[Federal Register Volume 71, Number 167 (Tuesday, August 29, 2006)]
[Notices]
[Pages 51222-51236]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 06-7137]
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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
[[Page 51223]]
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 August 4, 2006 to August 17, 2006. The last
biweekly notice was published on August 15, 2006 (71 FR 46929).
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
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
[[Page 51224]]
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].
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: April 11, 2006.
Description of amendment request: The proposed amendment would
modify Technical Specification 5.6.5 ``Core Operating Limits Report
(COLR)'' to add two U.S. Nuclear Regulatory Commission-approved topical
reports to the COLR methodologies list. These topical reports allow the
use of S-RELAP5 thermal-hydraulic analysis code for accident safety
analyses.
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 two topical reports have been reviewed and approved by the
NRC for use in determining core operating limits. The core operating
limits to be developed using the new methodologies for HBRSEP [H. B.
Robinson Steam Electric Plant], Unit No. 2, will be established in
accordance with the applicable limitations as documented in the NRC
Safety Evaluation Reports. In a May 11, 2001, NRC Safety Evaluation
Report, the NRC concluded that the S-RELAP5 code is capable of
addressing the thermal-hydraulic response of the target non-LOCA
[loss-of-coolant accident] events in a conservative manner and is,
therefore, an acceptable replacement for the ANF-RELAP code. In the
May 19, 2004, Safety Evaluation Report for Revision 1 to EMF-
2310(P)(A), the NRC concluded that the code remained acceptable for
use for the non-LOCA events. In a March 15, 2001, Safety Evaluation
Report, the NRC concluded that the code was acceptable for use for
small break LOCA analyses at Westinghouse pressurized water
reactors.
The proposed change, by itself, does not impact the current
design bases. The proposed change enables the use of new
methodologies to re-analyze certain events. Revised analyses may
either result in continued conformance with design bases, or may
change the design bases. If design basis changes result from a
revised analysis, then the specific design changes will be evaluated
in accordance with HBRSEP, Unit No. 2, design change procedures and
10 CFR 50.59.
The proposed change does not involve physical changes to any
plant structure, system, or component. Therefore, the probability of
occurrence for a previously analyzed accident is not significantly
increased.
The consequences of a previously analyzed accident are dependent
on the initial conditions assumed for the analysis, the behavior of
the fission product barriers during the analyzed accident, the
availability and successful functioning of the equipment assumed to
operate in response to the analyzed event, and the setpoints at
which these actions are initiated. The proposed methodologies will
ensure that the plant continues to meet applicable design and safety
analyses acceptance criteria. The proposed change does not affect
the performance of any equipment used to mitigate the consequences
of an analyzed accident. As a result, no analysis assumptions are
impacted and there are no adverse effects on the factors that
contribute to offsite or onsite dose as a result of an accident. The
proposed change does not affect setpoints that initiate protective
or mitigative actions. The proposed change ensures that plant
structures, systems, and components are maintained consistent with
the safety analysis and licensing bases. Based on this evaluation,
there is no significant increase in the consequences of a previously
analyzed accident.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. The Proposed Change Does Not Create the Possibility of a New
or Different Kind of Accident From Any Previously Evaluated.
The proposed change does not involve any physical alteration of
plant systems, structures, or components, other than allowing for
fuel design in accordance with NRC-approved methodologies. No new or
different equipment is being installed. No installed equipment is
being operated in a different manner. There is no change to the
parameters within which the plant is normally operated or in the
setpoints that initiate protective or mitigative actions. As a
result, no new failure modes are being introduced. Therefore, the
proposed change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. The Proposed Change Does Not Involve a Significant Reduction
in the Margin of Safety.
There is no impact on any margin of safety resulting from the
incorporation of these new topical reports into the Technical
[[Page 51225]]
Specifications. If design basis changes result from a revised
analysis that uses these new methodologies, the specific design
changes will be evaluated in accordance with HBRSEP, Unit No. 2,
design change procedures and 10 CFR 50.59. Any potential reduction
in the margin of safety would be evaluated for that specific design
change.
Therefore, 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: 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: L. Raghavan.
Carolina Power & Light Company, Docket No. 50-261, H. B. Robinson Steam
Electric Plant, Unit No. 2, Darlington County, South Carolina
Date of amendment request: July 17, 2006.
