[Federal Register Volume 69, Number 91 (Tuesday, May 11, 2004)]
[Notices]
[Pages 26184-26197]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 04-10305]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility Operating
Licenses
Involving No Significant Hazards Considerations
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC staff) is publishing this regular biweekly notice.
The Act requires the Commission publish notice of any amendments
issued, or proposed to be issued and grants the Commission the
authority to issue and make immediately effective any amendment to an
operating license upon a determination by the Commission that such
amendment involves no significant hazards consideration,
notwithstanding the pendency before the Commission of a request for a
hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, April 16 through April 29, 2004. The last
biweekly notice was published on April 27, 2004 (69 FR 22877).
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
[[Page 26185]]
no significant hazards consideration. Under the Commission's
regulations in 10 CFR 50.92, this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR Part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the
[[Page 26186]]
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; (2) courier, express mail, and expedited delivery
services: Office of the Secretary, Sixteenth Floor, One White Flint
North, 11555 Rockville Pike, Rockville, Maryland 20852, Attention:
Rulemaking and Adjudications Staff; (3) E-mail addressed to the Office
of the Secretary, U.S. Nuclear Regulatory Commission,
[email protected]; or (4) facsimile transmission addressed to the
Office of the Secretary, U.S. Nuclear Regulatory Commission,
Washington, DC, Attention: Rulemakings and Adjudications Staff at (301)
415-1101, verification number is (301) 415-1966. A copy of the request
for hearing and petition for leave to intervene should also be sent to
the Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and it is requested that copies be
transmitted either by means of facsimile transmission to 301-415-3725
or by 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 requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition,
request and/or the contentions should be granted based on a balancing
of the factors specified in 10 CFR 2.309(a)(1)(i)-(viii).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster
Creek Nuclear Generating Station, Ocean County, New Jersey
Date of amendment request: March 23, 2004.
Description of amendment request: The licensee requested to revise
the Technical Specifications (TSs), deleting the requirements for the
Independent Onsite Safety Review Group (IOSRG) and locating them intact
to a licensee-controlled document, the company-wide Quality Assurance
Topical Report (QATR). The requirements are in the administrative
section of the TSs and include IOSRG organization, function
description, member qualifications, and recordkeeping. The relocation
is proposed per the guidance of Nuclear Regulatory Commission (NRC)
Administrative Letter 95-06. In addition, the licensee proposed to
correct the reference for facility activities audits from a site-
specific document to the company-wide QATR.
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 reviewed the licensee's analysis and has
performed its own analysis as follows:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed amendment does not affect assumptions contained in
the current licensing basis plant safety analyses, will not lead to
physical changes of a plant structure, system, or component (SSC), and
will not alter the method of operation of any SSC. The IOSRG
requirements and conduct of IOSRG activities were not factors in any
previously analyzed accident or transient scenarios, and thus, the
elimination of IOSRG requirements from the TSs will have no effect on
the probability of occurrence and consequences of any previously
analyzed accident or transient.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed amendment is not the result of a design change or
method of operation change, and will not lead to such changes. Hence
no, new or different kind of accident can be created from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed amendment does not involve any change to current
analysis models, assumptions, limiting conditions for operation,
operational parameters, action statements, and surveillance
requirements. Hence, there is no reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
its own analysis, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice
President, General Counsel and Secretary, Exelon Generation Company,
LLC, 300 Exelon Way, Kennett Square, PA 19348.
NRC Section Chief: Richard J. Laufer.
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: March 23, 2004.
Description of amendment request: The licensee requested to revise
the Technical Specifications (TSs), deleting the requirements for the
Independent Onsite Safety Review Group (IOSRG) and locating them intact
to a licensee-controlled document, the company-wide Quality Assurance
Topical Report (QATR). The requirements are in the administrative
section of the TSs and include IOSRG organization, function
description, member qualifications, and recordkeeping. The relocation
is proposed per the guidance of Nuclear Regulatory Commission (NRC)
Administrative Letter 95-06. In addition, the licensee proposed to
correct the reference for facility activities audits from a site-
specific document to the company-wide QATR.
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 reviewed the licensee's analysis and has
performed its own analysis as follows:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed amendment does not affect assumptions contained in
the current licensing basis plant safety analyses, will not lead to
physical changes of a plant structure, system, or component (SSC), and
will not alter the method of operation of any SSC. The IOSRG
requirements and conduct of IOSRG activities were not factors in any
previously analyzed accident or transient scenarios, and thus, the
elimination of IOSRG requirements from the TSs will have no effect on
the probability of occurrence and consequences of any previously
analyzed accident or transient.
[[Page 26187]]
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
No. The proposed amendment is not the result of a design change or
method of operation change, and will not lead to such changes. Hence,
no new or different kind of accident can be created from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No. The proposed amendment does not involve any change to current
analysis models, assumptions, limiting conditions for operation,
operational parameters, action statements, and surveillance
requirements. Hence, there is no reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
its own analysis, it appears that the three standards of 10 CFR
50.92(c) are satisfied. Therefore, the NRC staff proposes to determine
that the proposed amendment involves no significant hazards
consideration.
Attorney for licensee: Edward J. Cullen, Jr., Esquire, Vice
President, General Counsel and Secretary, Exelon Generation Company,
LLC, 300 Exelon Way, Kennett Square, PA 19348
NRC Section Chief: Richard J. Laufer.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No.
50-458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: March 4, 2004.
Description of amendment request: The proposed amendment deletes
requirements from the Technical Specifications to maintain hydrogen
recombiners and hydrogen and oxygen monitors. Licensees were generally
required to implement upgrades as described in NUREG-0737,
``Clarification of TMI [Three Mile Island] Action Plan Requirements,''
and Regulatory Guide (RG) 1.97, ``Instrumentation for Light-Water-
Cooled Nuclear Power Plants to Assess Plant and Environs Conditions
During and Following an Accident.'' Implementation of these upgrades
was an outcome of the lessons learned from the accident that occurred
at TMI, Unit 2. Requirements related to combustible gas control were
imposed by Order for many facilities and were added to or included in
the technical specifications (TS) for nuclear power reactors currently
licensed to operate. The revised 10 CFR 50.44, ``Standards for
Combustible Gas Control System in Light-Water-Cooled Power Reactors,''
eliminated the requirements for hydrogen recombiners and relaxed safety
classifications and licensee commitments to certain design and
qualification criteria for hydrogen and oxygen monitors.
