[Federal Register Volume 72, Number 146 (Tuesday, July 31, 2007)]
[Notices]
[Pages 41780-41793]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E7-14350]
-----------------------------------------------------------------------
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 July 4, 2007 to July 18, 2007. The last
biweekly notice was published on July 17, 2007 (72 FR 39081).
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)
[[Page 41781]]
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, Rulemaking,
Directives and Editing Branch, Division of Administrative Services,
Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Written comments may also
be delivered to Room 6D22, Two White Flint North, 11545 Rockville Pike,
Rockville, Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays.
Copies of written comments received may be examined at the Commission's
Public Document Room (PDR), located at One White Flint North, Public
File Area O1F21, 11555 Rockville Pike (first floor), Rockville,
Maryland. The filing of requests for a hearing and petitions for leave
to intervene is discussed below.
Within 60 days after the date of publication of this notice, the
licensee may file a request for a hearing with respect to issuance of
the amendment to the subject facility operating license and any person
whose interest may be affected by this proceeding and who wishes to
participate as a party in the proceeding must file a written request
for a hearing and a petition for leave to intervene. Requests for a
hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.309, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed by: (1) First class mail addressed to the Office of the
Secretary of the Commission, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, Attention: Rulemaking and Adjudications
Staff; (2) courier, express
[[Page 41782]]
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].
Carolina Power & Light Company, et al., Docket No. 50-400, Shearon
Harris Nuclear Power Plant, Unit 1, Wake and Chatham Counties, North
Carolina
Date of amendment request: June 15, 2007.
Description of amendment request: The amendment proposes to
relocate the inservice testing requirements to the administrative
section of the technical specifications (TS), remove the inservice
inspection activities from TS and locate them in an owner-controlled
program, and establish a TS Bases Control Program. All of these changes
are proposed to be consistent with NUREG-1431, Revision 3, ``Standard
Technical Specifications Westinghouse Plants.''
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, 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 does not involve a significant increase in
the probability or consequences of an accident previously evaluated,
and it does not change an accident previously evaluated in the Final
Safety Analysis Report (FSAR). The proposed change is administrative in
nature, and it will continue to ensure that the inspection and testing
requirements required by regulations are met. The American Society of
Mechanical Engineers (ASME) Code requirements are established, reviewed
and approved by ASME, the industry, and ultimately endorsed by the NRC
for inclusion into 10 CFR 50.55a. Updates to the ASME Code reflect
advances in technology and consider information obtained from plant
operating experience to provide enhanced inspection and testing. Thus,
the proposed change will revise TS to appropriately reference the ASME
Code required by 10 CFR 50.55a for performing inservice testing,
specifically referencing the ASME Code for Operation and Maintenance of
Nuclear Power Plants, rather than the ASME Section XI Code.
The proposed change does not affect operations, and the inspection
and testing required is not an accident initiator.
Therefore, this amendment 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 of
different kind of accident from any accident previously evaluated?
Response: No.
The proposed amendment does not create the possibility of a new or
different kind of accident from any accident previously evaluated, and
it does not change an accident previously evaluated in the Final Safety
Analysis Report (FSAR). As noted above, the proposed change is
administrative in nature, the inspection and testing required is not an
accident initiator, and no new accident precursors are being
introduced. The proposed change will revise TS to appropriately
reference the ASME Code required by 10 CFR 50.55a for performing
inservice testing, which will continue to ensure that the inspection
and testing requirements required by regulations are met. Since
inservice testing will continue to be performed in accordance with
regulations, adequate assurance is provided to ensure that the safety-
related pumps and valves will continue to operate as required. No new
testing is required that could create a new or different type of
accident.
Therefore, this amendment does not create the possibility of a new
or different kind of accident from any accident previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment does not involve a significant reduction in
a margin of safety. The proposed amendment does not adversely affect a
plant safety limit or a limiting safety system setting, and does not
alter a design basis limit for a parameter evaluated in the FSAR. The
proposed change is administrative in nature, and it will continue to
ensure that the inspection and testing requirements required by
regulations are met. Since inservice testing will continue to be
performed in accordance with regulations, adequate assurance is
provided to ensure that the safety-related pumps and valves will
continue to operate as required and perform their intended safety
function.
Therefore, this amendment 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: David T. Conley, Associate General Counsel
II--Legal Department, Progress Energy Service Company, LLC, Post Office
Box 1551, Raleigh, North Carolina 27602.
NRC Branch Chief: Thomas H. Boyce.
Entergy Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 3, 2007.
Description of amendment request: The proposed change relocates the
quality and quantity requirements
[[Page 41783]]
associated with the emergency diesel generator (EDG) fuel oil within
the Technical Specifications (TS) through the creation of a new TS
Limiting Condition for Operation and the Diesel Fuel Oil Testing
Program.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes in the diesel fuel oil testing program will
continue to ensure that new and stored diesel fuel oil properties are
maintained within specified limits to assure EDG operation. The testing
of diesel generator fuel oil is not considered an initiator or a
mitigating factor in any previously evaluated accidents.
The deletion of the requirement to drain and inspect the fuel oil
storage tank (FOST) does not impact any of the previously analyzed
accidents. Periodic testing of the fuel oil as required by the Diesel
Fuel Oil Testing Program will identify poor quality oil. Actions are
included that will require the quality of the oil to be maintained
within acceptable limits. Draining and inspecting the FOST are not
considered an accident initiator or mitigating factor in any previously
evaluated accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change results in changes to the existing diesel fuel
oil testing program and the deletion of the [Surveillance Requirements]
associated with the performance of periodic draining and inspection of
the FOSTs. No plant modifications are required to support the proposed
TS changes. There is no impact to plant structures, systems, or
components, or in the design of the plant structures, systems, or
components.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change does not result in any plant modifications.
Diesel generator fuel oil quantity and quality will continue to be
maintained within acceptable limits to assure the ability of the EDG to
perform its intended function.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Council--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Thomas G. Hiltz.
