exception from performing a test does not involve a 
physical change to the plant nor does it change the operation of the 
plant. Thus it cannot introduce a new failure mode.
    Therefore adding a Technical Specification requirement that 
provides an exception for Unit 1 from the steam generator 
replacement post-modification integrated leak rate testing 
requirements does not create the possibility of a new or different 
kind of accident from any previously evaluated.

[[Page 2745]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    The proposed change would provide the Prairie Island Nuclear 
Generating Plant an exception from performing a required containment 
integrated leak rate test following the replacement of the steam 
generators in Unit 1.
    The intent of post-modification integrated leak rate testing 
requirements is to assure the leak-tight integrity of the area 
affected by the modification. This intent will be satisfied by 
performing American Society of Mechanical Engineers code required 
inspections and tests. The acceptance criterion for American Society 
of Mechanical Engineers code system pressure testing for the base 
metal and welds is no leakage. In addition, the test pressure for 
the system pressure test will be several times that required during 
an integrated leak rate test. Since the leak-tight integrity of the 
primary containment boundary affected by replacement of the steam 
generators will be assured, there is no change in the primary 
containment boundary's ability to confine radioactive materials 
during an accident.
    Therefore, adding a Technical Specification requirement that 
provides an exception for Unit 1 from the steam generator 
replacement post-modification integrated leak rate testing 
requirements 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 requests involve 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: L. Raghavan.

PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam 
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania

    Date of amendment request: December 22, 2003.
    Description of amendment request: The proposed amendment would 
revise the Unit 1 and 2 Technical Specifications (TSs) by adding TS 
3.3.1.3, ``Oscillation Power Range Monitor (OPRM) Instrumentation,'' 
and revising TS 3.4.1, ``Recirculation Loops Operating,'' and TS 5.6.5, 
``Core Operating Limits Report,'' to remove specifications and 
information related to current stability specifications which will no 
longer be needed.
    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 of occurrence or consequences of an accident 
previously evaluated?
    Response: No.
    The OPRM most directly affects the APRM [average power range 
monitor] and LPRM [local power range monitor] portions of the Power 
Range Neutron Monitoring system. Its installation does not affect 
the operation of these sub-systems. None of the accidents or 
equipment malfunctions affected by these sub-systems are affected by 
the presence or operation of the OPRM. The APRM channels provide the 
primary indication of neutron flux within the core and respond 
almost instantaneously to neutron flux changes. The APRM Fixed 
Neutron Flux-High function is capable of generating a trip signal to 
prevent fuel damage or excessive reactor pressure. For the ASME 
[American Society of Mechanical Engineers] overpressurization 
protection analysis in FSAR [Final Safety Analysis Report] Chapter 
5, the APRM Fixed Neutron Flux-High function is assumed to terminate 
the main steam isolation valve closure event. The high flux trip, 
along with the safety/relief valves, limits the peak reactor 
pressure to less than the ASME Code limits. The control rod drop 
accident (CRDA) analysis in Chapter 15 takes credit for the APRM 
Fixed Neutron Flux-High function to terminate the CRDA. The 
Recirculation Flow Controller Failure event (pump runup) is also 
terminated by the high neutron flux trip. The APRM Fixed Neutron 
Flux-High function is required to be OPERABLE in MODE 1 where the 
potential consequences of the analyzed transients could result in 
the Safety Limits (e.g., MCPR [minimum critical power ratio] and 
Reactor pressure) being exceeded.
    The installation of the OPRM equipment does not increase the 
consequences of a malfunction of equipment important to safety. The 
APRM and RPS [Reactor Protection System] systems are designed to 
fail in a tripped (fail safe) condition; the OPRM will have no 
affect on the consequences of the failure of either system. An 
inoperative trip signal is received by the RPS any time an APRM mode 
switch is moved to any position other than Operate, an APRM module 
is unplugged, the electronic operating voltage is low, or the APRM 
has too few LPRM inputs. These functions are not specifically 
credited in the accident analysis, but are retained for the RPS as 
required by the NRC approved licensing basis.
    The OPRM allows operation under operating conditions presently 
restricted by the current Technical Specifications by providing 
automatic suppression functions in the area of concern in the event 
an instability occurs. The consequences of any accident or equipment 
malfunction are not increased by operating under those conditions. 
Although protected by the OPRM from thermal-hydraulic core 
instabilities above 30% core power, operation under natural core 
circulation conditions is not allowed. No accidents or transients of 
a type not analyzed in the FSAR are created by operating under these 
conditions with the protection of the OPRM system.
    This change does not increase the probability of an accident as 
previously evaluated. The OPRM is designed and installed to not 
degrade the existing APRM, LRPM, and RPS systems. These systems will 
still perform all of their intended functions. The new equipment is 
tested and installed to the same or more restrictive environmental 
and seismic envelopes as the existing systems. The new equipment has 
been designed and tested to electromagnetic interference (EMI) 
requirements which assure correct operation of the existing 
equipment. The new system has been designed to single failure 
criteria and is electrically isolated from equipment of different 
electrical divisions and from non-1E equipment. The electrical 
loading is within the capability of the existing power sources and 
the heat loads are within the capability of existing cooling 
systems. The OPRM allows operation under operating conditions 
presently forbidden or restricted by the current Technical 
Specifications. No other transient or accident analysis assumes 
these operating restrictions.
    Therefore, this 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.
    This proposal does not create the possibility of a new or 
different type of accident from any accident previously evaluated. 
The OPRM system is a monitoring and accident mitigation system that 
cannot create the possibility for an accident not previously 
evaluated.
    The OPRM will allow operation in conditions restricted by the 
current Technical Specifications. Although protected by the OPRM 
from thermal-hydraulic core instabilities above 30% core power, 
operation under natural circulation conditions is not allowed. No 
accidents or transients of a type not analyzed in the FSAR are 
created by operating under these conditions with the protection of 
the OPRM system. No new failure modes of either the new OPRM 
equipment or of the existing APRM equipment have been introduced. 
Quality software design, testing, implementation and module self-
health testing provides assurance that no new equipment malfunctions 
due to software errors are created. The possibility of an accident 
of a new or different type than any evaluated previouly is not 
created.
    The new OPRM equipment is designed and installed to the same 
system requirements as the existing APRM equipment and is designed 
and tested to have no impact on the existing functions of the APRM 
system. Appropriate isolation is provided where new interconnections 
between redundant separation groups are formed. The OPRM modules 
have been designed and tested to assure that no new failure modes 
have been introduced.
    Therefore, the proposed change does not create the possibility 
of a new or different kind of accident from any accident previously 
evaluated.

