[Federal Register Volume 70, Number 11 (Tuesday, January 18, 2005)]
[Notices]
[Pages 2886-2907]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 05-779]
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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 December 23, 2004, through January 5, 2005.
The last biweekly notice was published on January 4, 2005 (70 FR 398).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not (1)
involve a significant increase in the probability or consequences of an
accident previously evaluated; or (2) create the possibility of a new
or different kind of accident from any accident previously evaluated;
or (3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination. Within 60 days after the date of publication of this
notice, the licensee may file a request for a hearing with respect to
issuance of the amendment to the subject facility operating license and
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
O1F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed 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
[[Page 2887]]
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 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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, New London County, Connecticut
Date of amendment request: July 15, 2004, supplemented by letter
dated August 23, 2004.
Description of amendment request: The amendment would revise
Operating License DPR-65 to address the resolution of a non-
conservative Technical Specification (TS) associated with control room
isolation radiation monitoring instrumentation. Specifically, the
amendment would revise the TS to require two operable channels of
control room isolation radiation monitoring instrumentation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration. The NRC staff has reviewed the licensee's analysis
against the standards of 10 CFR 50.92(c). The NRC staff's review is
presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change involves requirements to maintain two
operable channels in order to add a level of detection capability
and greater assurance that the safety function for control room
isolation is met. In addition, the proposed change will not alter
the setpoint value for the radiation monitors nor will it affect the
method for control room air filtration during the emergency mode of
operation. Therefore, the proposed change from one operable channel
to two operable channels for the control room isolation radiation
monitoring instrumentation will not increase the probability of
consequences of any previously evaluated accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The proposed change involves radiation monitoring channels
designed to send a signal to isolate the control room when high
radiation levels are detected to limit the radiological dose to the
control room operators in the event of an accident. In addition, the
proposed change will not have an impact on the setpoint value to
change the radiation level at which control room isolation is
assumed to occur. Again, the proposed change will not introduce
failure modes, accident initiators, or malfunctions. Therefore, the
proposed change from one operable channel to two operable channels
for the control room isolation radiation monitoring instrumentation,
will not create the possibility of a new or different kind of
accident from any accident previously evaluated.
3. Involve a significant reduction in a margin of safety.
Increasing the number of radiation monitoring channels for the
control room isolation radiation monitoring instrumentation will not
reduce a margin of safety. The proposed change to add requirements
to the TS for a redundant radiation monitoring channel will increase
the reliability of the system to perform its intended function. In
addition, the proposed change will add appropriate compensatory
actions for conditions when both channels are not available.
Therefore, given that the proposed change will continue to meet the
current design basis, any reduction in a margin of safety would not
be significant.
Based on the NRC staff's analysis, it appears that the three
standards of 10 CFR 50.929(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: Darrell J. Roberts.
[[Page 2888]]
Dominion Nuclear Connecticut, Inc., Docket Nos. 50-245, 50-336, and 50-
423, Millstone Nuclear Power Station, Unit Nos. 1, 2, and 3, New London
County, Connecticut
Date of amendment request: September 8, 2002.
Description of amendment request: The proposed amendment would
modify the Technical Specifications to support the implementation of
the proposed Dominion Nuclear Facility Quality Assurance Program
(Topical Report DOM-QA-1). Implementation of this Topical Report would
create a common quality assurance program for all sites owned by
Dominion Nuclear Connecticut, Inc. Review of this proposed amendment
was requested to be done in concert with review of the Topical Report.
The Topical Report is available in the Agencywide Document Access and
Management System under accession number ML042470015.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed changes do not involve a significant increase in
the probability or consequence of an accident previously analyzed.
The changes involve the transfer of requirements from the
administrative section of the Technical Specifications to the
Consolidated Quality Assurance Program and other licensee controlled
documents. Therefore, the proposed changes are administrative in
nature, and have no effect on a design basis accident, and will not
increase the probability or consequences of any previously analyzed
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The implementation of the proposed changes does not create the
possibility of an accident of a different type than was previously
evaluated in the Updated Final Safety Analysis Report (UFSAR). The
transfer of requirements concerning facility staff qualifications
from the administrative section of the Technical Specifications to
the Consolidated Quality Assurance Program and other licensee
controlled documents can not initiate a new or different kind of
accident.
These changes do not alter the nature of events postulated in
the UFSAR nor do they introduce any unique precursor mechanisms.
Therefore, the proposed changes are administrative in nature and do
not create the possibility of a new or different kind of accident
from those previously analyzed.
3. Involve a significant reduction in a margin of safety.
The implementation of the proposed changes does not reduce the
margin of safety. The proposed changes to transfer certain
requirements from the administration section of the Technical
Specifications to the Consolidated Quality Assurance Program and
other licensee controlled documents have no effect on design bases
radiological events. It is thus concluded that the proposed changes
are administrative in nature and the margin of safety will not be
reduced by the implementation of the changes.
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: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Section Chief: Darrell J. Roberts.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: December 6, 2004.
Description of amendment request: The proposed amendment would make
administrative changes to the Technical Specifications (TSs) including
correction of references and deleting obsolete or redundant TS
requirements and surveillances.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes are administrative or editorial in nature
and do not involve any physical changes to the plant. The changes do
not revise the methods of plant operation which could increase the
probability or consequences of accidents. No new modes of operation
are introduced by the proposed changes such that a previously
evaluated accident is more likely to occur or more adverse
consequences would result.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
These changes are administrative or editorial in nature and do
not affect the operation of any systems or equipment, nor do they
involve any potential initiating events that would create any new or
different kind of accident. There are no changes to the design
assumptions, conditions, configuration of the facility, or manner in
which the plant is operated and maintained. The changes do not
affect assumptions contained in plant safety analyses or the
physical design and/or modes of plant operation. Consequently, no
new failure mode is introduced.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
There are no changes being made to the Technical Specification
(TS) safety limits or safety system settings. The operating limits
and functional capabilities of systems, structures and components
are unchanged as a result of these administrative and editorial
changes. These changes do not affect any equipment involved in
potential initiating events or plant response to accidents. There is
no change to the basis for any TS that is related to the
establishment, or maintenance of, a nuclear safety margin.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: Allen G. Howe.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: December 7, 2004.
Description of amendment request: The proposed amendment would
revise the Technical Specifications (TSs) to: (1) Delete the
surveillance requirement (SR) associated with testing of the standby
liquid control (SLC) pump discharge pressure relief valves; and (2)
remove details from the SR for testing of the recirculation pump
discharge valves.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance
[[Page 2889]]
with the proposed amendment will not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed amendment removes details of SLC pressure relief
valve and recirculation pump discharge valve testing requirements
from the TS. Following implementation of the proposed change, the VY
TS will still require operability testing of the subject components
by reference to the VY IST [Inservice Testing] Program. Details of
SLC pressure relief valve and recirculation pump discharge valve
testing requirements will still be contained in the VY IST Program.
The SLC pressure relief valve and recirculation pump discharge valve
setpoint values related to the safety functions of those systems
will continue to be contained in the VY UFSAR [Updated Final Safety
Analysis Report]. Changes to the VY UFSAR are evaluated per the
requirements of 10 CFR 50.59. These controls are adequate to ensure
the required inservice testing is performed to verify the components
are operable and capable of performing their respective safety
functions. The proposed amendment introduces no new equipment or
changes to how equipment is operated. Neither the SLC pressure
relief valves nor the recirculation pump discharge valves are
initiators of any analyzed accidents. Therefore, operation of VY in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed amendment removes details of SLC pressure relief
valve and recirculation pump discharge valve testing requirements
from the TS. The proposed amendment does not change the design or
function of any component or system. No new modes of failure or
initiating events are being introduced. Therefore, operation of VY
in accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station (VY) in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The proposed amendment removes details of SLC pressure relief
valve and recirculation pump discharge valve testing requirements
from the TS. The proposed amendment does not change the design or
function of any component or system. The proposed amendment does not
involve any safety limits or limiting safety system settings.
Since the proposed controls are adequate to ensure the required
inservice testing is performed, there will still be high assurance
that the components are operable and capable of performing their
respective safety functions, and that the systems will respond as
designed to mitigate the subject events. Therefore, operation of VY
in accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW, Washington, DC 20037-1128.