Exelon Generation Company, LLC, Docket No. 50-237, Dresden Nuclear
Power Station (DNPS), Unit 2, Grundy County, Illinois
Date of amendment request: July 10, 2007.
Description of amendment request: The proposed amendment would
revise the values of the safety limit minimum critical power ratio
(SLMCPR) in Technical Specification (TS) Section 2.1.1, ``Reactor Core
SLs.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No
The probability of an evaluated accident is derived from the
probabilities of the individual precursors to that accident. The
consequences of an evaluated accident are determined by the operability
of plant systems designed to mitigate those consequences. Limits have
been established consistent with NRC-approved methods to ensure that
fuel performance during normal, transient, and accident conditions is
acceptable. The proposed change conservatively establishes the SLMCPR
for DNPS, Unit 2, Cycle 21 such that the fuel is protected during
normal operation and during plant transients or anticipated operational
occurrences (AOOs).
Changing the SLMCPR does not increase the probability of an
evaluated accident. The change does not require any physical plant
modifications, physically affect any plant components, or entail
changes in plant operation. Therefore, no individual precursors of an
accident are affected.
The proposed change revises the SLMCPR to protect the fuel during
normal operation as well as during plant transients or AOOs.
Operational limits will be established based on the proposed SLMCPR to
ensure that the SLMCPR is not violated. This will ensure that the fuel
design safety criterion (i.e., that at least 99.9% of the fuel rods do
not experience transition boiling during normal operation and AOOs) is
met. Since the proposed change does not affect operability of plant
systems designed to mitigate any consequences of accidents, the
consequences of an accident previously evaluated are not expected to
increase.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No
Creation of the possibility of a new or different kind of accident
requires creating one or more new accident precursors. New accident
precursors may be created by modifications of plant configuration,
including changes in allowable modes of operation. The proposed change
does not involve any plant configuration modifications or changes to
allowable modes of operation.
The proposed change to the SLMCPR assures that safety criteria are
maintained for DNPS, Unit 2, Cycle 21.
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 SLMCPR provides a margin of safety by ensuring that at least
99.9% of the fuel rods do not experience transition boiling during
normal operation and AOOs if the MCPR limit is not violated. The
proposed change will ensure the current level of fuel protection is
maintained by continuing to ensure that at least 99.9% of the fuel rods
do not experience transition boiling during normal operation and AOOs
if the MCPR limit is not violated. The proposed SLMCPR values were
developed using NRC-approved methods. Additionally, operational
[[Page 41784]]
limits will be established based on the proposed SLMCPR to ensure that
the SLMCPR is not violated. This will ensure that the fuel design
safety criterion (i.e., that no more than 0.1% of the rods are expected
to be in boiling transition if the MCPR limit is not violated) is met.
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
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket No. 50-373, LaSalle County
Station, Unit 1, LaSalle County, Illinois
Date of amendment request: June 18, 2007.
Description of amendment request: The proposed amendment would
revise technical specification TS 5.5.13, ``Primary Containment Leakage
Rate Testing Program,'' to reflect a one-time extension of the LaSalle
County Station (LSCS), Unit 1, primary containment Type A Integrated
Leak Rate Test (ILRT) date for the current requirement of no later than
June 13, 2009, prior to startup following the thirteenth LSCS Unit 1
refueling outage (L1R13).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No
The proposed changes will revise LSCS, Unit 1, TS 5.5.13, ``Primary
Containment Leakage Rate Testing Program,'' to reflect a one-time
extension of the primary containment Type A Integrated Leak Rate Test
(ILRT) date to ``prior to startup following L1R13.'' The current Type A
ILRT interval of 15 years, based on past performance, would be extended
on a one-time basis by approximately 5% of the current interval.
The function of the primary containment is to isolate and contain
fission products released from the reactor Primary Coolant System (PCS)
following a design basis Loss of Coolant Accident (LOCA) and to confine
the postulated release of radioactive material to within limits. The
test interval associated with Type A ILRTs is not a precursor of any
accident previously evaluated. Type A ILRTs provide assurance that the
LSCS Unit 1 primary containment will not exceed allowable leakage rate
values specified in the TS and will continue to perform their design
function following an accident. The risk assessment of the proposed
changes has concluded that there is an insignificant increase in total
population dose rate and an insignificant increase in the conditional
containment failure probability.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
Response: No
The proposed changes for a one-time extension of the Type A ILRT
for LSCS Unit 1 will not affect the control parameters governing unit
operation or the response of plant equipment to transient and accident
conditions. The proposed changes do not introduce any new equipment,
modes of system operation or failure mechanisms.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Does the change involve a significant reduction in a margin of
safety?
Response: No
LSCS Unit 1 is a General Electric BWR/5 plant with a Mark II
primary containment. The Mark II primary containment consists of two
compartments, the drywell and the suppression chamber. The drywell has
the shape of a truncated cone, and is located above the cylindrically
shaped suppression chamber. The drywell floor separates the drywell and
the suppression chamber. The primary containment is penetrated by
access, piping and electrical penetrations.
The integrity of the primary containment penetrations and isolation
valves is verified through Type B and Type C local leak rate tests
(LLRTs) and the overall leak tight integrity of the primary containment
is verified by a Type A ILRT, as required by 10 CFR 50, Appendix J,
``Primary Reactor Containment Leakage Testing for Water-Cooled Power
Reactors.'' These tests are performed to verify the essentially leak
tight characteristics of the primary containment at the design basis
accident pressure. The proposed changes for a one-time extension of the
Type A ILRT does not affect the method for Type A, B, or C testing or
the test acceptance criteria.
EGC has conducted a risk assessment to determine the impact of a
change to the LSCS Unit 1 Type A ILRT schedule from a baseline ILRT
frequency of three times in ten years to once in 15.67 years (i.e., 15
years plus 8 months) for the risk measures of Large Early Release
Frequency (i.e., LERF), Total Population Dose, and Conditional
Containment Failure Probability (i.e., CCFP). This assessment indicated
that the proposed LSCS ILRT interval extension has a minimal impact on
public risk.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: November 27, 2006.