[[Page 2746]]

    3. Does the proposed change involve a significant reduction in a 
margin of safety?
    Response: No.
    There has been no reduction in the margin of safety as defined 
in the basis for the Technical Specifications. The OPRM system does 
not negatively impact the existing APRM system. As a result, the 
margins in the Technical Specifications for the APRM system are not 
impacted by this addition.
    Current operation under the ICAs [interim corrective actions] 
provides an acceptable margin of safety in the event of an 
instability event as the result of preventive actions and Technical 
Specification controlled response by the control room operators. The 
OPRM system provides an increase in the reliability of the 
protection of the margin of safety by providing automatic protection 
of the MCPR safety limit, while the protection burden is 
significantly reduced for the control room operators. This 
protection is demonstrated as described above, and in the NRC 
reviewed and approved Topical Reports NEDO-32465-A and CENPD-400-P-
A.
    Replacement of the ICA operating restrictions from Technical 
Specifications with the OPRM system does not affect the margin of 
safety associated with any other system or fuel design parameter.
    Therefore, this change does not involve a 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: Bryan A. Snapp, Esquire, Assoc, General 
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3, 
Allentown, PA 18101,1179.
    NRC Section Chief: Richard J. Laufer.

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

    Date of amendment request: November 17, 2003.
    Description of amendment request: The proposed change would revise 
the Technical Specifications to delete the primary containment 
isolation valves and instrumentation associated with the permanent 
removal of the reactor vessel head spray piping.
    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 not involve a significant increase in the probability or 
consequences of an accident previously analyzed?
    Response: No.
    The proposed changes to Technical Specification Tables 3.3.2-1, 
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in 
structures, systems, or components that would affect the probability 
or consequences of any accident previously evaluated in the Hope 
Creek Updated Final Safety Analysis Report.
    The proposed changes involve eliminating piping and valves 
associated with the reactor head spray. The reactor head spray 
system was initially provided to cool down the steam dryer and 
separator during shutdown. The head spray system is not credited for 
the prevention or mitigation of any accident. Therefore, neither the 
offsite or control room radiological consequences are affected. The 
head spray piping removal and addition of a bolted flange on the 
reactor coolant pressure boundary enhances plant safety by 
eliminating a source of pipe whip and potential leakage. In 
addition, the drywell penetration will be capped and welded closed. 
This will maintain primary containment integrity and will be 
periodically tested in conjunction with the containment integrated 
leak rate test.
    Therefore, as discussed above, this modification does not 
involve a significant increase in the probability or consequences 
from any accident previously analyzed.
    2. Does not create the possibility of a new or different kind of 
accident from any accident previously analyzed?
    Response: No.
    The proposed changes to Technical Specification Tables 3.3.2-1, 
3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 do not involve a change in 
structures, systems, or components that would create a new or 
different kind of accident from any accident previously evaluated in 
the Hope Creek Updated Final Safety Analysis Report.
    The proposed change to eliminate the head spray piping and the 
addition of a bolted flange on the reactor coolant pressure boundary 
enhances plant safety by eliminating a source of pipe whip and 
potential leakage. In addition, the drywell penetration will be 
capped and welded closed. This will maintain primary containment 
integrity and will be tested in conjunction with the containment 
integrated leak rate test.
    Therefore, as discussed above, this modification does not create 
the possibility of a new or different kind of accident from any 
accident previously evaluated.
    3. Does not involve a significant reduction in the margin of 
safety?
    Response: No.
    The proposed change to delete the head spray valves from Tables 
3.3.2-1, 3.3.7.4-2, 3.4.3.2-1, and 3.6.3-1 does not reduce any 
margin of safety as defined in the Technical Specifications or 
Bases. The bolted flange that will be installed on the head spray 
penetration will maintain the integrity of the reactor coolant 
pressure boundary. This flange would then be tested as part of the 
reactor pressure vessel hydrostatic test. In addition, the drywell 
penetration will be capped and welded closed. This will maintain 
primary containment integrity and will be tested as part of the 
containment integrated leak rate test.
    Accordingly, based on the above, the proposed change does not 
involve a significant reduction in the margin of safety.