Description of amendment request: The proposed amendment would
revise the containment pressure requirements specified in Surveillance
Requirements 3.6.8 and 5.5.16 due to a revision in the Loss-of-Coolant
Accident (LOCA) containment pressure analysis.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated. The revised post-LOCA containment pressure and
temperature analysis used more conservative assumptions and the
increase in the calculated peak pressure was approximately 1 psig.
The revised value of 41.49 psig remains less than the containment
design pressure of 42 psig. The increase in the calculated peak
temperature was approximately 2 [deg]F, which was analyzed to have
no impact on structures or equipment. Although there is an increase
in the calculated pressure, the allowable containment leakage rate,
as measured at the peak pressure, is not being changed. Since there
is no increase in the allowable leakage, there is no increase in
consequences. The proposed change is related to containment pressure
analysis. There are no physical changes being made to the plant, or
to the manner in which the plant is operated. Surveillance
procedures for containment leakage have been conservatively testing
at pressures in excess of 42 psig and surveillance procedures for
the Isolation Valve Seal Water System have been conservatively
testing at pressures in excess of 46.2 psig. The change can have no
impact on the probability of an accident occurring. Therefore, the
proposed change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any previously evaluated?
No. The proposed change does not create the possibility of a new
or different kind of accident from any previously evaluated. There
are no physical changes being made to the plant or to the manner in
which the plant is operated. Surveillance procedures for containment
leakage have been conservatively testing at pressures in excess of
42 psig and surveillance procedures for the Isolation Valve Seal
Water System have been conservatively testing at pressures in excess
of 46.2 psig. The revised containment analysis results in a
calculated peak containment pressure that remains less than the
containment design pressure. The increase in the calculated peak
temperature was analyzed to have no impact on structures or
equipment. Therefore, this change does not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Do the proposed changes involve a significant reduction in
the margin of safety?
No. The proposed change does not involve a significant reduction
in the margin of safety. The proposed change imposes more
conservative surveillance test requirements. The calculated increase
in post-LOCA peak containment pressure is only 1 psig and the
revised value of 41.49 psig remains less than the containment design
pressure of 42 psig. The increase in the calculated peak temperature
was approximately 2 [deg]F, which was analyzed to have no impact on
structures or equipment. Although there was an increase in the
calculated pressure, the allowable containment leakage rate, as
measured at the peak pressure, is not being changed. Therefore, this
change does not involve a significant reduction in any 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: L. Raghavan.
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of amendment request: July 12, 2006.
Description of amendment request: The proposed amendment would
modify Conditions, Required Actions and Completion Times associated
with the inoperability of one or more emergency diesel generators
(EDGs) in Technical Specification (TS) 3.8.1.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
This license amendment request proposes a change to extend the
Technical Specification 3.8.1, ``AC Sources-Operating,'' Completion
Time. This change allows a single EDG to be inoperable for 7 days
more than Technical Specification 3.8.1 currently provides. The
Required Actions for CTG [Combustion Turbine Generator] 11-1 are
also removed from Condition A and TSTF [Technical Specification Task
Force] -439 is implemented for TS 3.8.1, removing the second
Completion Times.
The EDGs are safety related components which provide backup
electrical power supply to the onsite ESF [Engineered Safety
Feature] power distribution system. CTG 11-1 provides backup
electrical power to the Division 1 power distribution system.
Neither the EDGs nor CTG 11-1 are accident initiators, thus these
changes do not increase the probability of a previously evaluated
accident.
The plant ESF power distribution systems consist of two
divisions for 100% redundancy. Accident analyses demonstrate that
only one division is required for accident mitigation. Thus, with
one division inoperable the other division is capable of performing
the required safety function. Design basis analyses are not required
to be performed assuming extended loss of all power supplies to the
plant ESF power distribution system. Thus, this change does not
involve a significant increase in the consequences of a previously
analyzed accident.