Description of amendment request: The proposed amendments would
modify various technical specification (TS) requirements for emergency
diesel generators (EDGs). Specifically, the licensee stated that the
proposed changes would eliminate several accelerated tests and a test
table, modify acceptance criteria for fast start and load rejection
tests, and also, eliminate the EDG failure report. The proposed changes
are consistent with the Nuclear Regulatory Commission's (NRC's)
regulatory guidance presented in Generic Letter 93-05, ``Line-Item
Technical Specifications Improvement to Reduce Surveillance
Requirements for Testing During Power Operation,'' Generic Letter 94-
01, ``Removal of Accelerated Testing and Special Reporting Requirements
for Emergency Diesel Generators,'' and NUREG-1433, Rev. 3.1, ``Standard
Technical Specifications, General Electric Plants, BWR/4.''
[[Page 41785]]
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes are associated with the testing and reporting
requirements of the eight (four on each unit) Emergency Diesel
Generators (EDGs). The changes will eliminate unnecessary EDG testing
requirements that contribute to potential mechanical degradation of the
EDGs. The changes are based on the NRC guidance and recommendations
provided in Generic Letter 93-05 or Generic Letter 94-01, or are
consistent with NUREG-1433. The change to the reporting requirement is
administrative in nature.
The probability of an accident is not increased by these changes
because the EDGs are not assumed to be initiators of any design basis
event. Additionally, the proposed changes do not involve any physical
changes to plant systems, structures, or components (SSC), or the
manner in which these SSC are operated, maintained, or controlled. The
consequences of an accident will not be increased because the changes
to the EDGs and associated support systems still provide a high degree
of assurance that their operability is maintained.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident previously
evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes do not alter the physical design, safety
limits, or safety analysis assumptions, associated with the operation
of the plant. Accordingly, the proposed changes do not introduce any
new accident initiators, nor do they reduce or adversely affect the
capabilities of any plant structure or system in the performance of
their safety function.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes to the EDGs either: (1) Modify the test
acceptance criteria, (2) modify the accelerated testing schedules, or
(3) eliminate a reporting requirement. The change to the test
acceptance criteria is based on the recommendations of Regulatory Guide
1.9, and the change to the reporting requirement is enveloped by other
NRC reporting requirements. The other changes are consistent with NRC
guidance, and reduce unnecessary testing and improve EDG reliability.
Requirements to assure that a common mode failure has not affected the
remaining operable EDGs have been maintained. The existing routine
testing frequency, unaffected by these changes, has been shown to be
adequate for assuring the EDGs are operable based on operating
experience. The proposed changes do not impact the assumptions of any
design basis accident, and do not alter assumptions relative to the
mitigation of an accident or transient event.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. Bradley Fewell, Associate General
Counsel, Exelon Generation Company, LLC, 4300 Winfield Road,
Warrenville, IL 60555.
NRC Branch Chief: Harold K. Chernoff.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: March 29, 2007.
Description of amendment request: The proposed amendment would
revise the Seabrook Station, Unit No. 1 Technical Specifications to
increase the power level required for a reactor trip following a
turbine trip (P-9 setpoint).
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The analysis of the proposed change included an evaluation of loss
of load/turbine trip transient. With systems functioning as designed,
the proposed change to the P-9 setpoint does not impact [the] accident
analyses previously evaluated in the Updated Final Safety Analysis
Report (UFSAR). In the best estimate case (normal plant conditions; all
control systems functioning per design), the pressurizer power operated
relief valves (PORV) and the steam generator safety valves are not
challenged following the turbine trip without reactor trip.
Consequently, the proposed change does not adversely affect the
probability of a small break loss of coolant accident due to a stuck-
open PORV. The sensitivity study that assessed the affects of degraded
control systems found that a failure of all condenser steam dump valves
resulted in challenging the PORVs and the steam generator (SG) safety
valves. However, overfilling of the pressurizer will not occur and this
Condition 2 event will not initiate a Condition 3 event. The challenge
to the PORVs with all steam dump banks failed does not violate design
or licensing criteria. Therefore, the proposed setpoint change does not
significantly increase the probability or consequences of an accident
previously evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed setpoint change does not create the possibility of a
new or different kind of accident than any accident previously
evaluated in the FSAR. No new accident scenarios, failure mechanisms or
limiting single failures are introduced as a result of the proposed
change. The proposed Technical Specification changes have no adverse
effects on any safety-related system and do not challenge the
performance or integrity of any safety-related system. The revised
setpoint for the P-9 function ensures that accident/transient analyses
acceptance criteria continue to be met. This change makes no
modifications to the plant that would introduce new accident causal
mechanisms and has no affect on how the trip functions operate upon
actuation. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any previously
evaluated.
3. The proposed changes do not involve a significant reduction in
the margin of safety.
The proposed Technical Specification changes do not involve a
significant reduction in a margin of safety. The analyses supporting
the proposed change to the P-9 setpoint demonstrate that margin exists
between the setpoint and the corresponding safety analysis limits. The
calculations are based on plant instrumentation and calibration/
[[Page 41786]]
functional test methods and include allowances associated with the
setpoint change. The results of analyses and evaluations supporting the
proposed change demonstrate acceptance criteria continue to be met. The
reactor trip on turbine trip provides additional protection and
conservatism beyond that required for protection of public health and
safety; the safety analyses in chapter 15 of the UFSAR do not take
credit for this reactor trip. 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: M.S. Ross, Florida Power & Light Company,
P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Branch Chief: Harold K. Chernoff.
Indiana Michigan Power Company, Docket Nos. 50-315 and 50-316, Donald
C. Cook Nuclear Plant, Units 1 and 2, Berrien County, Michigan
Date of amendment request: June 27, 2007.