The NRC staff issued a notice of availability of a model no
significant hazards consideration determination 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 March 4, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The 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 and oxygen 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 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 and oxygen 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. Also, as part of the
rulemaking to revise 10 CFR 50.44, the Commission found that
Category 2, as defined in RG 1.97, is an appropriate categorization
for the oxygen monitors, because the monitors are required to verify
the status of the inert containment.
The regulatory requirements for the hydrogen and oxygen 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,
classification of the oxygen monitors as Category 2 and removal of
the hydrogen and oxygen monitors from TS will not prevent an
accident management strategy through the use of the severe accident
management guidelines, the emergency plan, the emergency operating
procedures, and site survey monitoring that support modification of
emergency plan protective action recommendations.
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen and oxygen 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 and oxygen monitor requirements,
including removal of these requirements from TS, will not result in
any failure mode not previously analyzed. The hydrogen recombiner
and hydrogen and oxygen monitor equipment was intended to mitigate a
design-basis hydrogen release. The hydrogen recombiner and hydrogen
and oxygen 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 and oxygen 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
[[Page 26188]]
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.
Category 2 oxygen monitors are adequate to verify the status of an
inerted containment.
Therefore, this change does not involve a significant reduction
in the margin of safety. The intent of the requirements established
as a result of the TMI, Unit 2 accident can be adequately met
without reliance on safety-related oxygen monitors. Removal of
hydrogen and oxygen monitoring from TS will not result in a
significant reduction in their functionality, reliability, and
availability.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas.
Date of amendment request: April 15, 2004.
Description of amendment request: The proposed amendment would
change the reactor coolant system (RCS) pressure/temperature (P/T)
limits in the technical specifications (TSs) by providing a single
maximum cooldown rate instead of a variable cooldown rate and by
revising the cooldown curve with one that is slightly more restrictive.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of occurrence of an accident previously
evaluated for ANO-2 [Arkansas Nuclear One, Unit 2] is not altered by
the proposed amendment to the TSs. The accidents remain the same as
currently analyzed in the ANO-2 Safety Analysis Report (SAR) as a
result of the change to the cooldown P/T limits. The new P/T
cooldown limits were based on NRC [Nuclear Regulatory Commission]
accepted methodologies along with ASME [American Society of
Mechanical Engineers] Code [Boiler and Pressure Vessel Code]
alternatives. The proposed change does not impact the integrity of
the reactor coolant pressure boundary (RCPB) (i.e., there is no
change to the operating pressure, materials, loadings, etc.) as a
result of this change. In addition, there is no increase in the
potential for the occurrence of a loss of coolant accident. The
proposed P/T cooldown limit curve is not considered to be an
initiator or contributor to any accident currently evaluated in the
ANO-2 SAR. The revised P/T cooldown limits ensure the long term
integrity of the RCPB. For each analyzed transient and steady state
condition, the allowable pressure was determined as a function of
reactor coolant temperature considering postulated flaws in the
reactor vessel beltline, inlet nozzle, outlet nozzle, and closure
head flange.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the P/T limits will not create a new
accident scenario. The requirements to have P/T protection are part
of the ANO-2 licensing basis. The proposed change in the P/T
cooldown limits is based on NRC approved methodologies performed by
Framatome ANP. This methodology complies with NRC and ASME
requirements for protecting the RCS. Therefore, the revised P/T
cooldown limits provide protection of the RCS from limiting
transients during normal cooldown.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The revision of the P/T limits and curves will ensure that ANO-2
continues to operate within the operating margins of the ASME Code.
The application of ASME Code Cases N-640 and N-588 presents
alternative procedures for calculating P/T temperatures and
pressures. These Code Cases allow certain assumptions to be
conservatively reduced. However, the procedures allowed by these
Code Cases still provide sufficient conservatism and ensure an
adequate margin of safety in the development of P/T operating and
pressure test limits to prevent non-ductile fractures.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio.
Date of amendment request: March 31, 2004.
Description of amendment request: This license amendment request
(LAR) proposes to eliminate the Technical Specification Surveillance
Requirements (SRs) that require each Main Steam Safety/Relief Valve (S/
RV) to open during the manual actuation portion of testing the valves.
In accordance with 10 CFR 50.55a, ``Codes and Standards,'' paragraph
(a)(3), this request also includes Relief Request VR-13. VR-13 is a
request for relief from the requirements of ASME/American National
Standards Institute (ANSI) Operation and Maintenance (OM) of Nuclear
Power Plants, OM-1995, Appendix I, Section 3.4.1(d) that after
isolation, the S/RVs are manually opened and closed.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed [License Amendment Request] LAR modifies TS
3.4.4.3, SR 3.5.1.7, and SR 3.6.1.6.1 to allow the uncoupling of the
S/RV stem from the S/RV actuator during manual actuation. The
proposed LAR does not change the manner in which the S/RVs are
intended to operate.
The performance of S/RV testing provides assurance that the S/
RVs are capable of depressurizing the Reactor Pressure Vessel (RPV).
This will protect the RPV from over pressurization and allows the
combination of the Low Pressure Coolant Injection (LPCI) system and
the Low Pressure Core Spray (LPCS) system to inject into the RPV as
designed. The proposed testing requirements are sufficient to
provide confidence that the S/RVs, [Automatic Depressurization
System] ADS valves, and the [Low-Low Set] LLS valves will perform
their intended design safety functions.