NRC Section Chief: Allen G. Howe.
Entergy Nuclear Vermont Yankee, LLC and Entergy Nuclear Operations,
Inc., Docket No. 50-271, Vermont Yankee Nuclear Power Station, Vernon,
Vermont
Date of amendment request: December 15, 2004.
Description of amendment request: The proposed amendment would
revise the limiting conditions for operation in Technical Specification
(TS) 3.3 and the surveillance requirements in TS 4.3 associated with
the control rod system. Specifically, the proposed changes would revise
the TSs associated with: (1) Control rod operability; (2) control rod
scram time testing; and (3) control rod accumulator operability.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
The proposed changes do not significantly affect the design or
fundamental operation and maintenance of the plant. Accident
initiators or the frequency of analyzed accident events are not
significantly affected as a result of the proposed changes;
therefore, there will be no significant change to the probabilities
of accidents previously evaluated.
The proposed changes do not significantly alter assumptions or
initial conditions relative to the mitigation of an accident
previously evaluated. The proposed changes continue to ensure
process variables, structures, systems, and components (SSCs) are
maintained consistent with the safety analyses and licensing basis.
The revised technical specifications continue to require that SSCs
are properly maintained to ensure operability and performance of
safety functions as assumed in the safety analyses. The design basis
events analyzed in the safety analyses will not change significantly
as a result of the proposed changes to the TS.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
The proposed changes do not involve any physical alteration of
the plant (no new or different type of equipment being installed)
and do not involve a change in the design, normal configuration or
basic operation of the plant. The proposed changes do not introduce
any new accident initiators. In some cases, the proposed changes
impose different requirements; however, these new requirements are
consistent with the assumptions in the safety analyses and current
licensing basis. Where requirements are relocated to other licensee-
controlled documents, adequate controls exist to ensure their proper
maintenance.
The proposed changes do not involve significant changes in the
fundamental methods governing normal plant operation and do not
require unusual or uncommon operator actions. The proposed changes
provide assurance that the plant will not be operated in a mode or
condition that violates the essential assumptions or initial
conditions in the safety analyses and that SSCs remain capable of
performing their intended safety functions as assumed in the same
analyses. Consequently, the response of the plant and the plant
operator to postulated events will not be significantly different.
Therefore, the proposed TS change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The operation of Vermont Yankee Nuclear Power Station in
accordance with the proposed amendment will not involve a
significant reduction in a margin of safety.
Margin of safety is related to the confidence in the ability of
the fission product barriers to perform their design functions
during and following an accident situation. The proposed changes do
not significantly affect any of the assumptions, initial conditions
or inputs to the safety analyses. Plant design is unaffected by
these proposed changes and will continue to provide adequate
defense-in-depth and diversity of safety functions as assumed in the
safety analyses.
There are no proposed changes to any of the Safety Limits or
Limiting Safety System Setting requirements. The proposed changes
maintain requirements consistent with safety analyses assumptions
and the licensing basis. Fission product barriers will continue to
meet their design capabilities without any significant impact to
their ability to maintain parameters within acceptable limits. The
safety functions are maintained within acceptable limits without any
significant decrease in capability.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three
[[Page 2890]]
standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff
proposes to determine that the amendment request involves no
significant hazards consideration.
Attorney for licensee: Mr. David R. Lewis, Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037-1128.
NRC Section Chief: Allen G. Howe.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: December 20, 2004.
Description of amendment request: The requested change will delete
the requirements in Technical Specification (TS) 5.6.1, ``Occupational
Radiation Exposure Report,'' and TS 5.6.4, ``Monthly Operating
Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 20, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Michael A. Webb (Acting).
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: December 20, 2004.
Description of amendment request: The requested change will delete
the requirements in Technical Specification (TS) 6.6.1, ``Occupational
Radiation Exposure Report,'' and TS 6.6.4, ``Monthly Operating
Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 20, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Michael A. Webb (Acting).
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: October 14, 2004.
Brief description of amendments: The proposed change will revise
the surveillance requirement (SR) 3.6.6.8 frequency of every 10 years.
Instead, the proposed change to SR 3.6.6.8 will require verification
that spray nozzles are unobstructed following maintenance that could
result in nozzle blockage.
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 and states that the amendment
request:
1. Does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed change modifies the [Surveillance Requirements] SR
to verify that the [Reactor Building] RB spray nozzles are
unobstructed after maintenance that could introduce material that
could result in nozzle blockage. The spray nozzles are not assumed
to be initiators of any previously analyzed
[[Page 2891]]
accident. Therefore, the change does not increase the probability of
any accident previously evaluated. The spray nozzles are assumed in
the accident analyses to mitigate design basis accidents. The
revised SR to verify system OPERABILITY following maintenance is
considered adequate to ensure OPERABILITY of the RB spray system.
Since the system will still be able to perform its accident
mitigation function, the consequences of accidents previously
evaluated are not increased. Therefore, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does not create the possibility of a new or different type of
accident from any accident previously evaluated.
The proposed change revises the SR to verify that the RB spray
nozzles are unobstructed after maintenance that could result in
nozzle blockage. The change does not introduce a new mode of plant
operation and does not involve physical modification to the plant.
The change will not introduce new accident initiators or impact the
assumptions made in the safety analysis. Therefore, the proposed
change does not create the possibility of a new or different kind of
accident from any previously evaluated.
3. Does not involve a significant reduction in the margin of
safety.
The proposed change revises the frequency for performance of the
SR to verify that the RB spray nozzles are unobstructed. The
frequency is changed from every 10 years to following maintenance
that could result in nozzle blockage. This requirement, along with
foreign material exclusion programs and the remote physical location
of the spray nozzles, provides assurance that the spray nozzles will
remain unobstructed. As the spray nozzles are expected to remain
unobstructed and able to perform their post-accident mitigation
function, plant safety is not significantly affected. 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: Steven R. Carr, Associate General Counsel--
Legal Department, Progress Energy Service Company, LLC, Post Office Box
1551, Raleigh, North Carolina 27602.
NRC Section Chief: Michael L. Marshall, Jr.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: November 22, 2004.
Description of amendment request: The requested change will delete
the requirements in Technical Specification (TS) 5.6.1, ``Occupational
Radiation Exposure Report,'' and TS 5.6.4, ``Monthly Operating
Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated November 22, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating report
of shutdown experience and operating statistics if the equivalent
data is submitted using an industry electronic database. It also
eliminates the TS reporting requirement for an annual occupational
radiation exposure report, which provides information beyond that
specified in NRC regulations. The proposed change involves no
changes to plant systems or accident analyses. As such, the change
is administrative in nature and does not affect initiators of
analyzed events or assumed mitigation of accidents or transients.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Mr. John R. McPhail, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Section Chief: Michael K. Webb (Acting).
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: April 29, 2004, as supplemented November
23, 2004.
Description of amendment request: The proposed amendment is a
selective-scope application of an alternative source term (AST) for the
fuel handling accident (FHA) in accordance with Title 10 of the Code of
Federal Regulations (10 CFR) Section 50.67, ``Accident Source Term.''
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 involves implementation of the AST for
the fuel handling accident at MNGP [Monticello Nuclear Generating
Plant]. There are no physical design modifications to the plant
associated with the proposed amendment. The revised calculations do
not impact the initiators of an FHA in any way.
The changes also do not impact the initiators for any other
design basis accident (DBA) or events. Therefore, because DBA
initiators are not being altered by adoption of the AST analyses,
the probability of an accident previously evaluated is not affected.
With respect to consequences, the only previously evaluated
accident that could be affected is the FHA. The AST is an input to
calculations used to evaluate the consequences of the accident, and
does not, in and of itself, affect the plant response or the actual
pathways to the environment utilized by the radiation/activity
released by the fuel. It does however, better represent the physical
characteristics of the release, so that appropriate mitigation
techniques may be applied. For the FHA, the AST analyses demonstrate
acceptable doses that are within regulatory limits after 24 hours of
radiological decay, without credit for Secondary Containment
integrity, selected ESF [engineered safety feature] filtration
system operation (i.e., SBGT [standby gas treatment] System or
Control Room EFT [emergency filtration] System) or Control Room
isolation. Therefore, the consequences of an accident previously
evaluated are not significantly increased.