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

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: October 13, 2003.
    Description of amendment request: The proposed amendments would 
revise the Technical Specifications (TS) limiting conditions for 
operation 3.8.4, 3.8.5, and 3.8.6, on direct current sources, operating 
and shutdown, and battery cell parameters. The proposed amendments 
creates TS 5.5.19, for a battery monitoring and maintenance program. 
The bases are revised to be consistent with these changes. The proposed 
amendments are based on Technical Specification Task Force (TSTF) 
Traveler, TSTF-360, Revision 1.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Do the proposed changes involve a significant increase in the 
probability or consequences of an accident previously evaluated?
    No. The proposed changes increase the Completion Time for an 
inoperable battery, relocate preventative maintenance requirements 
to licensee controlled programs, and generally restructure the TS 
[technical specification] requirements for DC [direct current] 
sources. The revised requirements will allow licensed operators to 
focus their attention on battery parameters that are indicative of 
battery operability as opposed to preventative maintenance issues. 
The increased Completion Time for an inoperable battery will allow 
corrective maintenance to be accomplished via a more orderly and 
effective work process. It will also minimize the potential for an 
additional shutdown/restart transient to comply with the TS in order 
to accomplish the required maintenance. The DC sources are not 
initiators to any analyzed accident sequence. Operation in 
accordance with the proposed TS will continue to ensure that the DC 
sources remain capable of performing their safety function and that 
all analyzed accidents will continue to be mitigated as previously 
analyzed.

[[Page 2747]]

    2. Do the proposed changes create the possibility of a new or 
different kind of accident from any previously evaluated?
    No. The proposed changes do not introduce any new equipment, 
create new failure modes for existing equipment, or create any new 
limiting single failures. Plant operation will not be altered, and 
all safety functions previously addressed in accident analyses will 
continue to be performed.
    3. Do the proposed changes involve a significant reduction in a 
margin of safety?
    No. The proposed changes will not adversely affect operation of 
plant equipment--principally the four Class 1E DC sources and the 
equipment supported by them. The changes aimed at restructuring the 
TS requirements for DC sources will have the effect of reducing the 
burden on licensed operators by focusing the TS requirements on 
conditions that impair DC source operability. Requirements related 
to preventive maintenance will be addressed via new Specification 
5.5.19 and the plant maintenance program. Margin to the battery 
operability requirements will continue to be maintained at current 
levels in accordance with IEEE-450. The extended Completion Time for 
an inoperable battery has been shown to have a negligible impact on 
plant risk using the criteria of Regulatory Guides 1.174 and 1.177.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
Nations Bank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: John A. Nakoski.

Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424 
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke 
County, Georgia

    Date of amendment request: December 15, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications surveillance requirement (SR) 3.3.1.2 
for the nuclear instrumentation system power range daily surveillance 
when operating above 15-percent rated thermal power. In addition, the 
format of SR 3.3.1.3 is being revised to be consistent with the format 
of the proposed change to SR 3.3.1.2.
    Basis for proposed no significant hazards consideration 
determination: As required by 10 CFR 50.91(a), the licensee has 
provided its analysis of the issue of no significant hazards 
consideration, which is presented below:

    1. Does the proposed change involve a significant increase in 
the probability or consequences of an accident previously evaluated?
    The proposed change to SR [surveillance requirement] 3.3.1.2 
does not significantly increase the probability or consequences of 
an accident previously evaluated in the FSAR [Final Safety Analysis 
Report]. This modification does not directly initiate an accident. 
The consequences of accidents previously evaluated in the FSAR are 
not adversely affected by this proposed change because the change to 
the NIS [nuclear instrumentation system] Power Range channel 
adjustment requirement ensures the conservative response of the 
channel even at part power levels. The proposed change to SR 3.3.1.3 
is to change the format consistent with the format of the proposed 
change to SR 3.3.1.2.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The proposed change to SR 3.3.1.2 does not create the 
possibility of a new or different kind of accident than any accident 
already evaluated in the FSAR. No new accident scenarios, failure 
mechanisms, or limiting single failures are introduced as a result 
of the proposed change. The proposed Technical Specifications change 
does not challenge the performance or integrity of any safety-
related systems. The proposed change to SR 3.3.1.3 is to change the 
format to be consistent with the format of the proposed change to SR 
3.3.1.2.
    3. Does the proposed change involve a significant reduction in 
the margin of safety?
    The proposed change to SR 3.3.1.2 does not involve a significant 
reduction in a margin of safety. The proposed change does require a 
revision to the criterion for implementation of Power Range channel 
adjustment based on secondary power calorimetric calculation; 
however, the change does not eliminate any RTS [reactor trip system] 
surveillances or alter the frequency of surveillances required by 
the Technical Specifications. The revision to the criterion for 
implementation of the daily surveillance will have a conservative 
effect on the performance of the NIS Power Range channel, 
particularly at part power after normalization at 100% RTP [rated 
thermal power] conditions. The nominal trip setpoints specified by 
the Technical Specifications and the safety analysis limits assumed 
in the transient and accident analysis are unchanged. The margin of 
safety associated with the acceptance criteria for any accident is 
unchanged. The proposed change to SR 3.3.1.3 is to change the format 
to be consistent with the format of the proposed change to SR 
3.3.1.2.

    The NRC staff has reviewed the licensee's analysis and, based on 
this review, it appears that the three standards of 10 CFR 50.92(c) are 
satisfied. Therefore, the NRC staff proposes to determine that the 
amendment request involves no significant hazards consideration.
    Attorney for licensee: Mr. Arthur H. Domby, Troutman Sanders, 
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta, 
Georgia 30308-2216.
    NRC Section Chief: John A. Nakoski.

Southern Nuclear Operating Company, Inc., Georgia Power Company, 
Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, 
City of Dalton, Georgia, Docket Nos. 50-321 and 50-366, Edwin I. Hatch 
Nuclear Plant, Units 1 and 2, Appling County, Georgia

    Date of amendment request: December 1, 2003.
    Description of amendment request: The proposed amendments would 
revise Technical Specifications Section 5.5.12, ``Primary Leakage Rate 
Testing Program,'' to change the peak calculated post accident primary 
containment internal pressure to support a 10 psi increase in the 
nominal Unit 1 and 2 reactor operating pressure.
    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?
    The proposed change to TS [technical specification] section 
5.5.12, ``Primary Containment Leakage Rate Testing Program'', 
involves an increase to the peak post accident primary containment 
pressure. It does not involve physical changes to the primary 
containment structure itself, nor to any of its support systems and 
components, nor does it involve changes to any other systems and 
components designed for the prevention of previously analyzed 
events. Consequently, the proposed amendment does not involve a 
significant increase in the probability of occurrence of a 
previously evaluated event.
    The increase in operating pressure for the Hatch reactors from 
1035 psig to 1045 psig results in an increase to the peak post-
accident primary containment internal pressure. This pressure 
increases from 50.5 to 50.8 psig for Unit 1 and from 46.9 to 47.3 
psig for Unit 2. This is a very small increase with respect to the 
Unit 1 and 2 primary containment design pressure of 56 psig and with 
the maximum code allowable pressure of 62 psig. The primary 
containment thus remains capable of withstanding the post accident 
pressure and thus the consequences of a previously evaluated event 
are not increased.
    2. Does the proposed change create the possibility of a new or 
different kind of accident from any accident previously evaluated?
    The primary containment boundary will not be altered by the 
proposed change to

[[Page 2748]]

Technical Specifications sections 5.5.12, Primary Containment 
Leakage Rate Testing Program. Furthermore, the primary containment 
will function as presently described in the Updated Final Safety 
Analysis Report and will be subject to the same structural and 
functional requirements. The containment will be operated, 
maintained and surveilled as before, with the exception of the 
increased peak post accident pressure, which changes the post 
accident test pressure acceptance criteria. As a result, no new 
modes of operation are introduced by this Technical Specifications 
change and therefore, the possibility of a new or different kind of 
accident from any previously evaluated is not created.
    3. Does the proposed change involve a significant decrease in 
the margin of safety?
    The change in the analyzed peak post accident containment 
pressure will require that the containment be tested to ensure that 
it meets leakage acceptance criteria at the new pressures of 50.8 
psig and 47.3 psig for Units 1 and 2 respectively. Therefore, the 
primary containment's ability to sustain the slightly higher 
pressures will be verified during leak rate testing at the required 
intervals.
    The Unit 1 peak pressure increases from 50.5 to 50.8 psig and 
the Unit 2 pressure increases from 46.9 to 47.3 psig. The primary 
containment design pressure is 56 psig for both units and the 
maximum code allowable pressure is 62 psig. Therefore, the margin to 
the design and maximum code allowable pressures has not been 
significantly affected. As a result, this proposed Technical 
Specifications change does not significantly reduce the margin of 
safety associated with the primary containment function.