The proposed change also eliminates the second Completion Time
from TS 3.8.1. These second Completion Times are not an initiator to
any accident previously evaluated. As a result, the probability of
an accident previously evaluated is not affected. The consequences
of an accident during the revised Completion Times are no different
than the consequences of the same accident during the existing
Completion Times. As a result, the consequences of an accident
previously evaluated are not affected by this change. Therefore, the
proposed change does not involve a significant increase in the
[[Page 51226]]
probability or consequences of an accident previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The changes do not involve a physical alteration of the plant
(i.e., no new or different type of equipment will be installed) or a
change in the methods governing normal plant operation. The changes
do not alter any 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 accident previously
evaluated.
3. The change does not involve a significant reduction in the
margin of safety.
This license amendment request proposes Technical Specification
changes to extend the Technical Specification 3.8.1, ``AC Sources-
Operating,'' Completion Time for an inoperable EDG to 14 days. These
changes allow an emergency diesel generator to be inoperable for 7
days more than TS 3.8.1 currently provides.
Deterministic and probabilistic risk assessments evaluated the
effect of the proposed TS changes on the availability of an
electrical power supply to the plant emergency safeguards features
systems. These assessments concluded that the proposed TS changes do
not involve a significant increase in the risk of power supply
unavailability.
This license amendment request proposes TS changes to remove the
Required Actions for CTG 11-1 from TS 3.8.1 Condition A. If CTG 11-1
is inoperable at the same time that any single EDG is inoperable for
the entire proposed 14 day period with no other equipment in
maintenance, the risk remains within RG [Regulatory Guide] 1.174
(Reference 2 [in the application]) thresholds for a ``very small''
classification.
The proposed change to delete the second Completion Time 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 affected by this change.
The proposed changes will not result in plant operation in a
configuration outside of the design basis.
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: David G. Pettinari, Legal Department, 688
WCB, Detroit Edison Company, 2000 2nd Avenue, Detroit, Michigan 48226-
1279.
NRC Acting Branch Chief: Martin Murphy.
Dominion Nuclear Connecticut, Inc., Docket No. 50-336, Millstone Power
Station, Unit No. 2, (MPS2) New London County, Connecticut
Date of amendment request: June 13, 2006.
Description of amendment request: The proposed amendment would
allow changes to the Technical Specifications based on the radiological
dose analysis margins obtained by using an alternate source term
consistent with 10 CFR 50.67.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated?
No.
The proposed amendment does not involve a significant increase
in the probability or consequence of an accident previously
analyzed. The MPS2 control room emergency ventilation system only
functions following the initiation of a design basis radiological
accident. Therefore, the change to the value used for in-leakage
rate test acceptance criteria following a design basis accident will
not increase the probability of any previously analyzed accident.
The proposed 200 cfm control room habitability envelope inleakage
surveillance acceptance criteria has no adverse impact on control
room habitability analyses for postulated toxic chemical release
events. These habitability analyses do not credit automatic or
manual isolation of the control room fresh air ventilation flow
during a toxic chemical release event. The control room's forced
ventilation fresh air exchange rate (e.g., 800 cfm) is much greater
than the proposed 200 cfm envelope inleakage rate acceptance
criteria. The MPS2 containment purge valve isolation signal is not
credited in the accident analyses. The requirements contained in
this specification do not meet any of 10 CFR 50.36(c)(2)(ii)
criteria on items for which technical specifications must be
established. Deletion of this technical specification will not
increase the probability of any previously analyzed accident. The
MPS2 containment and the containment systems function to prevent or
control the release of radioactive fission products following a
postulated accident. Therefore, the change to the value used for
primary to secondary leak rate acceptance criteria, and for all
secondary containment bypass leakage paths following a design basis
accident, will not increase the probability of any previously
analyzed accident.
These systems are not initiators of any design bases accident.
Revised dose calculations, which take into account the changes
proposed by this amendment and the use of the alternative source
term, have been performed for the MPS2 design basis radiological
accidents. The results of these revised calculations indicate that
public and control room doses will not exceed the limits specified
in 10 CFR 50.67 and Regulatory Guide 1.183. There is not a
significant increase in predicted dose consequences for any of the
analyzed accidents. Therefore, the proposed changes do not involve a
significant increase in the consequences of any previously analyzed
accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated?
No.
The implementation of the proposed changes does not create the
possibility of an accident of a different type than was previously
evaluated in the UFSAR. Although the proposed changes could affect
the operation of the control room emergency ventilation system, and
the containment and containment systems following a design basis
radiological accident, none of these changes can initiate a new or
different kind of accident since they are only related to system
capabilities that provide protection from accidents that have
already occurred. 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 do not create the
possibility of a new or different kind of accident from those
previously analyzed.