Therefore, the proposed LAR does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed LAR changes TS [Surveillance Requirements] SR
3.4.4.3, SR 3.5.1.7, and SR 3.6.1.6.1. The changes to these SRs do
not effect the assumed accident performance of the S/RVs, nor any
plant structure, system or component previously evaluated. The LAR
does not install any new equipment, nor does it cause existing
equipment to be operated in a new or
[[Page 26189]]
different manner. The S/RVs continue to be bench-tested to verify
the safety and relief modes of valve operation. The changes will
allow the testing of the manual actuation electrical circuitry,
solenoid and air control valve, and the actuator without causing the
S/RV to open. No setpoints are being changed which would alter the
dynamic response of plant equipment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from that previously evaluated.
3. The proposed change will not involve a single reduction in
the margin of safety.
The proposed LAR will allow the uncoupling of the S/RV stem from
the other components associated with the manual actuation testing of
the S/RVs. The proposed changes will allow the testing of the manual
actuation electrical circuitry, solenoid and air control valve, and
the actuator without causing the S/RV to open. The S/RVs will
continue to be manually actuated by the bench-test of the valve
control system and setpoint testing program prior to installation in
the plant. The changes do not effect the valve setpoint or
operational criteria that directs the S/RVs to be manually opened
during plant transients. There are no changes which alter the
setpoints at which protective actions are initiated.
Therefore, the proposed change does not involve a significant
reduction in any margins of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry
Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio
Date of amendment request: April 5, 2004.
Description of amendment request: This license amendment request
(LAR) proposes to modify the existing Minimum Critical Power Ratio
(MCPR) Safety Limit contained in Technical Specification 2.1.1.2.
Specifically, the change modifies the MCPR Safety Limit values, as
calculated by Global Nuclear Fuel (GNF), by decreasing the limit for
two recirculating loop operation from 1.10 to 1.08, and decreasing the
limit for single recirculation loop operation from 1.11 to 1.10. The
change resulted from a core reload analysis performed during the PNPP
Fuel Cycle 10.
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.
Perry Nuclear Power Plant (PNPP) Updated Safety Analysis Report
(USAR) Section 4.2, ``Fuel System Design,'' states the PNPP fuel
system design bases are provided in the General Electric Topical
Report, NEDE-24011-P-A, ``General Electric Standard Application for
Reactor Fuel (GESTAR II).'' The Minimum Critical Power Ratio (MCPR)
Safety Limit is one of the limits used to protect the fuel in
accordance with the design basis. The MCPR Safety Limit establishes
a margin to the onset of transition boiling. The basis of the MCPR
Safety Limit remains the same, ensuring that greater than 99.9 % of
all fuel rods in the core avoid transition boiling. The methodology
used to determine the MCPR Safety Limit values is contained within
GESTAR II and is NRC approved. The change does not result in any
physical plant modifications or physically affect any plant
components. As a result, there is no increase in the probability of
occurrence of a previously analyzed accident.
The fundamental sequences of accidents and transients have not
been altered. The Safety Limit MCPR is established to avoid fuel
damage in response to anticipated operational occurrences.
Compliance with a MCPR Safety Limit greater than or equal to the
calculated value will ensure that less than 0.1% of the fuel rods
will experience boiling transition. This in turn ensures fuel damage
does not occur following transients due to excessive thermal
stresses on the fuel cladding. The MCPR Operating Limits are set
higher (i.e., more conservative) than the Safety Limit such that
potentially limiting plant transients prevent the MCPR from
decreasing below the MCPR Safety Limit during the transient.
Therefore, there is no impact on any of the limiting USAR Appendix
15B transients. The radiological consequences remain the same as
previously stated in the USAR. Therefore, the consequences of an
accident do not increase over previous evaluations in the USAR.
Therefore, the proposed LAR does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The MCPR Safety Limit basis is preserved, which is to ensure
that transition boiling does not occur in at least 99% of the fuel
rods in the core as a result of the postulated limiting transient.
The values are calculated in accordance with GESTAR II. The GESTAR
II analyses have been accepted by the NRC. The MCPR Safety Limit is
one of the limits established to ensure the fuel is protected in
accordance with the design basis. The function, location, operation,
and handling of the fuel remain unchanged. No changes in the design
of the plant or the method of operating the plant are associated
with these revised safety limit valves. Therefore, no new or
different kind of accident from any previously evaluated is created.
3. The proposed change will not involve a single reduction in
the margin of safety.
This change revises the PNPP MCPR Safety Limit values. The new
MCPR Safety Limit values reflect changes due to Cycle 10 core
design, but do not alter the design or function of any plant system,
including the fuel. The new MCPR Safety Limit values were calculated
using NRC-approved methods described in GESTAR II. The proposed MCPR
Safety Limit values continue to satisfy the fuel design safety
criteria which ensures that transition boiling does not occur in at
least 99.9% of the fuel rods in the core as a result of the
postulated limiting transient. Therefore, the proposed values for
the MCPR Safety Limit do not involve a significant reduction in a
safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry
Nuclear Power Plant (PNPP), Unit 1, Lake County, Ohio
Date of amendment request: April 5, 2004.
Description of amendment request: This license amendment request
(LAR) proposes to modify the existing Minimum Critical Power Ratio
(MCPR) Safety Limit contained in Technical Specification 2.1.1.2.
Specifically, the change modifies the MCPR Safety Limit values, as
calculated by Global Nuclear Fuel (GNF), by decreasing the limit for
two recirculating loop operation from 1.10 to 1.08, and decreasing the
limit for single recirculation loop operation from 1.11 to 1.10. The
change resulted from a core reload analysis performed during the PNPP
Fuel Cycle 10.
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.