    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: Ernest L. Blake, Jr., Esquire, Shaw, 
Pittman, Potts and Trowbridge, 2300 N Street, NW., Washington, DC 
20037.
    NRC Section Chief: John A. Nakoski.

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.

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, as revised by letter dated 
December 5, 2003.
    Brief description of amendment request: By letter dated December 5, 
2003, Entergy submitted a revised application for amendment to 
Technical Specification (TS) 3.3.6.1, ``Primary Containment and Drywell 
Isolation Instrumentation,'' to add a provision to the APPLICABILITY 
function that will eliminate the requirement that the Residual Heat 
Removal System Isolation, Reactor Vessel Water Level-Low, Level 3, be 
OPERABLE under certain conditions during refueling outages. 
Specifically, the proposed change requested in the original application 
dated May 12, 2003, would remove the requirement for this isolation 
function, specified in Table 3.3.6.1-1, when the upper containment 
reactor cavity is at the High Water Level condition specified in TS 
3.5.2, ``Emergency Core Cooling Systems Shutdown.'' The revised 
application adds a new surveillance requirement (SR) 3.3.6.1.9 to 
verify that the water level in the upper containment pool is greater 
than or equal to 22 feet 8 inches above the reactor pressure vessel 
flange every four hours, and adds a footnote to Table 3.3.6.1-1, Item 
5.b, for MODE 5 that states that the function is not required when the 
upper containment reactor cavity and transfer canal gates are removed 
and SR 3.3.6.1.9 is met. The proposed SR and footnote are only 
applicable in MODE 5. The May 12, 2003, application was previously 
noticed in the Federal Register on June 10, 2003 (68 FR 34665).
    Date of publication of individual notice in Federal Register: 
December 15, 2003 (68 FR 69726).
    Expiration date of individual notice: January 14, 2004.

Notice of Issuance of Amendments to Facility Operating Licenses

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application complies 
with the standards and requirements of the Atomic Energy Act of 1954, 
as amended (the Act), and the Commission's rules and regulations. The 
Commission has made appropriate findings as required by the Act and the 
Commission's rules and regulations in 10 CFR Chapter I, which are set 
forth in the license amendment.
    Notice of Consideration of Issuance of Amendment to Facility 
Operating License, Proposed No Significant Hazards Consideration 
Determination, and Opportunity for a Hearing in connection with these 
actions was published in the Federal Register as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
applications for amendment, (2) the amendment, and (3) the Commission's 
related letter, Safety Evaluation and/or Environmental Assessment as 
indicated. All of these items are available for public inspection at 
the Commission's Public Document Room, located at One White Flint 
North, Public File Area 01F21, 11555 Rockville Pike (first floor), 
Rockville, Maryland. Publicly available records will be accessible from 
the Agencywide Documents Access and Management Systems (ADAMS) Public 
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS 
or if there are problems in accessing the documents located in ADAMS, 
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].

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

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

    Date of application for amendments: July 14, 2003, as supplemented 
by letter dated October 1, 2003.

[[Page 2749]]

    Brief description of amendments: The amendments extend from 1 hour 
to 24 hours the completion time for Condition B of Technical 
Specification 3.5.1, which defines requirements for the restoration of 
an emergency core cooling system accumulator when it has been declared 
inoperable for a reason other than boron concentration.
    Date of issuance: December 23, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 30 days from the date of issuance December 23, 2003.
    Amendment Nos.: 211, 205, 218, and 200.
    Renewed Facility Operating License Nos. NPF-35, NPF-52, NPF-9, and 
NPF-17: Amendments revised the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59214).
    The supplement dated October 1, 2003, provided clarifying 
information that did not change the scope of the July 14, 2003, 
application nor the initial proposed no significant hazards 
consideration determination.
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated December 23, 2003.
    No significant hazards consideration comments received: No.

Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations, 
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon, 
Vermont

    Date of application for amendment: January 9, 2003.
    Brief description of amendment: The proposed Technical 
Specification (TS) amendment request changes the definition of a Logic 
System Functional Test, deletes the definition of a Simulated Automatic 
Actuation, clarifies Surveillance Requirement 4.5.G.1.a regarding 
simulated automatic actuation testing, and revises associated TS Bases.
    Date of Issuance: December 23, 2003.
    Effective date: As of the date of issuance, and shall be 
implemented within 60 days.
    Amendment No.: 216.
    Facility Operating License No. DPR-28: Amendment revised the 
Technical Specifications.
    Date of initial notice in Federal Register: February 4, 2003 (68 FR 
5674).
    The Commission's related evaluation of this amendment is contained 
in a Safety Evaluation dated December 23, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: October 24, 2003.
    Brief description of amendment: The amendment revises TS 3.1.8, 
``Scram Discharge Volume (SDV) Vent and Drain Valves,'' for the 
condition of having one or more SDV vent or drain lines with one valve 
inoperable.
    Date of issuance: December 30, 2003.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 161.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: November 25, 2003 (68 
FR 66135).
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated December 30, 2003.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: May 8, 2003, as supplemented by 
letter dated October 24, 2003.
    Brief description of amendment: The amendment changes 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.
    Date of issuance: January 8, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment No: 162.
    Facility Operating License No. NPF-29: The amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: June 10, 2003 (68 FR 
34664).
    The October 24, 2003, supplemental letter provided clarifying 
information that did not change the scope of the original Federal 
Register notice or the original no significant hazards consideration 
determination.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 8, 2004.
    No significant hazards consideration comments received: No.

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

    Date of application for amendment: September 3, 2003.
    Brief description of amendment: The amendment modified Technical 
Specification (TS) requirements for mode change limitations to adopt 
the TS Task Force (TSTF) change TSTF-359, ``Increase Flexibility in 
Mode Restraints.''
    Date of issuance: January 5, 2004.
    Effective date: As of the date of issuance to be implemented within 
60 days.
    Amendment No.: 109.
    Facility Operating License No. NPF-69: Amendment revises the 
Technical Specifications.
    Date of initial notice in Federal Register: September 30, 2003 (68 
FR 56345).
    The staff's related evaluation of the amendment is contained in a 
Safety Evaluation dated January 5, 2004.
    No significant hazards consideration comments received: No.

PSEG Nuclear LLC, Docket No. 50-311, Salem Nuclear Generating Station, 
Unit No. 2, Salem County, New Jersey

    Date of application for amendment: July 1, 2003, as supported by 
letter dated June 16, 2003, and supplemented on November 11, 2003.
    Brief description of amendment: The requested changes revise 
License Condition 2.C.(10) to document changes to the Salem Post-Fire 
Safe Shutdown (SSD) strategy for Fire Areas 2-FA-AB-64B, 2-FA-AB-84B, 
and 2-FA-AB-84C. The licensee requested changes to the SSD as a result 
of recent plant modifications implemented in response to the resolution 
of Electrical Raceway Fire Barrier System issues at Salem.
    Date of issuance: January 7, 2004.
    Effective date: As of the date of issuance, and shall be 
implemented within 30 days.
    Amendment No.: 242.
    Facility Operating License No. DPR-75: This amendment revised the 
Facility Operating License.
    Date of initial notice in Federal Register: July 16, 2003 (68 FR 
42134). The supporting and supplemental

[[Page 2750]]

letters dated June 16, and November 11, 2003, contained clarifying 
information that did not change the NRC staff's proposed finding of no 
significant hazards consideration.
    The Commission's related evaluation of the amendment is contained 
in a Safety Evaluation dated January 7, 2004.
    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 amendment request: May 22, 2003.
    Brief description of amendments: The amendments revise Technical 
Specification (TS) 3/4.3.2, ``Engineering Safety Features Actuation 
System Instrumentation,'' and TS 3/4.9.9, ``Refueling Operations--
Containment Ventilation Isolation System,'' governing radiation 
monitoring instrumentation, to relax restrictions on containment purge 
valve operation.
    Date of issuance: January 5, 2004.
    Effective date: As of the date of issuance and shall be implemented 
within 60 days of issuance.
    Amendment Nos.: 160 and 150.
    Facility Operating License Nos. NPF-76 and NPF-80: The amendments 
revised the Technical Specifications.
    Date of initial notice in Federal Register: October 14, 2003 (68 FR 
59221).
    The Commission's related evaluation of the amendments is contained 
in a Safety Evaluation dated January 5, 2004.
    No significant hazards consideration comments received: No.