Based on the above conclusions, this proposed amendment does not
involve a
[[Page 2892]]
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 amendment does not involve a physical alteration of
the plant. No new or different types of equipment will be installed
and there are no physical modifications to existing equipment
associated with the proposed changes. Also, no changes are proposed
to the methods governing plant/system operation during handling of
irradiated fuel, so no new initiators or precursors of a new or
different kind of accident are created. New equipment or personnel
failure modes that might initiate a new type of accident are not
created as a result of the proposed amendment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Does the proposed change involve a significant reduction in
the margin of safety?
Response: No.
The proposed amendment is associated with the implementation of
a new licensing basis for the MNGP FHA. Approval of this change from
the original source term to an alternative source term derived in
accordance with the guidance of RG 1.183 [``Alternative Radiological
Source Terms for Evaluating Design Basis Accidents at Nuclear Power
Reactors''] is being requested. The results of the FHA accident
analysis, revised in support of the proposed license amendment, are
subject to revised acceptance criteria. The AST FHA analysis has
been performed using conservative methodologies, as specified in RG
1.183. Safety margins have been evaluated and analytical
conservatism has been utilized to ensure that the analyses
adequately bound the postulated limiting event scenario. The dose
consequences of the limiting FHA remain within the acceptance
criteria presented in 10 CFR 50.67 and RG 1.183.
The proposed changes continue to ensure that the doses at the
Exclusion Area Boundary (EAB) and Low Population Zone (LPZ)
boundaries, as well as the Control Room, are within the
corresponding regulatory limits. For the FHA, RG 1.183
conservatively sets the EAB and LPZ limits below the 10 CFR 50.67
limit, and sets the Control Room limit consistent with 10 CFR 50.67.
Since the proposed amendment continues to ensure the doses at
the EAB, LPZ and Control Room are within corresponding regulatory
limits, the proposed license amendment does not involve a
significant reduction in a margin of safety.
Based on the above, NMC has determined that operation of the
facility in accordance with the proposed change does not involve a
significant hazards consideration as defined in 10 CFR 50.92(c), in
that it: (1) Does not involve a significant increase in the
probability or consequences of an accident previously evaluated; (2)
does not create the possibility of a new or different kind of
accident from any accident previously evaluated; and (3) does not
involve a significant reduction in a margin of safety.
The U. S. Nuclear Regulatory Commission (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: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: June 30, 2004, as supplemented November
5, 2004.
Description of amendment request: The proposed amendment would
revise the technical specifications (TSs) to implement a 24-month fuel
cycle.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) 50.91(a), the licensee has provided its analysis
of the issue of no significant hazards consideration (NSHC), which is
presented below:
1. The proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
a. Surveillance Testing Interval Extensions
The proposed Technical Specification (TS) changes involve
changes in the surveillance testing to facilitate a change in the
operating cycle from 18 months to 24 months. The proposed TS changes
do not physically impact the normal operation of the plant, nor do
they impact any design or functional requirements of the associated
systems. That is, the proposed TS changes neither impact the TS SRs
[surveillance requirements] themselves nor the manner in which the
surveillances are performed.
In addition, the proposed TS changes do not introduce any
accident initiators, since no accidents previously evaluated relate
to the frequency of surveillance testing. Also, evaluations of the
proposed TS changes demonstrate that the availability of equipment
and systems required to prevent or mitigate the radiological
consequences of an accident are not significantly affected because
of other, more frequent testing that is performed, the availability
of redundant systems and equipment, or the high reliability of the
equipment. Since the impact on the systems is minimal NMC [Nuclear
Management Company] has concluded that the overall impact on the
plant safety analysis is negligible.
A historical review of surveillance test results and associated
maintenance records indicated that there was no evidence of any
failure that would invalidate the above conclusions.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of an accident previously evaluated.
b. TS Trip Setting Changes
Changes are proposed to the Monticello TS Trip Settings. The
proposed changes are a result of application of the Monticello
Instrument Setpoint Methodology using plant-specific drift values.
Application of this methodology results in Trip Setpoints that more
accurately reflect total instrumentation loop accuracy, as well as
that of test equipment and calculated drift between surveillances.
The proposed changes will not result in hardware changes. The
instrumentation is not assumed to be initiators of any analyzed
events, nor do they impact any design or functional requirements of
the associated systems. Existing operating margins between plant
conditions and actual plant setpoints are not significantly reduced
due to the proposed changes. The role of the instrumentation is in
mitigating and thereby, limiting the consequences of accidents.
The Nominal Trip Setpoints were developed to ensure the design
and safety analysis limits are satisfied. The methodology used for
the development of the Trip Settings ensures: (1) The affected
instrumentation remains capable of mitigating design basis events as
described in the safety analysis; and, (2) the results and
radiological consequences described in the safety analysis remain
bounding. The proposed changes do not alter the plant's ability to
detect and mitigate events.
Therefore, these changes do not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
c. Surveillance Testing Interval Reductions
The proposed TS changes involve reductions in the surveillance
testing intervals from once per operating cycle or refueling outage
to once every three (3) months or once per quarter for the equipment
associated with these TS SRs. The shorter intervals are based upon
the plant-specific results of a review of the surveillance test
history for this equipment. The implementing procedures for these
SRs have been performed on a once per three (3) month or once per
quarter interval for a number of years, and these changes more
accurately reflect actual plant maintenance practices. The proposed,
more restrictive TS changes do not physically impact the plant, nor
do they impact any design or functional requirements of the
associated systems. That is, the proposed TS changes neither degrade
the performance of, nor increase the challenges to, any safety
system assumed to function in the safety analysis. These proposed TS
changes neither impact the TS SRs themselves nor the manner in which
the surveillances are performed.
The proposed TS changes do not introduce any accident
initiators, since no accident previously evaluated relate to the
frequency
[[Page 2893]]
of surveillance testing. The proposed TS intervals demonstrate that
the equipment and systems required to prevent or mitigate the
radiological consequences of an accident are continuing to meet the
assumptions of the setpoint evaluation on a more frequent basis.
Since the impacts on systems are minimal and the assumptions of the
safety analyses are maintained, NMC has concluded that the overall
impact on the plant safety analysis is negligible.
Therefore, the proposed TS changes do not significantly increase
the probability or consequences of any accident previously
evaluated.
2. The proposed amendment does not create the possibility of a
new or different kind or accident from any accident previously
evaluated.
a. Surveillance Testing Interval Extensions
The proposed TS changes involve changes in the surveillance
testing intervals to facilitate a change in the operating cycle
length. The proposed TS changes do not introduce any failure
mechanisms of a different type than those previously evaluated.
There are no physical changes being made to the facility. No new or
different equipment is being installed. No installed equipment is
being operated in a different manner. As a result no new failure
modes are introduced. The SRs themselves, and the manner in which
surveillance tests are performed, remain unchanged.
A historical review of surveillance test results and associated
maintenance records indicated that there was no evidence of any
failure that would invalidate the above conclusions.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
b. TS Trip Setting Changes
The proposed changes to the Trip Settings are a result of
applying the Monticello Instrument Setpoint Methodology using plant-
specific drift values. The application of this methodology does not
create the possibility of any new or different kinds of accidents
from any accidents previously evaluated. This is based upon the fact
that the method and manner of plant operations are unchanged.
The use of the proposed Trip Setpoints does not impact the safe
operation of the plant in that the safety analysis limits are
maintained. The proposed changes in Trip Settings involve no system
additions or physical modifications to plant systems. The Trip
Settings are revised to ensure the affected instrumentation remains
capable of mitigating accidents and transients. Plant equipment will
not be operated in a manner different from previous operation. Since
operational methods remain unchanged and the operating parameters
were evaluated to maintain the plant within existing design basis
criteria no different type of failure or accident is created.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
c. Surveillance Testing Interval Reductions
The proposed TS changes involve reductions in the surveillance
testing intervals from once per operating cycle or refueling outage
to once every three (3) months or once per quarter for the equipment
associated with these TS SRs. The shorter intervals are based upon
the plant-specific results of a review of the surveillance test
history for this equipment. The implementing procedures for these
SRs have been performed on a once per three (3) month or once per
quarter interval for a number of years and these changes more
accurately reflect actual plant maintenance practices. The proposed
more restrictive TS changes do not physically impact the plant, nor
do they impact any design or functional requirements of the
associated systems. That is, the proposed TS changes neither degrade
the performance of, nor increase the challenges to, any safety
system assumed to function in the safety analysis. These proposed TS
changes neither impact the TS SRs themselves nor the manner in which
the surveillances are performed.