Perry Nuclear Power Plant (PNPP) Updated Safety Analysis Report
(USAR) Section 4.2, ``Fuel System Design,'' states the PNPP fuel
system design bases are provided in the General Electric Topical
Report, NEDE-
[[Page 26190]]
24011-P-A, ``General Electric Standard Application for Reactor Fuel
(GESTAR II).'' The Minimum Critical Power Ratio (MCPR) Safety Limit
is one of the limits used to protect the fuel in accordance with the
design basis. The MCPR Safety Limit establishes a margin to the
onset of transition boiling. The basis of the MCPR Safety Limit
remains the same, ensuring that greater than 99.9 % of all fuel rods
in the core avoid transition boiling. The methodology used to
determine the MCPR Safety Limit values is contained within GESTAR II
and is NRC approved. The change does not result in any physical
plant modifications or physically affect any plant components. As a
result, there is no increase in the probability of occurrence of a
previously analyzed accident.
The fundamental sequences of accidents and transients have not
been altered. The Safety Limit MCPR is established to avoid fuel
damage in response to anticipated operational occurrences.
Compliance with a MCPR Safety Limit greater than or equal to the
calculated value will ensure that less than 0.1% of the fuel rods
will experience boiling transition. This in turn ensures fuel damage
does not occur following transients due to excessive thermal
stresses on the fuel cladding. The MCPR Operating Limits are set
higher (i.e., more conservative) than the Safety Limit such that
potentially limiting plant transients prevent the MCPR from
decreasing below the MCPR Safety Limit during the transient.
Therefore, there is no impact on any of the limiting USAR Appendix
15B transients. The radiological consequences remain the same as
previously stated in the USAR. Therefore, the consequences of an
accident do not increase over previous evaluations in the USAR.
Therefore, the proposed LAR does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change would not create the possibility of a new
or different kind of accident from any previously evaluated.
The MCPR Safety Limit basis is preserved, which is to ensure
that transition boiling does not occur in at least 99% of the fuel
rods in the core as a result of the postulated limiting transient.
The values are calculated in accordance with GESTAR II. The GESTAR
II analyses have been accepted by the NRC. The MCPR Safety Limit is
one of the limits established to ensure the fuel is protected in
accordance with the design basis. The function, location, operation,
and handling of the fuel remain unchanged. No changes in the design
of the plant or the method of operating the plant are associated
with these revised safety limit valves. Therefore, no new or
different kind of accident from any previously evaluated is created.
3. The proposed change will not involve a single reduction in
the margin of safety.
This change revises the PNPP MCPR Safety Limit values. The new
MCPR Safety Limit values reflect changes due to Cycle 10 core
design, but do not alter the design or function of any plant system,
including the fuel. The new MCPR Safety Limit values were calculated
using NRC-approved methods described in GESTAR II. The proposed MCPR
Safety Limit values continue to satisfy the fuel design safety
criteria which ensures that transition boiling does not occur in at
least 99.9% of the fuel rods in the core as a result of the
postulated limiting transient. Therefore, the proposed values for
the MCPR Safety Limit do not involve a significant reduction in a
safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: August 27, 2003.
Description of amendment requests: The proposed amendments would
amend Unit 1 and Unit 2 Technical Specifications (TS) 4.0.3. TS 4.0.3
describes the relationship between meeting the surveillance requirement
and operability. The proposed change will modify TS 4.0.3 to allow a
missed surveillance to be completed within 24 hours or up to the limit
of the specified interval, whichever is greater. Additionally, a
statement that a risk evaluation shall be performed for any
surveillance delayed greater than 24 hours and that the risk impact
shall be managed is being added to the TS.
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?
No.
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
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?
No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed)
or a change in the methods governing normal plant operation. A
missed surveillance will not, in and of itself, introduce new
failure modes or effects and any increased chance that a standby
system might fail to perform its safety function due to a missed
surveillance would not, in the absence of other unrelated failures,
lead to an accident beyond those previously evaluated. The addition
of a requirement to assess and manage the risk introduced by the
missed surveillance will further minimize possible concerns. The
format changes are intended to improve readability and appearance
and do not alter any requirements. Thus, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
No.
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the limiting condition for
operation is met. Failure to perform a surveillance within the
prescribed frequency does not cause equipment to become inoperable.
The only effect of the additional time allowed to perform a missed
surveillance on the margin of safety is the extension of the time
until inoperable equipment is discovered to be inoperable by the
missed surveillance. However, given the rare occurrence of
inoperable equipment, and the rare occurrence of a missed
surveillance, a missed surveillance on inoperable equipment would be
very unlikely. This must be balanced against the real risk of
manipulating the plant equipment or condition to perform the missed
surveillance. In addition, parallel trains and alternate equipment
are typically available to perform the safety function of the
equipment not tested. Thus, there is confidence that the equipment
can perform its assumed safety function. The format changes are
intended to improve readability and appearance and do not alter any
requirements. Therefore, this change does not involve a significant
reduction in a margin of safety.
Based upon the reasoning presented above, the requested change
does not involve a significant hazards consideration.
[[Page 26191]]
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: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: February 14, 2004.
Description of amendment requests: The proposed amendments would
revise the Technical Specifications (TS) governing containment
penetrations and the Containment Purge and Exhaust Isolation System,
which are applicable during CORE ALTERATIONS and movement of irradiated
fuel, such that those TSs are only applicable during the movement of
recently irradiated fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No.
The proposed changes incorporate line item improvements that are
based on assumptions in the postulated fuel handling accident (FHA)
analysis. These proposed changes remove the applicability of the
Technical Specifications (TS) governing containment penetrations and
the Containment Purge and Exhaust Isolation System when handling
fuel assemblies that have decayed for a sufficient period of time.
The containment penetration and Containment Purge and Exhaust
Isolation System are not initiators to any accident previously
evaluated. As a result, the probability of any accident previously
evaluated is not significantly increased. The only previously
analyzed accident affected by the proposed change is an FHA. The
current, Nuclear Regulatory Commission (NRC)-approved analysis of an
FHA does not assume any holdup of the postulated radioactivity
release by the containment building nor does it assume the operation
of the Containment Purge and Exhaust Isolation System. As a result,
the proposed change does not affect the assumed mitigation or
consequences of that event.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes incorporate line item improvements that are
based on assumptions in the postulated FHA analysis. These proposed
changes remove the applicability of the TS governing containment
penetrations and the Containment Purge and Exhaust Isolation System
when handling fuel assemblies that have decayed for a sufficient
period of time. The proposed changes do not involve the addition or
modification of equipment nor do they alter the design of the plant.