Description of amendment request: The proposed amendment would
revise Technical Specifications (TS) 3.3.3, ``Post Accident Monitoring
(PAM) Instrumentation,'' to include containment recirculation sump
level instrumentation which will be used for indication of
recirculation sump strainer blockage. Additionally, the amendment would
revise TS 3.5.2, ``ECCS [Emergency Core Cooling System]--Operating,''
by replacing the term ``trash racks and screens'' with the more
descriptive term ``strainers.'' Finally, the amendment would revise TS
3.6.14, ``Containment Recirculation Drains,'' to include Limiting
Conditions for Operation, Actions, and Surveillance Requirements to
ensure the operability of flow paths credited in the evaluation of
potential adverse effects of post-accident debris on the containment
recirculation function pursuant to NRC Generic Letter 2004-02.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated?
Response: No.
The proposed change consists of a revision to the Technical
Specifications (TS) for post accident monitoring (PAM) instrumentation
to include new containment recirculation sump level instrumentation, a
revision to the TS for Emergency Core cooling System (ECCS) to replace
the term ``trash rack and screen'' with the term ``strainer,'' and a
revision to the TS for containment recirculation drains to add two flow
paths credited in the evaluation of the effects of post-accident debris
on the containment recirculation functions pursuant to Nuclear
Regulatory Commission Generic Letter 2004-02.
The proposed TS revisions will not increase the probability of an
accident because the associated components, i.e., the new sump level
instruments, the new strainers, and the two flow paths, are not, and
will not become, accident initiators. The activities involving these
components pursuant to the proposed TS revisions consist of
implementing Surveillance Requirements for the new sump level
instruments and flow paths and actions to be taken if these components
are inoperable. These activities will not increase the likelihood of an
accident. The TS change associated with the sump strainers is editorial
in that it reflects the terminology that has been applied to new pocket
strainers that continue to perform the trash rack and screen functions.
The change in terminology will not result in any new activities.
The proposed TS revision will not increase the consequences of an
accident because the associated components all provide mitigative
functions for an accident, and their ability to perform their
mitigative functions is not reduced by the associated TS changes. The
TS changes associated with the new sump level instrumentation and the
recirculation [flow paths] will provide increased assurance that these
components will be available to perform their mitigative function if
needed. The TS change associated with the sump strainers is editorial
and does not affect the mitigative capability of the screens.
Therefore, the proposed change will 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 TS revisions will not create the possibility of a new
or different kind of accident from any accident previously evaluated
because the associated components, i.e., the new sump level
instruments, the new strainers, and the two flow paths, are components
that will not initiate any accident. The proposed TS changes associated
with these components will not cause them to be operated in any manner
not previously evaluated for the specific components or for similar
components, or cause them to become other than passive components.
Therefore, the proposed change will 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 margin of safety associated with the proposed TS revisions
involves the ability of the associated components, i.e., the new sump
level instruments, the new strainers, and the two flow paths, to assure
the ECCS and containment spray recirculation function can be adequately
accomplished. The TS changes associated with the new sump level
instrumentation and the recirculation [flow paths] will provide
increased assurance that this function can be fulfilled. The TS change
associated with the sump strainers is editorial and does not affect
this function.
Therefore, the proposed change will not involve a significant
reduction in the margin of safety.
The 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 requests involve no
significant hazards consideration.
Attorney for licensee: Kimberly Harshaw, Esquire, One Cook Place,
Bridgman, MI 49106.
NRC Acting Branch Chief: Travis Tate.
Nine Mile Point Nuclear Station (NMPNS), LLC, Docket No. 50-410, Nine
Mile Point Nuclear Station Unit No. 2 (NMP2), Oswego County, New York
Date of amendment request: May 31, 2007.
Description of amendment request: The proposed amendment would
revise the accident source term used in the NMP2 design basis
radiological consequence analyses in accordance with Title 10 of the
Code of Federal Regulations (10 CFR), Part 50.67. The revised accident
source term replaces the current methodology that is based
[[Page 41787]]
on TID-14844, ``Calculation of Distance Factors for Power and Test
Reactor Sites,'' with the alternative source term (AST) methodology
described in Regulatory Guide (RG) 1.183, ``Alternative Source Terms
for Evaluating Design Basis Accidents at Nuclear Power Reactors.'' The
amendment request is for full implementation of the AST as described in
RG 1.183, with the exception that TID-14844 will continue to be used as
the radiation dose basis for equipment qualification and vital area
access. Proposed changes include the following: Revision of the
Technical Specification (TS) definition of Dose Equivalent I-131 to be
consistent with the AST analyses; TS changes that reflect revised
design requirements regarding the use of the standby liquid control
system (SLCS) to buffer the suppression pool pH to prevent iodine re-
evolution following a postulated design basis loss-of-coolant accident
(LOCA); revisions to the TS operability requirements for the control
room envelope filtration system and the control room envelope air
conditioning system, consistent with the assumptions contained in the
AST fuel-handling accident (FHA) analysis; and credit for operation of
the residual heat removal system in the drywell spray mode for the
post-LOCA removal of airborne elemental iodine and particulates from
the drywell atmosphere. Because NMPNS is considering an extended power
uprate (EPU) project that would increase the maximum licensed reactor
core power level to 3,988 megawatts thermal (MWt), the AST analyses
have been performed using a bounding core isotopic inventory that is
based on operation at 3,988 MWt in lieu of the currently licensed power
of 3,467 MWt.
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.
Adoption of the AST and those plant systems affected by
implementing AST do not initiate DBAs [design-basis accidents]. The AST
does not affect the design or manner in which the facility is operated;
rather, for postulated accidents, the AST is an input to calculations
that evaluate the radiological consequences. The AST does not by itself
affect the post-accident plant response or the actual pathway of the
radiation released from the fuel. It does, however, better represent
the physical characteristics of the release, so that appropriate
mitigation techniques may be applied. Implementation of the AST has
been incorporated in the analyses for the limiting DBAs at NMP2.