The proposed TS changes do not introduce any failure mechanism
of a different type than those previously evaluated. The proposed
changes make no physical changes to the plant. No new or different
equipment is being installed. No installed equipment is being
operated in a different manner.
A historical review of surveillance test results and associated
maintenance records indicate that there is no evidence of any
failure that would invalidate the above conclusions.
Therefore, the proposed TS changes do not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. The proposed amendment will not involve a significant
reduction in a margin of safety.
a. Surveillance Testing Interval Extensions
Although the proposed TS changes result in changes in the
interval between surveillance tests, the impact, if any, on system
availability is minimal based upon other, more frequent testing that
is performed, the existence of redundant systems and equipment or
overall system reliability. Evaluations show there is no evidence of
any time-dependant failure that would impact system availability.
The proposed changes do not significantly impact the condition
or performance of structures, systems and components relied upon for
accident mitigation. The proposed TS changes do not physically
impact the plant, nor do they impact any design or functional
requirements of the associated systems. The proposed changes do not
significantly impact any safety analysis assumptions or results.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
b. TS Trip Setting Changes
The proposed changes do not involve a reduction in a margin of
safety. The proposed changes were developed using a Monticello
Instrument Setpoint Methodology using plant-specific drift values.
This methodology ensures no safety analysis limits are exceeded. The
proposed TS changes do not physically impact the plant, nor do they
impact any design or functional requirements of the associated
systems.
As such, these proposed changes do not involve a reduction in a
margin of safety.
c. Surveillance Testing Interval Reductions
The proposed TS changes result in a shorter interval between
surveillance tests to ensure the assumptions of the safety analysis
are maintained. The impact, if any, on system availability is
minimal, as a result of the more frequent testing that is performed.
The proposed changes do not significantly impact the condition or
performance of structures, systems and components relied upon for
accident mitigation. The proposed TS changes do not physically
impact the plant, nor do they impact any design or functional
requirements of the associated systems. The proposed changes do not
significantly impact any safety analysis assumptions or results.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The U. S. Nuclear Regulatory Commission (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 NSHC.
Attorney for licensee: Jonathan Rogoff, Esquire, Vice President,
Counsel & Secretary, Nuclear Management Company, LLC, 700 First Street,
Hudson, WI 54016.
NRC Section Chief: L. Raghavan.
Omaha Public Power District, Docket No. 50-285, Fort Calhoun Station,
Unit No. 1, Washington County, Nebraska
Date of amendment request: September 8, 2004.
Description of amendment request: The proposed amendment deletes
the requirements from the technical specifications (TS) to maintain
containment hydrogen 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 TS for nuclear
power reactors currently licensed to operate. The revised 10 CFR 50.44,
``Standards for Combustible Gas Control System in
[[Page 2894]]
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
relevant portions of the model NSHC determination (TS for Fort Calhoun
do not include requirements for hydrogen recombiners) in its
application dated September 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 revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (EOPs), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
Based upon the reasoning presented above, the requested change does
not involve a significant hazards consideration.
Attorney for licensee: James R. Curtiss, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: November 1, 2004.
Description of amendment requests: The requested change will delete
Technical Specification (TS) 5.6.1, ``Occupational Radiation Exposure
Report,'' and TS 5.6.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated November 1, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients.
[[Page 2895]]
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve a significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Robert A. Gramm.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: September 22, 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, Sec. 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 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 September 22, 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: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: December 10, 2004.
Description of amendment requests: The proposed amendment will
delete the requirements from the Technical Specifications (TS) to
maintain hydrogen recombiners and hydrogen 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 TS for nuclear power reactors currently licensed to operate.
[[Page 2896]]
The revised Sec. 50.44 of Title 10 of the Code of Federal Regulations
(10 CFR), ``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 proposed license amendment will revise TS 3.3.11, ``Post
Accident Monitoring Instrumentation (PAMI),'' to delete the Note in
Condition C. Also in TS 3.3.11, Condition D will be deleted. In TS
Table 3.3.11-1, Item 10, ``Containment Hydrogen Monitors,'' is deleted.
Other TS changes included in this application are limited to
renumbering and formatting changes that resulted directly from the
deletion of the above requirements related to hydrogen monitors. The
changes to TS requirements result in changes to various TS Bases
sections. The TS Bases changes will be submitted with a future update
in accordance with TS 5.4.4, ``Technical Specifications (TS) Bases
Control.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) 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 December 10, 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 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 monitors no longer meet the definition
of Category 1 in RG 1.97. As part of the rulemaking to revise 10 CFR
50.44 the Commission found that Category 3, as defined in RG 1.97,
is an appropriate categorization for the hydrogen monitors because
the monitors are required to diagnose the course of beyond design-
basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the SAMGs [severe accident
management guidelines], the emergency plan (EP), the emergency
operating procedures (EOP), and site survey monitoring that support
modification of emergency plan protective action recommendations
(PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction in
the margin of safety. Removal of hydrogen monitoring from TS will not
result in a significant reduction in their functionality, reliability,
and availability.
Based upon the reasoning presented above and the previous
discussion of the amendment request, the requested change does not
involve a significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Robert A. Gramm.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of amendment requests: December 17, 2004.
Description of amendment requests: The proposed amendments would
revise Technical Specification (TS) 3.8.1, ``AC Sources--Operating,''
TS 3.8.4, ``DC Sources--Operating,'' TS 3.8.5, ``DC Sources--
Shutdown,'' TS 3.8.6, ``Battery Cell Parameters,'' TS 3.8.7,
``Inverters--Operating,'' and TS 3.8.9, ``Distribution Systems--
Operating.'' This change will also add a new Battery Monitoring and
Maintenance Program, section 5.5.2.16. The proposed change will provide
operational flexibility to credit DC electrical subsystem design
upgrades that are in progress. These upgrades will provide increased
capacity batteries, additional battery chargers, and the
[[Page 2897]]
means to cross-connect DC subsystems while meeting all design battery
loading requirements. With these modifications in place, it will be
feasible to perform routine surveillance as well as battery
replacements online.
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. Will operation of the facility in accordance with this
proposed change involve a significant increase in the probability or
consequences of an accident previously evaluated?
Response: No.
The proposed changes to Technical Specifications (TS) 3.8.4 and
3.8.6 would allow extension of the Completion Time (CT) for
inoperable Direct Current (DC) distribution subsystems to manually
cross-connect DC distribution buses of the same safety train of the
operating unit for a period of 30 days. Currently the CT only allows
for 2 hours to ascertain the source of the problem before a
controlled shutdown is initiated. Loss of a DC subsystem is not an
initiator of an event. However, complete loss of a Train A
(subsystems A and C) or Train B (subsystems B and D) DC system would
initiate a plant transient/plant trip.
Operation of a DC Train in cross-connected configuration does
not affect the quality of DC control and motive power to any system.
Therefore, allowing the cross-connect of DC distribution systems
does not significantly increase the probability of an accident
previously evaluated in Chapter 15 of the Updated Final Safety
Analysis Report (UFSAR).
The above conclusion is supported by Probabilistic Risk Analysis
(PRA) evaluation which encompasses all accidents, including UFSAR
Chapter 15.
Modification to the Frequency for Surveillance Requirement (SR)
3.8.6.1 is consistent with the recommendations of TSTF 360 Rev. 1
and IEEE 450-2002, and similarly does not impact safety
considerations.
Further changes are made of an editorial nature or provide
clarification only. For example, discussions regarding electrical
`Trains' and `Subsystems' will be in more conventional terminology.
Limiting Condition for Operations (LCOs) affected by editorial
changes include 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9.
Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have been
incorporated into LCOs 3.8.4 and 3.8.6. TSTF-360, Rev. 1 was
previously approved by the NRC, and IEEE 450-2002 includes industry-
generic recommendations.