The revised operations are consistent with the FHA analysis and do
not require any new or different ways of operating the plant
equipment.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes incorporate line item improvements that are
based on assumptions in the postulated FHA analysis. These proposed
changes remove the applicability of the TS governing containment
penetrations and the Containment Purge and Exhaust Isolation System
when handling fuel assemblies that have decayed for a sufficient
period of time. The calculated offsite and Control Room doses
resulting from an FHA are not affected by this change as the
proposed TS changes are revised to be consistent with the
assumptions used in these analyses. As a further measure, [Indiana
Michigan Power Company] I&M has committed to maintaining a single
normal or contingency method to promptly close containment
penetrations following an FHA. These prompt methods will enable the
ventilation systems to draw the release from a postulated FHA such
that it can be treated and monitored. This will provide a further
margin of safety beyond that assumed in the accident analysis.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107
NRC Section Chief: L. Raghavan.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: February 14, 2004.
Description of amendment requests: The proposed amendments would
modify the Technical Specification (TS) 3.9.2 limiting condition for
operation, to delete TS Surveillance Requirements (SRs) 4.9.2.a and b
for the Source Range Neutron Flux Monitor channel functional test, to
revise SR 4.9.2.c for the channel check test, and to add a requirement
to perform a channel calibration every 18 months as well as revise TS
4.10.4.2 and 4.10.3.2 (Units 1 and 2 respectively) for Intermediate and
Power Range channel functional test.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment replaces the Technical Specification (TS)
3.9.2 limiting condition for operation (LCO) requirement for an
audible indication in the containment (both units) and control room
(Unit 2) with a requirement that a source range audible count rate
circuit be operable. This involves no physical changes to the plant,
and maintains the capability to alert the operators to changes in
core reactivity. Thus, neither the probability of an accident nor
the consequences are significantly increased.
The proposed amendment revises the TS SR for the Power Range,
Intermediate Range, and the Source Range Neutron Flux Monitors to
reduce redundant testing. Surveillance testing is not an initiator
to any accident previously evaluated. As a result, the proposed
changes will not result in a significant increase in the probability
of any accident previously evaluated.
The Power Range, Intermediate Range, and the Source Range
Neutron Flux Monitors are used to detect and mitigate accidents
previously evaluated. However, the LCOs continue to require the
subject flux monitors to be operable and the remaining testing is
sufficient to ensure the flux monitors are capable of performing
their detection and mitigation functions. Thus, the consequences of
an accident are not significantly changed.
Based on the above, [Indiana Michigan Power Company] I&M
concludes that 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 and accident previously evaluated?
Response: No.
The proposed amendment replaces the TS 3.9.2 LCO requirement for
an audible indication in the containment (both units) and control
room (Unit 2) with a requirement that a source range audible count
rate circuit be operable.
[[Page 26192]]
The change does not make any physical changes to the plant.
Thus, the change does not create the possibility of a new or
different kind of accident.
The proposed amendment revises the TS SR for the Power Range,
Intermediate Range, and the Source Range Neutron Flux Monitors to
reduce redundant testing. The proposed changes do not change the
design function or operation of any plant equipment. No new failure
mechanisms, malfunctions, or accident initiators are being
introduced by the proposed changes. Therefore, the proposed change
does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment replaces the TS 3.9.2 LCO requirement for
an audible indication in the containment (both units) and control
room (Unit 2) with a requirement that a source range audible count
rate circuit be operable. The source range audible count rate
circuit will continue to perform its function of alerting the
operators to changes in core reactivity.
The proposed amendment revises the TS Surveillance Requirement
(SR) for the Power Range, Intermediate Range, and the Source Range
Neutron Flux Monitors to reduce redundant testing. The elimination
of redundant testing does not reduce the reliability of the tested
flux monitors. The flux monitors continue to be tested in a manner
and at a frequency necessary to provide confidence that the
equipment can perform its assumed safety function.
Therefore, there is no 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: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107
NRC Section Chief: L. Raghavan.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment requests: April 6, 2004.
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS) design features for fuel assemblies
and new fuel storage criticality limitations. In addition, the licensee
requests approval of the criticality analysis methodology supporting
the spent fuel storage rack and new fuel storage rack in accordance
with 10 CFR 50.59(c)(2)(viii).
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 proposed Technical Specification (TS) changes allow the
zirconium-based alloy, M5, to be used in addition to Zircaloy-4 and
ZIRLO in Donald C. Cook Nuclear Plant fuel assemblies. TS changes
are also proposed to allow Gadolinia to be used in fuel assemblies
in the new fuel storage racks to ensure adequate reactivity margin.
In addition, methodology changes were proposed for a criticality
analysis supporting new and spent fuel rack design criteria. M5 is a
Nuclear Regulatory Commission (NRC)-approved alloy for fuel cladding
and Gadolinia is an NRC-approved fuel burnable absorber used in the
maintenance of reactivity margin in the new fuel storage rack. The
use of NRC-approved cladding and fuel absorbers and methodology
changes to criticality analyses to support TS design criteria for
the spent and new fuel storage racks are not initiators of any
accident previously evaluated. As a result, the probability of any
accident previously evaluated is not significantly increased. M5
cladding has been shown to meet all 10 CFR 50.46 acceptance
criteria. Analysis has shown that the use of Gadolinia assures
sufficient reactivity margin to prevent a criticality accident in
the new fuel storage rack. Changes in methodology for criticality
analyses were performed to demonstrate TS requirements are met or to
support proposed TS changes and do not affect plant equipment.
Therefore, the consequences of an accident are not significantly
increased.