The structures, systems and components affected by the proposed
change mitigate the consequences of accidents after the accident has
been initiated. Application of the AST does result in changes to NMP2
Updated Safety Analysis Report (USAR) functions (e.g., Standby Liquid
Control system [SLCS]). As a condition of application of AST, NMPNS is
proposing to use the [SLCS] to control the suppression pool pH
following a LOCA. These changes do not require any physical
modifications to the plant. As a result, the proposed changes do not
involve a revision to the parameters or conditions that could
contribute to the initiation of a DBA discussed in Chapter 15 of the
NMP2 USAR. Since design basis accident initiators are not being altered
by adoption of the AST, the probability of an accident previously
evaluated is not affected.
Plant-specific AST radiological analyses have been performed and,
based on the results of these analyses, it has been demonstrated that
the dose consequences of the limiting events considered in the analyses
are within the acceptance criteria provided by the NRC for use with the
AST. These criteria are presented in 10 CFR 50.67 and Regulatory Guide
1.183. Even though the AST dose limits are not directly comparable to
the previously specified whole body and thyroid dose guidelines of
General Design Criterion 19 and 10 CFR 100.11, the results of the AST
analyses have demonstrated that the 10 CFR 50.67 limits are satisfied.
Therefore, it is concluded that adoption of the AST does not involve a
significant increase in the consequences of an accident previously
evaluated.
Based on the above discussion, it is concluded that the proposed
changes do not involve a significant increase in the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Implementation of AST and the proposed changes does not alter or
involve any design basis accident initiators. These changes do not
involve any physical changes to the plant and do not affect the design
function or mode of operations of systems, structures, or components in
the facility prior to a postulated accident. Since systems, structures,
and components are operated essentially no differently after the AST
implementation, no new failure modes are created by this proposed
change.
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The changes proposed are associated with a new licensing basis for
analysis of NMP2 DBAs. Approval of the licensing basis change from the
original source term to the AST is being requested. The results of the
accident analyses performed in support of the proposed changes are
subject to revised acceptance criteria. The limiting DBAs have been
analyzed using conservative methodologies, in accordance with the
guidance contained in Regulatory Guide 1.183, to ensure that analyzed
events are bounding and that safety margin has not been reduced. The
dose consequences of these limiting events are within the acceptance
criteria presented in 10 CFR 50.67 and Regulatory Guide 1.183. Thus,
the proposed changes continue to ensure that the doses at the exclusion
area boundary and low population zone boundary, as well as in the
control room, are within corresponding regulatory criteria.
Therefore, by meeting the applicable regulatory criteria for AST,
it is concluded that the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mark J. Wetterhahn, Esquire, Winston &
Strawn, 1700 K Street, NW., Washington, DC 20006.
NRC Branch Chief: Mark G. Kowal.
Pacific Gas and Electric Co., Docket No. 50-133, Humboldt Bay Power
Plant (HBPP), Unit 3 Humboldt County, California
Date of amendment request: April 4, 2007.
Description of amendment request: The licensee has proposed
amending the existing license to allow the results of near-term
surveys, performed on a portion of the plant site, to be included
[[Page 41788]]
in the eventual Final Status Survey (FSS) for license termination.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change would allow survey results for a specific area
within the licensed site area, performed prior to Humboldt Bay Power
Plant (HBPP) Unit 3 decommissioning and dismantlement activities, to be
used in the overall licensed site area Final Status Survey (FSS) for
license termination. The FSS will be performed following completion of
HBPP Unit 3 decommissioning and dismantlement activities. This proposed
change would not change plant systems or accident analysis, and as
such, would not affect initiators of analyzed events or assumed
mitigation of accidents. Therefore, the proposed change does not
increase the probability or consequences of an accident previously
evaluated.
(2) Does the change create the possibility of a new or different
kind of accident from any accident evaluated?
Response: No.
The proposed change does not involve a physical alteration to the
plant or require existing equipment to be operated in a manner
different from the present design. Implementation of a cross
contamination prevention and monitoring plan will be done in accordance
with plant procedures and licensing bases documents. Therefore, the
proposed change does not create the possibility of a new or different
kind of accident from any accident evaluated.
(3) Does the change involve a significant reduction in a margin of
safety?
Response: No.
The proposed change has no effect on existing plant equipment,
operating practices, or safety analysis assumptions. 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: Ms. Jennifer K. Post, Pacific Gas and
Electric Company, 77 Beale Street, B30A, San Francisco, CA.
NRC Branch Chief: Bruce Watson.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: March 22, 2007.
Description of amendment request: The proposed amendment supports
full-scope implementation of an alternative source term (AST)
methodology, in accordance with Section 50.67, ``Accident source
term,'' of Title 10 of the Code of Federal Regulations (10 CFR) with
the exception that Technical Information Document (TID) 14844,
``Calculation of Distance Factors for Power and Test Reactor Sites,''
will continue to be used as the radiation dose basis for equipment
qualification.
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 changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The implementation of AST assumptions has been evaluated in
revisions to the analyses of the following limiting DBAs [design-basis
accidents].
Loss-of-Coolant Accident.
Fuel Handling Accident.
Control Rod Ejection Accident.
Locked Rotor Accident.
Main Steam Line Break Accident.
Steam Generator Tube Rupture Accident.
Based upon the results of these analyses and evaluations, it has
been demonstrated that, with the requested changes, the dose
consequences of these limiting events satisfies the dose limits in 10
CFR 50.67 and are within the regulatory guidance provided by the NRC
for use with the AST methodology. The AST is an input to calculations
used to evaluate the consequences of an accident and does not affect
the plant response or the actual pathway of the activity released from
the fuel. Therefore, it is concluded that AST does not involve a
significant increase in the consequences of an accident previously
evaluated.