The changes being proposed do not affect assumptions contained
in other safety analyses or the physical design of the plant other
than the upgrades of the electrical systems described in this
change, nor do they affect other Technical Specifications that
preserve safety analysis assumptions.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant increase in the
probability or consequences of an accident previously analyzed.
2. Will operation of the facility in accordance with this
proposed change create the possibility of new or different kind of
accident from any accident previously evaluated?
Response: No.
The proposed change to Technical Specifications 3.8.4 will
enable the cross-tie of subsystems. New equipment, swing battery
chargers, distribution panels, and associated protective devices are
added to increase overall DC system reliability. Both administrative
and mechanical controls will be in place to ensure the design and
operation of the DC distribution systems continue to perform to
applicable design standards. During cross connecting of subsystem
buses, two batteries would be paralleled for a short duration. An
electrical fault during that duration could exceed the interrupting
duties of the protective devices. This is standard industry practice
during transfer of power sources and is considered to be an
acceptable minimal risk. For example, the design of the 1E 4kV power
system is based on this practice as well. Therefore, the addition of
new equipment does not create the possibility of a new or different
kind of accident from any previously evaluated.
Enhancements from TSTF-360, Rev. 1 and IEEE 450-2002 have been
incorporated into LCOs 3.8.4 and 3.8.6. TSTF-360, Rev. 1 is
previously approved and IEEE 450-2002 includes industry-generic
recommendations. Enhancements, including surveillance intervals or
required completion times, will not create the possibility of a new
or different kind of accident from any previously evaluated.
LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 are revised to
incorporate editorial changes. Since these changes do not affect
plant design but enhance clarity, these modifications do not create
the possibility of a new or different kind of accident from any
previously evaluated.
Therefore, operation of the facility in accordance with this
proposed change will not create the possibility of new or different
kind of accident from any accident previously evaluated.
3. Will operation of the facility in accordance with this
proposed change involve a significant reduction in a margin of
safety?
Response: No.
The proposed change does not alter the bases for assurance that
safety-related activities are performed correctly or the basis for
any Technical Specification that is related to the establishment of
or maintenance of a safety margin. Specifically, battery sizing
calculations continue to show that new upgraded capacity batteries
will meet the most limiting load profile that includes margin for
growth, with aging and temperature correction. Battery modified
performance discharge testing will demonstrate on an on-going basis
that battery capacity will be greater than or equal to 80% of
original design requirements at all times during service life and
that the service profiles will be met as is currently required by
Surveillance Requirements 3.8.4.7 and 3.8.4.8. The addition of the
DC cross-tie capability proposed for LCO 3.8.4 will ensure
appropriate operations of the DC buses during maintenance activities
such as battery testing or replacement. Enhancements from TSTF-360,
Rev. 1 and IEEE 450-2002 have been incorporated into LCOs 3.8.4 and
3.8.6. TSTF-360, Rev. 1 is previously approved and IEEE 450-2002
includes industry-generic recommendations. Enhancements including
surveillance intervals or required completion times will not involve
a significant reduction in a margin of safety.
Also, LCOs 3.8.1, 3.8.4, 3.8.5, 3.8.6, 3.8.7, and 3.8.9 are
revised to incorporate editorial changes. Since these changes do not
affect plant design or operations but should enhance clarity, these
modifications would not involve a significant reduction in margin of
safety.
Therefore, operation of the facility in accordance with the
proposed amendment would not involve a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Douglas K. Porter, Esquire, Southern
California Edison Company, 2244 Walnut Grove Avenue, Rosemead,
California 91770.
NRC Section Chief: Robert A. Gramm.
Southern Nuclear Operating Company, Inc., et al., Docket Nos. 50-424
and 50-425, Vogtle Electric Generating Plant, Units 1 and 2, Burke
County, Georgia
Date of amendment request: October 26, 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, Sec. 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 to reflect the related
changes to LCO 3.0.4, and Surveillance Requirement 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-
[[Page 2898]]
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 October
26, 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: Mr. Arthur H. Domby, Troutman Sanders,
NationsBank Plaza, Suite 5200, 600 Peachtree Street, NE., Atlanta,
Georgia 30308-2216.
NRC Section Chief: John A. Nakoski.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant, Unit 1, Limestone County, Alabama
Date of amendment request: December 6, 2004 (TS 426).
Description of amendment request: The proposed amendment would
revise the current Unit 1 Diesel Generators (DG) Allowed Outage Time
(AOT) in the Technical Specifications (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 amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
No. The DGs are designed as backup AC power sources in the event
of loss of offsite power. The proposed DG TS AOT does not change the
conditions, operating configurations, or minimum amount of operating
equipment assumed in the safety analysis for accident mitigation. No
changes are proposed in the manner in which the DGs provide plant
protection or which create new modes of plant operation. In
addition, a PSA [probabilistic safety assessment] evaluation
concluded that the risk contribution of the DG TS AOT extension is
non-risk significant. Therefore, the proposed amendment does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
No. The proposed amendment does not introduce new equipment,
which could create a new or different kind of accident. No new
external threats, release pathways, or equipment failure modes are
created. Therefore, the implementation of the proposed amendment
will not create a possibility for an accident of a new or different
type than those previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
No. BFN's emergency AC system is designed with sufficient
redundancy such that a DG may be removed from service for
maintenance or testing. The remaining DGs are capable of carrying
sufficient electrical loads to satisfy the UFSAR [Updated Final
Safety Analysis Report] requirements for accident mitigation or unit
safe shutdown.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-259, Browns Ferry Nuclear
Plant (BFN), Unit 1, Limestone County, Alabama
Date of amendment request: December 6, 2004 (TS 428).
Description of amendment request: The proposed amendment would
revise the reactor vessel Pressure-Temperature (P-T) curves depicted in
the Technical Specification (TS) Figure 3.4.9-1 and adds a new TS
Figure 3.4.9-2.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 2899]]
No. The proposed changes deal exclusively with the reactor
vessel P-T curves, which define the permissible regions for
operation and testing. Failure of the reactor vessel is not
considered as a design basis accident. Through the design
conservatisms used to calculate the P-T curves, reactor vessel
failure has a low probability of occurrence and is not considered in
the safety analyses. The proposed changes adjust the reference
temperature for the limiting material to account for irradiation
effects and provide the same level of protection as previously
evaluated and approved.
The adjusted reference temperature calculations were performed
in accordance with the requirements of 10 CFR 50 Appendix G using
the guidance contained in Regulatory Guide 1.190, ``Calculational
and Dosimetry Methods for Determining Pressure Vessel Neutron
Fluence,'' to reflect use of the operating limits to no more than 16
Effective Full Power Years (EFPY). These changes do not alter or
prevent the operation of equipment required to mitigate any accident
analyzed in the BFN Final Safety Analysis Report.
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?
No. The proposed changes to the reactor vessel P-T curves do not
involve a modification to plant equipment. No new failure modes are
introduced. There is no effect on the function of any plant system,
and no new system interactions are introduced by this change.
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?
No. The proposed curves conform to the guidance contained in
Regulatory Guide (RG) 1.190, ``Calculational and Dosimetry Methods
for Determining Pressure Vessel Neutron Fluence,'' and maintain the
safety margins specified in 10 CFR 50 Appendix G. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority (TVA), Docket No. 50-328, Sequoyah Nuclear
Plant, Unit 2, Hamilton County, Tennessee
Date of amendment request: December 2, 2004.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 3/4.4.5, ``Steam Generators,''
including associated Bases 3/4.4.5 to change the inspection scope of
steam generator tubing in the Westinghouse Electric Company explosive
tube expansion region below the top of the tubesheet. Additionally, the
proposed TS change removes the axial primary water stress corrosion
cracking at dented tube support plate alternate repair criteria and the
associated note for the exclusion made for Unit 2 Cycle 12 operation
only and changes the current definition of plugging limit to exclude
possible indications below the W* distance.
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.