Therefore, the proposed change does not involve a significant
increase in the probability of occurrence or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change to use the M5 alloy is based on an NRC-
approved topical report which demonstrates that the material
properties of the M5 alloy are not significantly different from
those of Zircaloy-4. The design and performance criteria continue to
be met and no new failure mechanisms have been identified.
Therefore, M5 fuel rod cladding and fuel assembly structural
components will perform similarly to those fabricated from Zircaloy-
4, thus precluding the possibility of the fuel becoming an accident
initiator and causing a new or different type of accident.
The proposed TS change to use Gadolinia to ensure adequate
reactivity margin for higher enrichment fuel assemblies prevents
reactivity limits from being exceeded. An NRC-approved topical
report demonstrates that Gadolinia is acceptable for use in fuel
assemblies. The proposed change only modifies the type of fuel
burnable absorber and does not affect any permanent plant equipment
or plant operating procedures, and can not be an initiator of an
accident.
The proposed criticality analysis supports TS design criteria
for spent and new fuel racks. The analysis evaluates reactivity
margin based on conservative assumptions on fuel assembly design and
burnup and does not affect any plant equipment. The criticality
analysis can not be an initiator of an accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed TS change to allow the use of fuel rods clad with
the M5 alloy does not change the reactor fuel reload design and
safety limits. For each cycle reload core, the fuel assembly design
and core configuration are evaluated using NRC-approved reload
design methods, including consideration of the core physics analysis
peaking factors and core average linear heat rate effects. The
design basis and modeling techniques for fuel assemblies with
Zircaloy-4 and ZIRLO clad fuel rods remain valid for fuel assemblies
with M5 clad fuel rods. Use of the M5 alloy as cladding material has
no effect on the criticality analysis for the spent fuel storage
racks and the new fuel storage racks. Furthermore, it has no effect
on the thermal-hydraulic and structural analysis for the spent fuel
pool. Therefore, the design and safety analysis limits specified in
the TS are maintained with this proposed change.
The proposed TS change to use Gadolinia as a fuel burnable
absorber for fuel assemblies with higher enrichments of Uranium-235
to ensure proper reactivity control in the spent fuel storage rack
is consistent with the current method of reducing reactivity of high
enrichment fuel assemblies. Each method reduces the equivalent
uranium enrichment to below that found acceptable by the NRC for
safe storage of new fuel.
The proposed criticality analyses use NRC-approved codes with a
methodology different than previously approved by the NRC. The
criticality analysis results for the spent fuel storage rack flooded
with unborated water condition and for the new fuel storage rack
moderated by aqueous foam condition remain less than the limiting TS
values. Analysis results for the new fuel storage rack flooded with
unborated water condition are consistent with previous analysis
results.
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
[[Page 26193]]
Attorney for licensee: David W. Jenkins, Esq., 500 Circle Drive,
Buchanan, MI 49107.
NRC Section Chief: L. Raghavan.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Units 1 and 2, Somervell County, Texas
Date of amendment request: April 13, 2004.
Brief description of amendments: The requested amendments will
revise the Technical Specification 3.3.2, ``Engineered Safety Features
Actuation System (ESFAS) Instrumentation,'' to revise the trip setpoint
allowable value for Refueling Water Storage Tank (RWST) Low-Low Level
(ESFAS function 7.b) for Unit 2 to be the same as it is for Unit 1.
Also, the frequency of calibration of the RWST water level transmitters
will be revised from once in 9 months to once in 18 months.
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 by focusing on the three standards set forth in 10 CFR
50.92. The licensee's analysis of three standards 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 in the trip setpoint allowable value for
Unit 2 Refueling Water Storage Tank (RWST) Low-Low Level has no
impact on the probability of any accident previously evaluated.
Since none of the accident analyses are affected by this change, the
consequences of all previously evaluated accidents remain unchanged.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of these changes. There will be no adverse effect or challenges
imposed on any safety-related system as a result of these changes.
There are no changes in the method by which any safety-related plant
system performs its safety function. Overall protection system
performance will remain within the bounds of the previously
performed accident analyses and the protection systems will continue
to function in a manner consistent with the plant design basis. The
proposed changes do not affect the probability of any event
initiators. The proposed changes do not alter any assumptions or
change any mitigation actions in the radiological consequence
evaluations in the Final Safety Analysis Report (FSAR).
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not affect the acceptance criteria for
any analyzed event nor is there a change to any Safety Analysis
Limit (SAL). There will be no effect on the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined nor will there be any effect on those plant
systems necessary to assure the accomplishment of protection
functions. There will be no impact on the overpower limit, the
Departure from Nucleate Boiling Ratio (DNBR) limits, the Heat Flux
Hot Channel Factor (FQ), the Nuclear Enthalpy Rise Hot
Channel Factor (F[Delta]H), the Loss of Coolant Accident Peak
Centerline Temperature (LOCA PCT), peak local power density, or any
other margin of safety. The radiological dose consequence acceptance
criteria listed in the Standard Review Plan will continue to be met.
Therefore the proposed change does not involve a reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involves no significant hazards consideration.
Attorney for licensee: George L. Edgar, Esq., Morgan, Lewis and
Bockius, 1800 M Street, NW., Washington, DC 20036.
NRC Section Chief: Robert A. Gramm.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of amendment request: April 8, 2004.
Description of amendment request: The proposed amendment revises TS
5.5.7, ``Reactor Coolant Pump Flywheel Inspection Program,'' to extend
the allowable inspection interval to 20 years.
The NRC staff issued a notice of opportunity for comment in the
Federal Register on June 24, 2003 (68 FR 37590), on possible amendments
to extend the inspection interval for reactor coolant pump (RCP)
flywheels, 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 October 22, 2003, (68
FR 60422). The licensee affirmed the applicability of the model NSHC
determination in its application dated April 8, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
The proposed change to the RCP flywheel examination frequency
does not change the response of the plant to any accidents. The RCP
will remain highly reliable and the proposed change will not result
in a significant increase in the risk of plant operation. Given the
extremely low failure probabilities for the RCP motor flywheel
during normal and accident conditions, the extremely low probability
of a loss-of-coolant accident (LOCA) with loss of offsite power
(LOOP), and assuming a conditional core damage probability (CCDP) of
1.0 (complete failure of safety systems), the core damage frequency
(CDF) and change in risk would still not exceed the NRC's acceptance
guidelines continued in Regulatory Guide (RG) 1.174 (<1.0E-6 per
year). Moreover, considering the uncertainties involved in this
evaluation, the risk associated with the postulated failure of an
RCP motor flywheel is significantly low. Even if all four RCP motor
flywheels are considered in the bounding plant configuration case,
the risk is still acceptably low.