3. Involve a significant reduction in the margin of safety?
No.
The implementation of the proposed changes does not reduce the
margin of safety. The proposed changes for the control room
ventilation system, and the containment and containment systems do
not affect the ability of these systems to perform their intended
safety functions to maintain dose less than the required limits
during design basis radiological events. The radiological analysis
results, when compared with the revised TEDE [total effective dose
equivalent] acceptance criteria, meet the applicable limits. These
acceptance criteria have been developed for application to analyses
performed with alternative source terms. These acceptance criteria
have been developed for the purpose of use in design basis accident
analyses such that meeting the stated limits demonstrates adequate
protection of public health and safety. It is thus concluded that
the margin of safety will not be reduced by the implementation of
the changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Acting Branch Chief: Brooke D. Poole.
[[Page 51227]]
Dominion Nuclear Connecticut, Inc., Docket No. 50-423, Millstone Power
Station, Unit No. 3, New London County, Connecticut
Date of amendment request: July 19, 2006.
Description of amendment request: The proposed amendment will
revise reactor core safety limits Technical Specifications (TSs) and
relocate the reactor core safety limit figure to the Core Operating
Limits Report (COLR) in the Millstone Power Station, Unit 3, Technical
Requirements Manual.
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:
Criterion 1:
Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The relocation of cycle-specific core operating limits from the
TS to the COLR and the addition of the RCS [reactor coolant system]
total flow rate to the COLR has no influence or impact on the
probability or consequences of a design basis accident. Adherence to
the COLR and methodologies acceptable for establishing COLR
parameters continues to be controlled by TS. The proposed amendment
still requires exactly the same actions to be taken when or if
limits are exceeded. Each accident analysis addressed in the final
safety analysis report (FSAR) will be examined with respect to
changes in cycle-dependent parameters, which are obtained from
application of the NRC-approved reload design methodologies, to
ensure that the transient evaluation of new core designs are bounded
by previously accepted analyses. This examination, which will be
performed in accordance with the requirements of 10 CFR 50.59,
ensures that future core designs will not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Criterion 2:
Does the proposed amendment create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The relocation and addition of the cycle-specific variables to
the COLR and adding new document references in TS Section 6.9.1.6.b
does not influence or impact, nor does it contribute in any way to
the probability or consequences of an accident. No safety-related
equipment, safety function, or plant operations will be altered as a
result of these proposed changes. The cycle-specific variables are
calculated using NRC-approved methods and submitted to the NRC to
allow the staff to continue to trend the values of these limits. The
TS will continue to require operation within the required core
operating limits and appropriate actions will be taken when or if
limits are exceeded. Therefore the proposed amendment does not in
any way create the possibility of a new or different kind of
accident from any accident previously evaluated.
Criterion 3:
Does the proposed amendment involve a significant reduction in a
margin of safety?
Response: No.
The proposed changes have no impact on plant equipment
operation. The proposed changes do not revise any setpoints or
acceptance criteria assumed in the accident analyses. Therefore, the
proposed changes will not result in a reduction in a margin of
safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Lillian M. Cuoco, Senior Nuclear Counsel,
Dominion Nuclear Connecticut, Inc., Rope Ferry Road, Waterford, CT
06385.
NRC Acting Branch Chief: Brooke D. Poole.
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2 (IP2), Westchester County, New York
Date of amendment request: July 10, 2006.
Description of amendment request: Energy Nuclear Operations, Inc.,
is planning to operate an Independent Spent Fuel Storage Installation
(ISFSI) facility at IP2 using the HOLTEC HI-STORM 100 Cask System. To
support this activity, the proposed amendment adds Spent Fuel Cask
loading requirements to IP2 Technical Specifications (TSs).
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 will revise the Indian Point Unit 2 TSs
associated with the SFP [spent fuel pool] to assure that the
regulatory requirements related to criticality in the SFP and
applied to the Holtec HI-STORM 100 Multi-Purpose Canister MPC-32
when in the SFP are reflected in the IP2 TS. The proposed change
does not require any physical changes to Part 50 structures,
systems, or components, nor will their performance requirements be
altered. Therefore, the response of the plant to previously analyzed
accidents and related radiological releases will not be adversely
impacted, and will bound those postulated during cask loading
activities in the cask storage area.