Of the various accidents previously evaluated, the proposed
changes only affect the steam generator tube rupture (SGTR) event
evaluation and the postulated steam line break (SLB) accident
evaluation. Loss-of-coolant accident (LOCA) conditions cause a
compressive axial load to act on the tube. Therefore, since the LOCA
tends to force the tube into the tubesheet rather than pull it out,
it is not a factor in this amendment request. Another faulted load
consideration is a safe shutdown earthquake (SSE); however, the
seismic analysis of Westinghouse 51 series SGs has shown that axial
loading of the tubes is negligible during an SSE.
TVA's amendment request takes credit for how the tubesheet
enhances the tube integrity in the Westinghouse Electric Company
explosive tube expansion (WEXTEX) region by precluding tube
deformation beyond its initial expanded outside diameter. For the
SGTR and SLB events, the required structural margins of the SG tubes
will be maintained due to the presence of the tubesheet. Tube
rupture is precluded for axial cracks in the WEXTEX region due to
the constraint provided by the tubesheet. Therefore, the normal
operating 3[Delta]P margin and the postulated accident 1.43[Delta]P
margin against burst are maintained.
The W* length supplies the necessary resistive force to preclude
pullout loads under both normal operating and accident conditions.
The contact pressure results from the WEXTEX expansion process,
thermal expansion mismatch between the tube and tubesheet, and from
the differential pressure between the primary and secondary side.
Therefore, the proposed change results in no significant increase in
the probability or the occurrence of an SGTR or SLB accident.
The proposed changes do not affect other systems, structures,
components or operational features. Therefore, based on the above
evaluation, the proposed changes do not involve a significant
increase in the probability of an accident previously evaluated.
The consequences of an SGTR event are primarily affected by the
primary-to-secondary flow rate and the time duration of the primary-
to-secondary flow during the event. Primary-to-secondary flow rate
through a postulated ruptured tube (i.e., complete severance of a
single SG tube) is not affected by the proposed change since the
flow rate is based on the inside diameter of a SG tube and the
pressure differential. TVA's amendment request does not change
either of these. The duration of primary-to-secondary leakage is
based on the time required for an operator to determine that a SGTR
has occurred, the time to identify and isolate the faulty SG, and
ensure termination of radioactive release to the atmosphere from the
faulty SG. TVA's amendment request does not affect the duration of
the primary-to-secondary leakage because it does not change the
control room indicators with which an operator would determine that
an SGTR has occurred. The consequences of an SGTR are secondarily
affected by primary-to-secondary leakage, which could occur due to
axial cracks remaining in service in the WEXTEX region in a non-
faulted SG. During a SGTR, the primary-to-secondary differential
pressure is less than or equal to the normal operating differential
pressure; therefore, the primary-to-secondary leakage due to axial
cracks in the WEXTEX region of a non-faulted SG during a SGTR would
be less than or equal to the primary-to-secondary leakage
experienced during normal operation. Primary-to-secondary leakage is
considered in the calculation determining the consequences of a SGTR
and the value is bounding.
The postulated SLB has the greatest primary-to-secondary
pressure differential, and therefore could experience the greatest
primary-to-secondary leakage. TVA's amendment request requires the
aggregate leakage, (i.e., the combined leakage for the tubes with
service induced degradation inside the tubesheet) plus the combined
leakage developed by other ARC [alternate repair criteria], to
remain below the maximum allowable SLB primary-to-secondary leakage
rate limit such that the doses are maintained to less than a
fraction of the 10 CFR 100 limits and also less than the general
design criteria (GDC)--19 limits.
TVA's proposed change also removes the existing axial PWSCC
[primary water stress corrosion cracking] at dented tube support
plate ARC and removes the exclusion made for Unit 2 Cycle 12
operation only from the TS. This ARC was not used on Unit 2 and was
only intended through the Unit 2 Cycle 12 operation. Therefore, this
change is inherently more conservative.
[[Page 2900]]
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.
TVA's amendment request does not introduce any physical changes
to the Sequoyah Unit 2 SGs. TVA's amendment request takes credit for
how the tubesheet enhances the SG tube integrity in the WEXTEX
region by precluding tube deformation beyond its initial expanded
outside. Removal of the existing PWSCC axial at dented tube support
plate ARC incorporates the more conservative TS limit for SG tube
plugging. A failure to meet SG tube integrity results in an SGTR.
Because degradation detected within the WEXTEX region are required
to be plugged, it is highly unlikely that a W* tube would fail as a
result of a circumferential defect. Therefore a tube severance,
which would strike neighboring tubes and create a multiple tube
rupture, is not credible.
The proposed change does not introduce any new equipment or any
change to existing equipment. No new effects on existing equipment
are created.
Based on the above evaluation, the proposed changes do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
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 amendment request maintains the structural margins of the SG
tubes for both normal and accident conditions that are required by
Regulatory Guide 1.121.
For cracking located within the tubesheet, tube burst is
precluded due to the presence of the tubesheet. WCAP-14797 defines a
length, W*, of degradation free expanded tubing that provides the
necessary resistance to tube pullout due to the pressure induced
forces (with applicable safety factor applied). Application of the
W* methodology will preclude unacceptable primary-to-secondary
leakage during all plant conditions. The methodology for determining
leakage provides for large margins between calculated and actual
leakage values in the W* criteria. TVA's proposed change to remove
PWSCC ARC from the TS does not compromise structural integrity or
leakage integrity of SG tubes.
Based on the above, it is concluded that the proposed changes do
not result in a significant reduction of margin with respect to
plant safety as defined in the safety analysis report or TSs.
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: September 30, 2004.
Description of amendment request: The proposed amendment would
revise the technical specifications to relocate the requirements for
the emergency diesel generator start loss of power instrumentation and
associated actions in the engineering safety features tables to a new
limiting conditions for operation (LCO). In addition, an upper
allowable value has been added to the voltage sensors for loss of
voltage and degraded voltage consistent with Technical Specification
Task Force (TSTF) Item TSTF-365 along with a lower allowable value
limit for the degraded voltage diesel generator start and load shed
timer. The auxiliary feedwater loss of power start setpoints and
allowable values have been relocated to this new LCO.
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 relocation and enhancement of the loss of power functions to
a new LCO does not alter the intended functions of this feature or
physically alter these systems. Changes to Avs [allowable values]
have been evaluated in accordance with TVA [Tennessee Valley
Authority] setpoint methodology and have been verified to acceptably
protect the associated safety limits. Format changes provide a
clearer representation of the requirements and provide more
consistency with the standard TSs [Technical Specifications] in
NUREG-1431. The EDG [emergency diesel generator] and AFW [auxiliary
feedwater] start functions provided by this instrumentation are
utilized for the mitigation of accident conditions and are not
considered to be a potential source for accident generation.
Additionally, these start functions are enhanced by the addition of
an upper allowable value limit such that the accident mitigation
functions are not challenged unnecessarily. This further assures the
ability to mitigate accidents and maintain acceptable offsite dose
limits. These changes continue to support or improve the required
safety functions; 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 for the loss of power instrumentation will
not alter plant processes, components, or operating practices. The
function to start the EDGs and AFW pumps on a loss of voltage or
degraded voltage to the shutdown boards will not be altered by the
proposed change. Additionally, the EDGs and AFW system is not
considered to be a source for the generation of postulated
accidents. Therefore, the proposed change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes do not alter any plant settings or
functions that are utilized to mitigate accident conditions. The
enhanced allowable values for the voltage sensors help to prevent
unnecessary actuation of mitigation systems to ensure their ability
to respond to actual accident conditions. The parameters that ensure
the required margin of safety will be maintained with the proposed
changes or improved. 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket Nos. 50-327 and 50-328, Sequoyah
Nuclear Plant, Units 1 and 2, Hamilton County, Tennessee
Date of amendment request: December 2, 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
[[Page 2901]]
place for complying with the requirements of title 10 of the Code of
Federal Regulations (10 CFR), part 50, Sec. 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 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 December 2, 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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: September 15, 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, Sec. 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 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 September 15, 2004.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or
[[Page 2902]]
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: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: November 8, 2004.
Description of amendment request: The requested change will delete
Technical Specification (TS) 5.9.1, ``Occupational Radiation Exposure
Report,'' and TS 5.9.4, ``Monthly Operating Reports.''
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated November 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Section Chief: Michael L. Marshall, Jr.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia; Docket Nos. 50-280 and 50-281, Surry Power Station, Units No.