The proposed change does not adversely affect accident
initiators or precursors, nor alter the design assumptions,
conditions, or configuration of the facility, or the manner in which
the plant is operated and maintained; alter or prevent the ability
of structures, systems, components (SSCs) from performing their
intended function to mitigate the consequences of an initiating
event within the assumed acceptance limits; or affect the source
term, containment isolation, or radiological release assumptions
used in evaluating the radiological consequences of an accident
previously evaluated. Further, the proposed change does not increase
the type or amount of radioactive effluent that may be released
offsite, nor significantly increase individual or cumulative
occupational/public radiation exposure. The proposed change is
consistent with the safety analysis assumptions and resultant
consequences. Therefore, the proposed 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 accident previously
evaluated.
The proposed change in flywheel inspection frequency does not
involve any change in the design or operation of the RCP. Nor does
the change to examination frequency affect any existing accident
scenarios, or create any new or different accident scenarios.
Further, the change does not involve a physical alteration of the
plant
[[Page 26194]]
(i.e., no new or different type of equipment will be installed) or
alter the methods governing normal plant operation. In addition, the
change does not impose any new or different requirements or
eliminate any existing requirements, and does not alter any
assumptions made in the safety analysis. The proposed change is
consistent with the safety analysis assumptions and current plant
operating practice. Therefore, the proposed change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
Criterion 3--The proposed change does not involve a significant
reduction in a margin of safety.
The proposed change does not alter the manner in which safety
limits, limiting safety system settings, or limiting conditions for
operation are determined. The safety analysis acceptance criteria
are not impacted by this change. The proposed change will not result
in plant operation in a configuration outside of the design basis.
The calculated impact on risk is insignificant and meets the
acceptance criteria contained in RG 1.174. There are no significant
mechanisms for inservice degradation of the RCP flywheel. Therefore,
the proposed 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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: April 8, 2004.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4 exceptions in individual TSs would be
eliminated, several notes or specific exceptions are revised or deleted
to reflect the related changes to LCO 3.0.4, and Surveillance
Requirement (SR) 3.0.4 is revised to reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated April 8, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
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
[[Page 26195]]
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].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station (OCNGS), Ocean County, New Jersey
Date of application for amendment: December 20, 2002, as
supplemented on May 30, September 10, and November 3, 2003.
Brief description of amendment: The amendment authorized the
revision of the OCNGS Updated Final Safety Analysis Report (UFSAR) to
reflect implementation of the Boiling Water Reactor Vessel and
Internals Project reactor pressure vessel Integrated Surveillance
Program (ISP) as the basis for demonstrating compliance with the
requirements of Appendix H, ``Reactor Vessel Material Surveillance
Program Requirements,'' to Title 10 of the Code of Federal Regulations,
Part 50.
Date of Issuance: April 27, 2004.
Effective date: The amendment is effective immediately. The ISP
shall be implemented prior to the next scheduled reactor vessel
surveillance capsule removal. The UFSAR is to be revised to reflect use
of the ISP in accordance with the schedule of 10 CFR 50.71(e).
Amendment No.: 242.
Facility Operating License No. DPR-16: Amendment revised the
Operating License DPR-16.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5669).
The May 30, September 10, and November 3, 2003, letters provided
clarifying information within the scope of the original application,
and did not change the staff's initial proposed no significant hazards
consideration determination. The Commission's related evaluation of
this amendment is contained in a Safety Evaluation dated April 27,
2004. No significant hazards consideration comments received: No.
Calvert Cliffs Nuclear Power Plant, Inc., Docket Nos. 50-317 and 50-
318, Calvert Cliffs Nuclear Power Plant, Unit Nos. 1 and 2, Calvert
County, Maryland
Date of application for amendments: May 12, 2003, as supplemented
December 5, 2003, February 23, 2004, March 26, 2004 and April 6, 2004.
Brief description of amendments: These amendments extend several
Required Action completion times for inoperable diesel generators
identified in Technical Specification 3.8.1, ``AC Sources Operating.''
Date of issuance: April 13, 2004.
Effective date: As of the date of issuance to be implemented within
30 days.
Amendment Nos.: 265 and 242.
Renewed Facility Operating License Nos. DPR-53 and DPR-69:
Amendments revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37576). The licensee's December 5, 2003, February 23, 2004, March 26,
2004, and April 6, 2004, letters provided additional information that
clarified the application, did not change the scope of the application
as originally noticed, and did not change the staff's original proposed
no significant hazards consideration determination.
The Commission's related evaluation of these amendments is
contained in a Safety Evaluation dated April 13, 2004.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendments: June 27, 2003.
Brief description of amendments: The amendments modified Technical
Specification 4.0.5.f and associated Bases, and Bases Section 3/4.4.8,
with regard to the commitment to perform piping inspections in
accordance with Generic Letter 88-01, by adding the words ``or in
accordance with alternate measures approved by the NRC staff.''
Date of issuance: As of date of issuance and shall be implemented
within 30 days.
Effective date: April 20, 2004.
Amendment Nos.: 171 and 133.
Facility Operating License Nos. NPF-39 and NPF-85. The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: August 19, 2003 (68 FR
49817).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 20, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: January 30, 2003.