Tennessee Valley Authority, Docket Nos. 50-260 and 50-296, Browns Ferry 
Nuclear Plant, Units 2 and 3, Limestone County, Alabama

    Date of application for amendments: April 14, 2003, as supplemented 
by letters dated September 5 and November 7, 2003.
    Brief description of amendments: The amendments revised Technical 
Specification (TS) 3.3.4.1, ``End-Of-Cycle Recirculation Pump Trip 
(EOC[dash]RPT) Instrumentation,'' and TS 3.7.5, ``Main Turbine Bypass 
System,'' to reference additional core limits adjustment factors for 
linear heat generation rate for equipment out-of-service conditions. 
Also, Section b of TS 5.65. ``Core Operating Limits Report (COLR),'' 
was revised to add references to the Framatome Advanced Nuclear Power 
analytical methods what will be used in the upcoming fuel cycles to 
determine core operating limits.
    Date of issuance: December 30, 2003.
    Effective date: Date of issuance, to be implemented within 60 days 
from the completion of Unit 3 Spring 2004 and Unit 2 Spring 2005 
refueling outages.
    Amendment Nos.: 287 & 245.
    Facility Operating License Nos. DPR-52, and DPR-68. Amendments 
revised the TSs.
    Date of initial notice in Federal Register: May 27, 2003 (68 FR 
28858).
    TVA's supplemental letters provided clarifying information that did 
not expand the scope of the original application 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 December 30, 2003.
    No significant hazards consideration comments received: No.

Notice of Issuance of Amendments to Facility Operating Licenses and 
Final Determination of No Significant Hazards Consideration and 
Opportunity for a Hearing (Exigent Public Announcement or Emergency 
Circumstances)

    During the period since publication of the last biweekly notice, 
the Commission has issued the following amendments. The Commission has 
determined for each of these amendments that the application for the 
amendment complies with the standards and requirements of the Atomic 
Energy Act of 1954, as amended (the Act), and the Commission's rules 
and regulations. The Commission has made appropriate findings as 
required by the Act and the Commission's rules and regulations in 10 
CFR Chapter I, which are set forth in the license amendment.
    Because of exigent or emergency circumstances associated with the 
date the amendment was needed, there was not time for the Commission to 
publish, for public comment before issuance, its usual 30-day Notice of 
Consideration of Issuance of Amendment, Proposed No Significant Hazards 
Consideration Determination, and Opportunity for a Hearing.
    For exigent circumstances, the Commission has either issued a 
Federal Register notice providing opportunity for public comment or has 
used local media to provide notice to the public in the area 
surrounding a licensee's facility of the licensee's application and of 
the Commission's proposed determination of no significant hazards 
consideration. The Commission has provided a reasonable opportunity for 
the public to comment, using its best efforts to make available to the 
public means of communication for the public to respond quickly, and in 
the case of telephone comments, the comments have been recorded or 
transcribed as appropriate and the licensee has been informed of the 
public comments.
    In circumstances where failure to act in a timely way would have 
resulted, for example, in derating or shutdown of a nuclear power plant 
or in prevention of either resumption of operation or of increase in 
power output up to the plant's licensed power level, the Commission may 
not have had an opportunity to provide for public comment on its no 
significant hazards consideration determination. In such case, the 
license amendment has been issued without opportunity for comment. If 
there has been some time for public comment but less than 30 days, the 
Commission may provide an opportunity for public comment. If comments 
have been requested, it is so stated. In either event, the State has 
been consulted by telephone whenever possible.
    Under its regulations, the Commission may issue and make an 
amendment immediately effective, notwithstanding the pendency before it 
of a request for a hearing from any person, in advance of the holding 
and completion of any required hearing, where it has determined that no 
significant hazards consideration is involved.
    The Commission has applied the standards of 10 CFR 50.92 and has 
made a final determination that the amendment involves no significant 
hazards consideration. The basis for this determination is contained in 
the documents related to this action. Accordingly, the amendments have 
been issued and made effective as indicated.
    Unless otherwise indicated, the Commission has determined that 
these amendments satisfy the criteria for categorical exclusion in 
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b), 
no environmental impact statement or environmental assessment need be 
prepared for these amendments. If the Commission has prepared an 
environmental assessment under the special circumstances provision in 
10 CFR 51.12(b) and has made a determination based on that assessment, 
it is so indicated.
    For further details with respect to the action see (1) the 
application for amendment, (2) the amendment to Facility Operating 
License, and (3) the Commission's related letter, Safety Evaluation 
and/or Environmental Assessment, as indicated. All of these items are 
available for public inspection at the Commission's Public Document

[[Page 2751]]