1 and 2, Surry County, VA
Date of amendment request: September 8, 2004.
Description of amendment request: The proposed amendments delete
the requirements from the technical specifications (TS) to maintain
hydrogen recombiners (North Anna Power Station only) and hydrogen
monitors (North Anna and Surry Power Stations). 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 TS for nuclear power reactors currently licensed to operate. The
revised title 10 of the Code of Federal Regulations (10 CFR), Sec.
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 September 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:
[[Page 2903]]
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident Previously
Evaluated
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design-
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design-basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
Category 1 in RG 1.97 is intended for key variables that most
directly indicate the accomplishment of a safety function for
design-basis accident events. The hydrogen monitors no longer meet
the definition of Category 1 in RG 1.97. As part of the rulemaking
to revise 10 CFR 50.44 the Commission found that Category 3, as
defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the severe accident
management guidelines (SAMGs), the emergency plan (EP), the
emergency operating procedures (EOP), and site survey monitoring
that support modification of emergency plan protective action
recommendations (PARs).
Therefore, the elimination of the hydrogen recombiner
requirements and relaxation of the hydrogen monitor requirements,
including removal of these requirements from TS, does not involve a
significant increase in the probability or the consequences of any
accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident From Any Previously Evaluated
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, will not result in any failure mode
not previously analyzed. The hydrogen recombiner and hydrogen
monitor equipment was intended to mitigate a design-basis hydrogen
release. The hydrogen recombiner and hydrogen monitor equipment are
not considered accident precursors, nor does their existence or
elimination have any adverse impact on the pre-accident state of the
reactor core or post accident confinement of radionuclides within
the containment building.
Therefore, this change does not create the possibility of a new
or different kind of accident from any previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The elimination of the hydrogen recombiner requirements and
relaxation of the hydrogen monitor requirements, including removal
of these requirements from TS, in light of existing plant equipment,
instrumentation, procedures, and programs that provide effective
mitigation of and recovery from reactor accidents, results in a
neutral impact to the margin of safety.
The installation of hydrogen recombiners and/or vent and purge
systems required by 10 CFR 50.44(b)(3) was intended to address the
limited quantity and rate of hydrogen generation that was postulated
from a design-basis LOCA. The Commission has found that this
hydrogen release is not risk-significant because the design-basis
LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage.
Category 3 hydrogen monitors are adequate to provide rapid
assessment of current reactor core conditions and the direction of
degradation while effectively responding to the event in order to
mitigate the consequences of the accident. The intent of the
requirements established as a result of the TMI, Unit 2 accident can
be adequately met without reliance on safety-related hydrogen
monitors.
Therefore, this change does not involve a significant reduction
in the margin of safety. Removal of hydrogen monitoring from TS will
not result in a significant reduction in their functionality,
reliability, and availability.
The NRC staff proposes to determine that the amendment requests
involve no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia; Docket Nos. 5050-280 and 50-281, Surry Power Station, Unit
No. 1 and No. 2, Surry County, Virginia
Date of amendment request: December 21, 2004.
Description of amendment request: The requested change will delete
Technical Specification requirements for the licensee to submit annual
occupational radiation exposure reports and monthly operating reports.
The NRC staff issued a notice of availability of a model no
significant hazards consideration (NSHC) determination for referencing
in license amendment applications in the Federal Register on June 23,
2004 (69 FR 35067). The licensee affirmed the applicability of the
model NSHC determination in its application dated December 21, 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:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change eliminates the Technical Specifications
(TSs) reporting requirements to provide a monthly operating letter
report of shutdown experience and operating statistics if the
equivalent data is submitted using an industry electronic database.
It also eliminates the TS reporting requirement for an annual
occupational radiation exposure report, which provides information
beyond that specified in NRC regulations. The proposed change
involves no changes to plant systems or accident analyses. As such,
the change is administrative in nature and does not affect
initiators of analyzed events or assumed mitigation of accidents or
transients. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant, add any new equipment, or require any existing equipment
to be operated in a manner different from the present design.
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.
[[Page 2904]]
This is an administrative change to reporting requirements of
plant operating information and occupational radiation exposure
data, and has no effect on plant equipment, operating practices or
safety analyses assumptions. For these reasons, the proposed change
does not involve a significant reduction in the margin of safety.
Based upon the reasoning presented above, the requested change does
not involve significance hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Wolf Creek Nuclear Operating Corporation, Docket No. 50-482, Wolf Creek
Generating Station, Coffey County, Kansas
Date of amendment request: December 13, 2004.
Description of amendment request: This amendment would revise
Technical Specification Surveillance Requirement (SR) 3.8.1.7 (fast-
start test), SR 3.8.1.12 (safety injection actuation signal test), SR
3.8.1.15 (hot restart test), and SR 3.8.1.20 (redundant unit test) to
clarify what voltage and frequency limits are applicable during the
transient and steady state portions of the diesel generator (DG) start
testing performed by these SRs.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change does not affect the DGs ability to supply
the minimum voltage and frequency within 12 seconds or the steady
state voltage and frequency. The DGs will continue to perform their
intended safety function, in accordance with the safety analysis.
The design of plant equipment is not being modified by the proposed
change. In addition, the DGs and their associated emergency loads
are accident mitigating features. As such, testing of the DGs
themselves is not associated with any potential accident-initiating
mechanism.
The proposed changes do 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. The proposed changes do not alter or
prevent the ability of structures, systems, and components (SSCs)
from performing their intended function to mitigate the consequences
of an initiating event within the assumed acceptance limits. The
proposed changes do not affect the source term, containment
isolation, or radiological release assumptions used in evaluating
the radiological consequences of an accident previously evaluated.
Further, the proposed changes do not increase the types or amounts
of radioactive effluent that may be released offsite, nor
significantly increase individual or cumulative occupational [or]
public radiation exposures. The proposed changes are consistent with
the safety analysis assumptions and resultant consequences.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different accident from any accident previously evaluated.
The proposed change revises surveillance requirements to clarify
what voltage and frequency limits are applicable during the
transient and steady state portions of the DG start testing. No
changes are being made in equipment hardware, operational
philosophy, testing frequency, system operation, or how the DGs are
physically tested.
The proposed changes do not result in a change in the manner in
which the electrical distribution subsystems provide plant
protection. The changes do not alter assumptions made in the safety
analysis. The proposed changes are consistent with the safety
analysis assumptions and current plant operating practice.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety is related to the confidence in the ability
of the fission product barriers to perform their design functions
during and following an accident situation. These barriers include
the fuel cladding, the reactor coolant system, and the containment
system. The proposed change does not directly affect these barriers,
nor do they involve any significantly adverse impact on the DGs
which serve to support these barriers in the event of an accident
concurrent with a loss of offsite power.
The proposed 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 these changes. The proposed changes will not
result in plant operation in a configuration outside the design
basis.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Jay Silberg, Esq., Shaw, Pittman, Potts and
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Robert A. Gramm.
Previously Published Notices of Consideration of Issuance of Amendments
to Facility Operating Licenses, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing
The following notices were previously published as separate
individual notices. The notice content was the same as above. They were
published as individual notices either because time did not allow the
Commission to wait for this biweekly notice or because the action
involved exigent circumstances. They are repeated here because the
biweekly notice lists all amendments issued or proposed to be issued
involving no significant hazards consideration.
For details, see the individual notice in the Federal Register on
the day and page cited. This notice does not extend the notice period
of the original notice.
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 amendment: June 7, 2004.
Brief description of amendment: The proposed amendment revised
Technical Specification 3.9.4, ``Shutdown Cooling (SDC) and Coolant
Circulation-High Water Level,'' to incorporate the use of an alternate
cooling method to function as a path for decay heat removal when in
MODE 6 with the refueling pool fully flooded. The spent fuel pool
cooling system is the alternative cooling method intended to be used as
a substitute for the SDC system during the refueling operations,
including during fuel movement.
Date of publication of individual notice in Federal Register:
November 29, 2004 (69 FR 69417).
Expiration date of individual notice: January 27, 2005.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment request: November 3, 2004.