Brief description of amendment: By letter dated January 30, 2003,
FirstEnergy Nuclear Operating Company, (FENOC), the licensee for Perry
Nuclear Power Plant (PNPP), Unit 1, submitted a request for Nuclear
Regulatory Commission review and approval of a license amendment to
modify the basis for their compliance with the requirements of Appendix
H to Title 10 Part 50 of the Code of Federal Regulations (Appendix H to
10 CFR Part 50), ``Reactor Vessel Material Surveillance Program
Requirements.'' In the license amendment submittal, FENOC requested
that they be approved to implement the Boiling Water Reactor Vessel and
Internals Project reactor pressure vessel integrated surveillance
program as the basis for demonstrating the compliance of PNPP, Unit 1,
with the requirements of Appendix H to 10 CFR Part 50.
Date of issuance: April 15, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 128.
Facility Operating License No. NPF-58: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 6, 2004 (69 FR
696).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 15, 2004.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-440, Perry Nuclear
Power Plant, Unit 1, Lake County, Ohio
Date of application for amendment: January 14, 2003.
[[Page 26196]]
Brief description of amendment: By letter dated January 14, 2003,
FirstEnergy Nuclear Operating Company, the licensee for Perry Nuclear
Power Plant, Unit 1, submitted a request for Nuclear Regulatory
Commission review and approval of a license amendment to modify the
Technical Specifications (TS) 5.1.1, 5.4.1, and 5.5.1 to replace the
requirement for the plant manager to approve administrative procedures
and the Offsite Dose Calculation Manual. The plant manager approval
signature will be replaced with the signature of a procedurally
authorized individual who would be the more appropriate authority for
approval of the activity. Additionally, a change is proposed to revise
License Condition 2.F, to replace the 30-day reporting period with a
direct reference to the 10 CFR 50.73 subsection that contains the
reporting period. The License Condition already references 10 CFR 50.73
for use in reporting plant issues.
Date of issuance: April 23, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 129.
Facility Operating License No. NPF-58: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: April 1, 2003 (68 FR
15761).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 23, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-282 and 50-306, Prairie
Island Nuclear Generating Plant, Units 1 and 2, Goodhue County,
Minnesota
Date of application for amendments: March 25, 2003, as supplemented
by your letters dated June 16, 2003, January 14, February 23, and April
7, 2004.
Brief description of amendments: The amendments revise the
technical specifications (TSs) to include implementation of relaxed
axial offset control of the reactor core through changes in TS 3.2.1
and TS 3.2.3; relocation of selected operating parameters from TS 2.0,
TS 3.1.8 and TS 3.3.1 to the Core Operating Limit Report (COLR) and the
revised pressurizer pressure-low allowable value in TS Table 3.3.1-1.
The TS changes also include, in TS 5.6.5, the topical reports
documenting the Nuclear Regulatory Commission-approved methodologies
that are used to support COLR implementation.
Date of issuance: April 28, 2004.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment Nos.: 162 and 153.
Facility Operating License Nos. DPR-42 and DPR-60: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 29, 2003 (68 FR
22750).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 28, 2004.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Units 1 and 2, San Luis Obispo County,
California
Date of application for amendment: May 29, 2003, as supplemented by
letters dated November 5, 2003 and December 23, 2003.
Brief description of amendments: The amendment revises Technical
Specification 3.8.1, ``AC Sources-Operating,'' to extend the completion
times for the required actions associated with restoration of an
inoperable diesel generator (DG). Specifically, the changes extend the
completion times for restoring an inoperable DG from 7 days to 14 days.
Date of issuance: April 20, 2004.
Effective date: April 20, 2004, and shall be implemented within 180
days of the date of issuance.
Amendment No.: Unit 1-166; Unit 2-167.
Facility Operating License Nos. DPR-80 and DPR-82: The amendment
revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37581).
The supplemental letters dated November 5, 2003 and December 23,
2003, provided additional clarifying information, did not expand the
scope of the application as originally noticed, and did not change the
NRC staff's original proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 20, 2004.
No significant hazards consideration comments received: No.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of application for amendments: February 28, 2003, as
supplemented by letters dated October 30, 2003, December 2, 2003, and
January 23, 2004.
Brief description of amendments: The amendments revise the Diablo
Canyon Power Plant Technical Specifications (TS) to add a surveillance
requirement to the Power Range Neutron Flux Rate--High Positive Rate
Trip function.
Date of issuance: April 22, 2004.
Effective date: April 22, 2004, and shall be implemented within 180
days from the date of issuance.
Amendment Nos.: Unit 1--167; Unit 2--168.
Facility Operating License Nos. DPR-80 and DPR-82: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18283).
The October 30, 2003, December 2, 2003, and January 23, 2004,
supplemental letters provided additional clarifying information, did
not expand the scope of the application as originally noticed, and did
not change the NRC staff's original proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated April 22, 2004
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of application for amendment: June 5, 2003.
Brief description of amendment: Revise the required actions in
Technical Specification (TS) 3.6.1.9 when a containment purge or
exhaust isolation valve is found inoperable as a result of leakage in
excess of the limit. The changes allow alternate methods to ensure flow
path isolation to the environment consistent with the methods allowed
for containment isolation valves in TS 3.6.3, ``Containment Isolation
Valves.''
Date of issuance: April 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 45 days of issuance.
Amendment Nos.: 290 & 280.
Facility Operating License No. DPR-77: Amendment revises the TSs.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40719).
[[Page 26197]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 21, 2004.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: August 22, 2003, as supplemented
March 19, 2004.
Brief description of amendment: The amendment revises Technical
Specification 3.3.1, ``Reactor Trip System Instrumentation.'' The
revision adds a Surveillance Requirement for response time to the
Source Range Neutron Flux Reactor Trip function.
Date of issuance: April 19, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: 52.
Facility Operating License No. NPF-90: Amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54753). The supplemental letter provided clarifying information that
was within the scope of the initial notice and did not change the
initial proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated April 19, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 30th day of April 2004.
For the Nuclear Regulatory Commisison.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 04-10305 Filed 5-10-04; 8:45 am]
BILLING CODE 7590-01-P