Implementation of AST provides for elimination of the Fuel Handling
Building ventilation system filtration TS [Technical Specification]
requirements and elimination of Control Room ventilation filtration TS
requirements in Modes 5 or 6. It also eliminates containment integrity
TS requirements while handling irradiated fuel and during core
alterations. The equipment affected by the proposed changes is
mitigative in nature and relied upon after an accident has been
initiated. The affected systems are not accident initiators; and
application of the AST methodology is not an initiator of a design
basis accident.
Elimination of the requirement to suspend operations involving
positive reactivity additions that could result in loss of required
SHUTDOWN MARGIN or required boron concentration if the control room
ventilation system is inoperable in Modes 5 or 6 does not increase the
probability of an accident because the proposed change does not affect
the design and operational controls to prevent dilution events. These
same design and operational controls prevent a loss of SHUTDOWN MARGIN
or a boron dilution event so that radiological consequences from these
events are precluded.
The proposed changes do not involve physical modifications to plant
equipment and do not change the operational methods or procedures used
for moving irradiated fuel assemblies. The proposed changes do not
affect any of the parameters or conditions that could contribute to the
initiation of any accidents. Relaxation of operability requirements
during the specified conditions will not significantly increase the
probability of occurrence of an accident previously analyzed. Since
design basis accident initiators are not being altered by adoption of
the AST, the probability of an accident previously evaluated is not
affected.
Administrative changes to delete a footnote from Technical
Specification surveillance requirement 4.7.7.e.3) and a note from
ACTION 20 of Technical Specification Table 3.3-3, in which the
provisions of the notes have expired, does not impact the probability
or consequences of an accident previously evaluated.
Based on the above discussion, the proposed changes do not involve
a significant increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes do not create the possibility of a new or
different kind of accident from any accident previously evaluated.
The proposed changes do not involve a physical change. The change
will allow the automatic start feature of systems no longer credited in
the accident analyses for mitigation to be disabled through the STPNOC
[STP Nuclear Operating Company]
[[Page 41789]]
modification process. Implementation of AST provides increased
operating margins for filtration system efficiencies. Application of
AST provides for relaxation of certain Control Room ventilation system
filtration requirements. The Fuel Handling Building filtration and
holdup is no longer credited in the AST analyses. Therefore, the Fuel
Handling Building Exhaust Air Ventilation system is no longer required
in the Technical Specifications. It also relaxes containment integrity
requirements while handling irradiated fuel and during core
alterations. Elimination of the requirement to suspend operations
involving positive reactivity additions that could result in loss of
required SHUTDOWN MARGIN or required boron concentration if the control
room ventilation system is inoperable in Mode 5 or Mode 6 does not
create the possibility of a new or different kind of accident because
these events have already been analyzed in the safety analysis with a
conclusion that adequate measures exist to prevent these events.
Similarly, the proposed changes do not require any physical changes
to any structures, systems or components involved in the mitigation of
any accidents. Therefore, 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 changes.
Administrative changes to delete a footnote from Technical
Specification surveillance requirement 4.7.7.e.3) and a note from
ACTION 20 of Technical Specification Table 3.3-3, in which the
provisions of the notes have expired, does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Based on the above discussion, 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.
Approval of a change from the original source term methodology
(i.e., TID 14844) to an AST methodology, consistent with the guidance
in RG [NRC Regulatory Guide] 1.183, will not result in a significant
reduction in the margin of safety. The safety margins and analytical
conservatisms associated with the AST methodology have been evaluated
and were found acceptable. The results of the revised DBA analyses,
performed in support of the proposed changes, are subject to specific
acceptance criteria as specified in RG 1.183. The dose consequences of
these DBAs remain within the acceptance criteria presented in 10 CFR
50.67 and RG 1.183.
Elimination of the requirement to suspend operations involving
positive reactivity additions that could result in loss of required
SHUTDOWN MARGIN or required boron concentration if the control room
ventilation system is inoperable in Mode 5 or Mode 6 does not result in
a reduction in a margin to safety because adequate measures exist to
preclude radiological consequences from these events.
The proposed changes continue to ensure that the doses at the
exclusion area boundary (EAB) and low population zone boundary (LPZ),
as well as the Control Room and Technical Support Center, are within
the specified regulatory limits.
Administrative changes to delete a footnote from Technical
Specification surveillance requirement 4.7.7.e.3) and a note from
ACTION 20 of Technical Specification Table 3.3-3, in which the
provisions of the notes have expired, does not impact the margin of
safety.
Therefore, based on the above discussion, 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 standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Thomas G. Hiltz.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: May 21, 2007.
Description of amendment request: The license amendment request
proposes revising the Technical Specification (TS) Surveillance
Requirement (SR) 4.5.2.d for the inspection of Emergency Core Cooling
System (ECCS) sumps for consistency with the new STP sump design. SR
4.5.2.d includes a noncomprehensive parenthetical list of sump
components, some of which have been removed in the new sump screen
design. The licensee proposes an administrative change to delete the
parenthetical reference to sump components in its entirety.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
(1) Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is an administrative editorial change to remove
unnecessary information from a surveillance requirement. It will not
affect how any system, structure, or component is designed or operated
and so has no potential to affect the mitigation of an accident. The
change does not affect an initiator of any accident previously
evaluated. Therefore, the 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 is an administrative editorial change to remove
unnecessary information from a surveillance requirement. It will not
affect how any system, structure, or component is designed or operated
or involve any new or different plant configurations. Therefore, the
change does not create the possibility of a new or different kind of
accident previously evaluated.
(3) Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is editorial and administrative and
consequently has no effect on the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: A.H. Gutterman, Esq., Morgan, Lewis &
Bockius, 1111 Pennsylvania Avenue, NW., Washington, DC 20004.
NRC Branch Chief: Thomas G. Hiltz.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of amendment request: June 8, 2007.