Brief description of amendment request: The proposed amendments
would revise Technical Specification (TS) 3.7.17 and TS 4.3 for Cycles
14-16 to allow installation and use of a temporary cask pit spent fuel
storage
[[Page 2905]]
rack (cask pit rack) for Diablo Canyon Power Plant, Unit Nos. 1 and 2.
The total spent fuel pool storage capacity for each unit would be
increased from 1324 fuel assemblies to 1478 fuel assemblies for Cycles
14-16.
Date of publication of individual notice in Federal Register:
December 21, 2004 (69 FR 76486).
Expiration date of individual notice: February 22, 2005.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room (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 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 PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Connecticut Yankee Atomic Power Company, Docket No. 50-213, Haddam Neck
Plant, Middlesex County, Connecticut
Date of amendment request: August 11, 2004.
Brief description of amendment: The amendment revises Technical
Specifications to eliminate operational requirements and certain design
requirements that will no longer be applicable following the transfer
of all of the spent fuel from the Haddam Neck Plant spent fuel pool
into dry cask storage at the Haddam Neck Plant Independent Spent Fuel
Storage Installation. The amendment relocates administrative
requirements to the Connecticut Yankee Quality Assurance Program. The
amendment also deletes the requirement for submittal of an annual
Occupational Radiation Exposure Report.
Date of issuance: December 20, 2004.
Effective date: As of the date that all reactor fuel has been
permanently removed from the spent fuel pool and stored in an
Independent Spent Fuel Storage Installation. The license amendment
shall be implemented within 60 days of its effective date.
Amendment No.: 201.
Facility Operating License No. DPR-61: The amendment revises the
Technical Specifications.
Date of initial notice in Federal Register: September 28, 2004 (69
FR 57978).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation Report, dated December 20, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: October 12, 2004.
Brief description of amendment: This amendment approves an
engineering evaluation performed in accordance with the Pilgrim Nuclear
Power Station Technical Specifications (TS). TS 3.6.D.3 requires the
licensee to perform an engineering evaluation when safety relief valve
(SRV) discharge pipe temperatures exceed 212 [deg]F during normal
reactor power operation for a period greater than 24 hours, and TS
3.6.D.4 further requires that power operation may not continue beyond
90 days from the initial discovery of discharge pipe temperatures in
excess of 212 [deg]F, without prior NRC approval of the engineering
evaluation. The Nuclear Regulatory Commission staff has reviewed the
engineering evaluation and has determined that the licensee has
adequately justified power operations beyond the end of the TS-required
90-day period for plant shutdown, until the next cold shutdown of 72
hours or more.
Date of issuance: December 23, 2004.
Effective date: As of the date of issuance, and shall be
implemented within 30 days.
Amendment No.: 208.
Facility Operating License No. DPR-35: Amendment does not revise
the Technical Specifications.
Date of initial notice in Federal Register: October 20, 2004 (69
FR 61695).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2004.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: December 8, 2003.
Brief description of amendment: The proposed amendment would delete
a portion of the Pilgrim Nuclear Power Station (Pilgrim) Technical
Specification (TS) 4.6.A.2, ``Primary System Boundary--Thermal and
Pressurization Limitations,'' and the associate TS Table 4.6-3,
``Reactor Vessel Material Surveillance Program Withdrawal Schedule.''
The amendment would replace the existing Reactor Vessel Material
Surveillance Program with the Boiling Water Reactor Vessel and Internal
Project (BWRVIP) Integrated Surveillance Program (ISP) and Supplemental
Surveillance Program (SSP). The BWRVIP ISP/SSP would be incorporated
into the Pilgrim Updated Final Safety Analysis Report (UFSAR).
Date of issuance: January 5, 2005.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 209.
Facility Operating License No. DPR-35: Amendment revised the
Technical Specifications and updated the UFSAR.
Date of initial notice in Federal Register: February 17, 2004 (69
FR 7521).
[[Page 2906]]
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated January 5, 2005.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: August 11, 2003, as supplemented
January 9, May 3, and July 19, 2004.
Brief description of amendment: This amendment relocates the
Technical Specification requirement to leak rate test the enclosure for
decay heat removal system valves DH-11 and DH-12 to the Technical
Requirements Manual.
Date of issuance: December 21, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 263.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54750).
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 amendment is contained
in a Safety Evaluation dated December 21, 2004.
No significant hazards consideration comments received: No.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: May 27, 2004, as supplement by letter
dated September 28, 2004.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) to lower the reactor vessel water level at which
the reactor water cleanup system isolates, secondary containment
isolates, and the control room emergency filter system starts.
Date of issuance: December 23, 2004.
Effective date: As of the date of issuance and shall be implemented
upon startup in Operating Cycle 23.
Amendment No.: 209.
Facility Operating License No. DPR-46: Amendment revised the TS.
Date of initial notice in Federal Register: June 22, 2004 (69 FR
34702).
The supplement dated September 28, 2004, provided additional
information that clarified the application, did not expand the scope of
the application as originally noticed, and did not change the staff's
original proposed no significant hazards consideration determination as
published in the Federal Register.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated December 23, 2004.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket Nos. 50-266 and 50-301, Point
Beach Nuclear Plant, Units 1 and 2, Town of Two Creeks, Manitowoc
County, Wisconsin
Date of application for amendments: December 23, 2003.
Brief description of amendments: The amendments modified TS
requirements to adopt the provisions of Industry/TS Task Force (TSTF)
change TSTF-359, ``Increased Flexibility in Mode Restraints.'' The
availability of TSTF-359 for adoption by licensees was announced in the
Federal Register on April 4, 2003 (68 FR 16579).
Date of issuance: December 22, 2004.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment Nos.: 215, 220.
Facility Operating License Nos. DPR-24 and DPR-27: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 16, 2004 (69
FR 55844)
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 22, 2004.
No significant hazards consideration comments received: No.
Southern California Edison Company, et al., Docket Nos. 50-361 and 50-
362, San Onofre Nuclear Generating Station, Units 2 and 3, San Diego
County, California
Date of application for amendments: August 4, 2003, as supplemented
by letters dated December 24, 2003, and June 3, August 24, and October
6 and 22, 2004.
Brief description of amendments: The proposed amendments would
revise Technical Specification 3.9.3, ``Containment Penetrations,'' by
adding a note to the limiting condition for operation that permits the
containment equipment hatch to be open during core alterations and
movement of irradiated fuel in containment during refueling operations.
Date of issuance: December 23, 2004.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 193/184.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: September 18, 2003 (68
FR 54752). The supplemental letters dated December 24, 2003, and June
3, August 24, October 6, and October 22, 2004, provided additional
information that clarified the application, 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 December 23, 2004.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket No. 50-498, South Texas Project,
Unit 1, Matagorda County, Texas
Date of amendment request: September 30, 2004.
Brief description of amendment: The amendment changes Technical
Specification (TS) Surveillance Requirement 4.4.4.2 to expand the range
of conditions under which quarterly testing of block valves for the
pressurizer power operated relief valves would be unnecessary.
Date of issuance: December 28, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment No.: Unit 1--166.
Facility Operating License No. NPF-76: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: October 26, 2004 (69 FR
62477).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 28, 2004.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment requests: September 22, 2003, and September 27,
2004.
Brief description of amendments: The amendments change Technical
Specification (TS) Surveillance Requirement 4.7.1.6, ``Atmospheric
Steam Relief Valves'' to provide consistency with TS 3.3.5.1,
[[Page 2907]]
``Atmospheric Steam Relief Valve Instrumentation,'' regarding
atmospheric steam relief valve automatic controls. The amendments also
correct typographical errors in TSs 3.7.1.6 and 3.2.4. The remaining
proposed changes associated with the September 22, 2003, application
were withdrawn as noted in the NRC staff's letter to the licensee dated
October 19, 2004.
Date of issuance: December 28, 2004.
Effective date: As of the date of issuance and shall be implemented
within 30 days of issuance.
Amendment Nos.: Unit 1--167; Unit 2--156.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: November 12, 2003 (68
FR 64139) for the September 22, 2003, application and October 26, 2004
(69 FR 62478) for the September 27, 2004, application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated December 28, 2004.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 10th day of January, 2005.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 05-779 Filed 1-14-05; 8:45 am]
BILLING CODE 7590-01-P