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 Assess 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 Public 
Document Room (PDR) Reference staff at 1-800-397-4209, 301-415-4737 or 
by email to [email protected].
    The Commission is also offering an opportunity for a hearing with 
respect to the issuance of the amendment. By February 19, 2004, the 
licensee may file a request for a hearing with respect to issuance of 
the amendment to the subject facility operating license and any person 
whose interest may be affected by this proceeding and who wishes to 
participate as a party in the proceeding must file a written request 
for a hearing and a petition for leave to intervene. Requests for a 
hearing and a petition for leave to intervene shall be filed in 
accordance with the Commission's ``Rules of Practice for Domestic 
Licensing Proceedings'' in 10 CFR part 2. Interested persons should 
consult a current copy of 10 CFR 2.714, which is available at the 
Commission's PDR, located at One White Flint North, Public File Area 
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, and 
electronically on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are problems in accessing the 
document, contact the PDR Reference staff at 1-800-397-4209, 301-415-
4737, or by e-mail to [email protected]. If a request for a hearing or 
petition for leave to intervene is filed by the above date, the 
Commission or an Atomic Safety and Licensing Board, designated by the 
Commission or by the Chairman of the Atomic Safety and Licensing Board 
Panel, will rule on the request and/or petition; and the Secretary or 
the designated Atomic Safety and Licensing Board will issue a notice of 
a hearing or an appropriate order.
    As required by 10 CFR 2.714, a petition for leave to intervene 
shall set forth with particularity the interest of the petitioner in 
the proceeding, and how that interest may be affected by the results of 
the proceeding. The petition should specifically explain the reasons 
why intervention should be permitted with particular reference to the 
following factors: (1) The nature of the petitioner's right under the 
Act to be made a party to the proceeding; (2) the nature and extent of 
the petitioner's property, financial, or other interest in the 
proceeding; and (3) the possible effect of any order which may be 
entered in the proceeding on the petitioner's interest. The petition 
should also identify the specific aspect(s) of the subject matter of 
the proceeding as to which petitioner wishes to intervene. Any person 
who has filed a petition for leave to intervene or who has been 
admitted as a party may amend the petition without requesting leave of 
the Board up to 15 days prior to the first prehearing conference 
scheduled in the proceeding, but such an amended petition must satisfy 
the specificity requirements described above.
    Not later than 15 days prior to the first prehearing conference 
scheduled in the proceeding, a petitioner shall file a supplement to 
the petition to intervene which must include a list of the contentions 
which are sought to be litigated in the matter. Each contention must 
consist of a specific statement of the issue of law or fact to be 
raised or controverted. In addition, the petitioner shall provide a 
brief explanation of the bases of the contention and a concise 
statement of the alleged facts or expert opinion which support the 
contention and on which the petitioner intends to rely in proving the 
contention at the hearing. The petitioner must also provide references 
to those specific sources and documents of which the petitioner is 
aware and on which the petitioner intends to rely to establish those 
facts or expert opinion. Petitioner must provide sufficient information 
to show that a genuine dispute exists with the applicant on a material 
issue of law or fact. Contentions shall be limited to matters within 
the scope of the amendment under consideration. The contention must be 
one which, if proven, would entitle the petitioner to relief. A 
petitioner who fails to file such a supplement which satisfies these 
requirements with respect to at least one contention will not be 
permitted to participate as a party.
    Those permitted to intervene become parties to the proceeding, 
subject to any limitations in the order granting leave to intervene, 
and have the opportunity to participate fully in the conduct of the 
hearing, including the opportunity to present evidence and cross-
examine witnesses. Since the Commission has made a final determination 
that the amendment involves no significant hazards consideration, if a 
hearing is requested, it will not stay the effectiveness of the 
amendment. Any hearing held would take place while the amendment is in 
effect.
    A request for a hearing or a petition for leave to intervene must 
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory 
Commission, Washington, DC 20555-0001, Attention: Rulemakings 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 the 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 petition for 
leave to intervene and request for hearing 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 the 
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).

STP Nuclear Operating Company, Docket No. 50-499, South Texas Project, 
Unit 2, Matagorda County, Texas

    Date of amendment request: December 27, 2003 as supplemented by 
letter dated December 27 and two letters dated December 28, 2003.
    Description of amendment request: The amendments revise Technical 
Specification (TS) 3.8.1, ``AC Sources--Operating,'' to extend the 
allowed outage time for Unit 2 Standby Diesel Generator (SDG) 22 from 
21 days to 113 days as a one-time change for the purpose of making 
repairs to SDG 22.
    Date of issuance: December 30, 2003.
    Effective date: December 30, 2003.
    Amendment No.: 149.
    Facility Operating License No. NPF-80: Amendment revised the 
Technical Specifications.

[[Page 2752]]

    Public comments requested as to final no significant hazards 
consideration (NSHC): No.
    The Commission's related evaluation of the amendment, finding of 
emergency circumstances, state consultation, and final NSHC 
determination are contained in a safety evaluation dated December 30, 
2003.
    Attorney for licensee: A.H. Gutterman, Esquire, Morgan, Lewis & 
Bockius, LLP, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
    NRC Section Chief: Robert A. Gramm.

    Dated at Rockville, Maryland, this 13th day of January 2004.

    For the Nuclear Regulatory Commission.
Eric J. Leeds,
Deputy Director, Division of Licensing Project Management, Office of 
Nuclear Reactor Regulation.
[FR Doc. 04-1104 Filed 1-16-04; 8:45 am]
BILLING CODE 7590-01-P