Description of amendment request: The proposed amendment would
revise
[[Page 41790]]
the technical specifications for Watts Bar Nuclear Plant, Unit 1 (WBN)
to allow relaxations of various Reactor Trip System (RTS) and
Engineered Safety Feature Actuation System (ESFAS) logic completion
times, bypass test times, allowable outage times, and surveillance
testing intervals. The proposed changes implement several Technical
Specifications Task Force travelers, which the NRC staff has previously
reviewed and approved for incorporation into the Standard Technical
Specifications for Westinghouse plants.
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, which is
presented below:
1. Does the proposed change involve a significant increase in the
probability or consequences of an accident previously evaluated?
The proposed changes do not result in any modifications to RTS and
ESFAS hardware, design requirements, or functions. No system
operational parameters are affected. The protection system will
continue to perform the intended design functions consistent with the
design bases and accident analyses. The proposed changes will not
modify any system interfaces and, therefore, could not increase the
likelihood of an accident described in the UFSAR [Updated Facility
Safety Analysis Report]. The proposed amendment will not change,
degrade or prevent actions, or alter any assumptions previously made in
evaluating the radiological consequences of an accident described in
the UFSAR.
Plant-specific evaluations confirm the applicability of the
[Westinghouse Topical Report] WCAP-14333 and WCAP-15376 analyses to
WBN. Implementation of the approved changes is in accordance with the
conditions of the NRC safety evaluations for these reports and will
result in an insignificant risk impact.
The proposed changes to the completion time, bypass test time, and
surveillance frequencies reduce the potential for inadvertent reactor
trips and spurious actuations and, therefore, do not increase the
probability of any accident previously evaluated. The proposed changes
to the allowed completion time, bypass test time, and surveillance
frequencies do not change the response of the plant to any accidents
and have an insignificant impact on the reliability of the RTS and
ESFAS signals. The RTS and ESFAS will remain highly reliable and the
proposed changes will not result in a significant increase in the risk
of plant operation. This is demonstrated by showing that the impact on
plant safety as measured by core damage frequency [CDF] is less than
1.0E-06 per year and the impact on large early release frequency [LERF]
is less than 1.0E-07 per year. In addition, for the completion time
change, the incremental conditional core damage probabilities [ICCDP]
and incremental conditional large early release probabilities [ICLERP]
are less than 5.0E-07 and 5.0E-08, respectively. These changes meet the
acceptance criteria in Regulatory Guides 1.174 and 1.177. Therefore,
since the RTS and ESFAS will continue to perform their functions with
high reliability as originally assumed, and the increase in risk as
measured by CDF, LERF, ICCDP, and ICLERP is within the acceptance
criteria of existing regulatory guidance, there will not be a
significant increase in the consequences of any accidents.
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 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/public radiation exposures. The proposed changes are
consistent with the safety analysis assumptions and resultant
consequences.
Therefore, this change does not increase the probability or
consequences of an accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
The proposed amendment does not require any design changes,
physical modifications or changes in normal operation of the RTS and
ESFAS instrumentation. Existing setpoints will be maintained. The
changes do not affect functional performance requirements of the
instrumentation. No changes are required to accident analysis
assumptions. The changes do not introduce different malfunctions,
failure modes, or limiting single failures. The changes to the
completion time, bypass test time, and surveillance frequency do not
change any existing accident scenarios nor create any new or different
accident scenarios.
Therefore, this change does not create the possibility of a new or
different kind of accident from any previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
The proposed changes do 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. Redundant RTS and ESFAS trains are
maintained, and diversity with regard to the signals that provide
reactor trip and engineered safety features actuation is also
maintained. All signals credited as primary or secondary and all
operator actions credited in the accident analyses will remain the
same. The proposed changes will not result in plant operation in a
configuration outside the design basis. The calculated impact on risk
is insignificant and meets the acceptance criteria contained in
Regulatory Guides 1.174 and 1.177. Although there was no attempt to
quantify any positive human factors benefit due to increased completion
time, bypass test time, and surveillance frequencies, it is expected
that there would be a net benefit due to a reduced potential for
spurious reactor trips and actuations associated with testing.
Therefore, it is concluded that this change does not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: General Counsel, Tennessee Valley Authority,
400 West Summit Hill Drive, ET 11A, Knoxville, Tennessee 37902.
NRC Branch Chief: Thomas H. Boyce.
[[Page 41791]]
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.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: June 6, 2006.
Brief description of amendment request: The proposed amendments
would provide a new action for selected Technical Specifications (TSs)
limiting conditions for operation to permit extension of the completion
times of action requirements, provided risk is assessed and managed. A
new program, the Configuration Risk Management Program, would be added
to the Administrative Controls of TSs.
Date of publication of individual notice in Federal Register: June
12, 2007.
Expiration date of individual notice: July 12, 2007.
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 51.22(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].
AmerGen Energy Company, LLC, Docket No. 50-461, Clinton Power Station,
Unit No. 1, DeWitt County, Illinois
Date of application for amendment: November 13, 2006.
Brief description of amendment: The proposed amendment revises the
technical specification (TS) testing frequency for the surveillance
requirement (SR) in TS 3.2.4, ``Control Rod Scram Times.''
Specifically, the proposed change would revise the frequency for SR
3.1.4.2, control rod scram time testing, from ``120 days cumulative
operation in MODE 1,'' to ``200 days cumulative operation in MODE 1.''
This operating license improvement was made available by the Nuclear
Regulatory Commission on August 23, 2004, as part of the consolidated
line item improvement process.
Date of issuance: July 5, 2007.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment No.: 177.
Facility Operating License No. NPF-62: The amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: April 10, 2007 (72 FR
17944). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 5, 2007.
No significant hazards consideration comments received: No.
Dominion Nuclear Connecticut, Inc., et al., Docket No. 50-423,
Millstone Power Station, Unit No. 3, New London County, Connecticut
Date of application for amendment: June 14, 2006, as supplemented
by letters dated November 27, 2006 and January 17, 2007.
Brief description of amendment: The amendment revised the technical
specifications (TSs) to allow a one-time change in the Appendix J, Type
A, Containment Integrated Leak Rate Test from the required 10 years to
15 years.
Date of issuance: June 29, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days from the date of issuance.
Amendment No. 239.
Facility Operating License No. NPF-49: Amendment revised the
technical specifications.
Date of initial notice in Federal Register: September 12, 2006 (71
FR 53717). The November 27, 2006 and January 17, 2007, letters provided
clarifying information that did not change the initial proposed no
significant hazards consideration determination or expand the
application beyond the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained in a
Safety Evaluation dated June 29, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Units 1 and 2, Montgomery County, Pennsylvania
Date of application for amendment: June 2, 2006.
Brief description of amendment: The amendment revises the Technical
Specifications (TSs) to incorporate revised requirements in Title 10 of
the Code of Federal Regulations (10 CFR), Part 20. Specifically, the
amendment revises the definitions for Members of the Public and
Unrestricted Area, adds a definition for Restricted Area, revises the
requirements for limitations on the concentrations of radioactive
material
[[Page 41792]]
released in liquid and gaseous effluents, and revises the references
for radioactive effluent control requirements.
Date of issuance: June 29, 2007.
Effective date: As of its date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 187 and 148.
Facility Operating License Nos. NPF-39 and NPF-85: This amendment
revised the license and Technical Specifications.
Date of initial notice in Federal Register: April 10, 2007 (72 FR
17949). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated June 29, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: June 8, 2006, as supplemented
by letter dated February 5, 2007.
Brief description of amendments: These amendments modify the
Technical Specifications by removing reference to ``the Banked Position
Withdrawal Sequence'' and replace it with ``the analyzed rod position
sequence.''
Date of issuance: June 29, 2007.
Effective date: As of the date of issuance, to be implemented
within 60 days.
Amendments Nos.: 260 and 264.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the License and Technical Specifications.
Date of initial notice in Federal Register: August 15, 2006 (71 FR
46934). The February 5, 2007, letter, provided clarifying information
that did not change the initial proposed no significant hazards
consideration determination or expand the application beyond the scope
of the original Federal Register notice. The Commission's related
evaluation of the amendments is contained in a Safety Evaluation dated
June 29, 2007.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Docket Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: July 14, 2006, as supplemented
by letter dated June 5, 2007.
Brief description of amendments: The proposed changes modified
Technical Specification (TS) requirements related to required end
states for TS action statements that are consistent with the NRC-
approved Revision 0 to Technical Specification Task Force (TSTF) Change
Traveler, TSTF-423, ``Risk Informed Modification to Selected Required
Action End States for BWR [boiling-water reactor] Plants.''
Date of issuance: July 12, 2007.
Effective date: As of the date of issuance, to be implemented
within 120 days.
Amendments Nos.: 261 and 265.
Renewed Facility Operating License Nos. DPR-44 and DPR-56: The
amendments revised the TSs.
Date of initial notice in Federal Register: December 19, 2006 (71
FR 75994). The letter dated June 5, 2007, provided clarifying
information that did not change the initial proposed no significant
hazards consideration determination or expand the application beyond
the scope of the original Federal Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 12, 2007.
No significant hazards consideration comments received: No.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit No. 1, Lake County, Ohio
Date of application for amendment: January 19, 2007.
Brief description of amendment: The amendment modifies the
technical specifications requirements for the diesel fuel oil program
by relocating references to specific standards for fuel oil testing to
licensee-controlled documents and adds alternate criteria to the
``clear and bright'' acceptance test for new fuel oil.
Date of issuance: July 12, 2007.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 146.
Facility Operating License No. NPF-58: This amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: April 10, 2007 (72 FR
17950). The Commission's related evaluation of the amendment is
contained in a Safety Evaluation dated July 12, 2007.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama
Date of amendment request: January 30, 2007, supplemented by your
letter dated April 11, 2007.
Brief description of amendment request: The amendments revise
Section 5 of the technical specifications to reflect the move to a site
vice president organizational structure for Joseph M. Farley Nuclear
Plant, Units 1 and 2.
Date of issuance: July 16, 2007.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 175, 168.
Renewed Facility Operating License Nos. NPF-2 and NPF-8: Amendments
revise the technical specifications.
Date of initial notice in Federal Register: February 13, 2007 (72
FR 6790). The supplement provided clarifying information that did not
change the scope of the application nor the initial proposed no
significant hazards consideration determination.
The Commission's related evaluation of the amendments is contained
in a safety evaluation dated July 16, 2007.
No significant hazards consideration comments received: No.
STP Nuclear Operating Company, Docket Nos. 50-498 and 50-499, South
Texas Project, Units 1 and 2, Matagorda County, Texas
Date of amendment request: August 2, 2004, as resubmitted on June
6, 2006, and supplemented by letters dated December 28, 2006, February
28, May 9, and May 17, 2007.
Brief description of amendments: The amendments provide for a new
action for selected Technical Specifications (TS) limiting conditions
for operation to permit extending the completion times allowed for
action requirements subject to the requirements that the risk is
assessed and managed. A new Configuration Risk Management Program is
added to the TS under Administrative Controls, as a risk assessment
tool.
Date of issuance: July 13, 2007.
Effective date: As of the date of issuance and shall be implemented
within 180 days of the date of issuance.
Amendment Nos.: Unit 1--179; Unit 2--166.
Facility Operating License Nos. NPF-76 and NPF-80: The amendments
revised the Facility Operating Licenses and Technical Specifications.
Date of initial notice in Federal Register: June 12, 2007 (72 FR
32332). The Commission's related evaluation of the amendments is
contained in a Safety Evaluation dated July 13, 2007.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 19th day of July 2007.
[[Page 41793]]
For the Nuclear Regulatory Commission
Catherine Haney,
Director, Division of Operating Reactor Licensing, Office of Nuclear
Reactor Regulation.
[FR Doc. E7-14350 Filed 7-30-07; 8:45 am]
BILLING CODE 7590-01-P