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.
Existing fuel handling procedures and associated administrative
controls remain applicable for cask loading operations within the
SFP. Additionally, the soluble boron concentration required to
maintain keff <= 0.95 for postulated criticality accidents
associated with cask loading operations was also evaluated. The
results of the analyses, using a methodology previously approved by
the NRC [Nuclear Regulatory Commission], demonstrate that the amount
of soluble boron required to compensate for the positive reactivity
associated with these postulated accidents (371 ppm [parts per
million]) remains well below the existing spent fuel pit minimum
boron concentration limit of 2000 ppm. Accordingly, the same limit
has been proposed for cask loading operations in the cask storage
area.
Therefore, the possibility of a new or different kind of
accident from any accident previously evaluated is not created.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
An NRC approved methodology was used to perform the criticality
analysis which provides the basis to incorporate a new family of
burnup versus enrichment curves, for various cooling times, into the
plant Technical Specifications to ensure criticality requirements
are met during spent fuel cask loading. Accordingly, the existing
minimum boron concentration limit for the spent fuel pit of 2000 ppm
will continue to remain bounding during cask loading operations.
This determination accounts for uncertainties at a 95 percent
probability, 95 percent confidence level. Should it be postulated
that a boron dilution event does occur during this time period, keff
will remain less than 1.0 should the cask storage area become fully
flooded with unborated water.
Therefore, there will not be 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. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Richard J. Laufer.
[[Page 51228]]
Exelon Generation Company, LLC, Docket No. 50-249, Dresden Nuclear
Power Station (DNPS), Unit 3, Grundy County, Illinois
Date of amendment request: July 21, 2006.
Description of amendment request: The proposed amendment would
revise the values of the safety limit minimum critical power ratio
(SLMCPR) in Technical Specification Section 2.1.1, ``Reactor Core SLs
[Safety Limits].'' Specifically, the proposed change would require that
for Unit 3, the minimum critical power ratio (MCPR) for Global Nuclear
Fuel fuel shall be >= 1.10 for two recirculation loop operation, or >=
1.11 for single recirculation loop operation. Additionally, the
proposed change would require that MCPR for Westinghouse fuel shall be
>= 1.12 for two recirculation loop operation, or >= 1.14 for single
recirculation loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. 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
[Nuclear Regulatory Commission]-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 3, Cycle 20 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 3, Cycle 20.
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 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, Assistant General
Counsel, Exelon Generation Company, LLC, 200 Exelon Way, Kennett
Square, PA 19348.
NRC Branch Chief: Daniel S. Collins.
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-440, Perry
Nuclear Power Plant, Unit 1 (PNPP), Lake County, Ohio
Date of amendment request: June 6, 2006.
Description of amendment request: The proposed amendment would
revise the Ventilation Filter Test Program (VFTP) in Technical
Specification (TS) 5.5.7. The license amendment is a corrective action
to revise the flow rate units specified in the VFTP from standard cubic
feet per minute to cubic feet per minute. This amendment will ensure
the PNPP TS are consistent with plant design documentation, testing
criteria, and the industry.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response:
The ESF [Engineered Safety Feature] Ventilation systems reduce
the concentration of airborne radioactive contaminants following a
design basis accident and therefore are not initiators of design
bases accidents. The proposed amendment does not change the manner
in which the ESF ventilation systems are operated or tested.
Implementation of the proposed amendment will ensure the ESF
ventilation systems perform their function when called upon and does
not affect the plant operations, design function or analysis that
verifies the capability of a [plant] structures, systems or
components.
The proposed amendment does not affect the design of the ESF
ventilation systems, the operational characteristics of the ESF
ventilation systems, the interfaces between the ESF ventilation
systems and those plant systems they support, or the reliability of
the ESF ventilation systems.
Therefore, the ESF ventilation systems will be capable of
performing their accident mitigation function and there is no
increase in the probability or consequences of an accident already
evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response:
The proposed amendment introduces no new mode of plant operation
and does not involve a physical modification to the plant. New
equipment is not installed with the proposed amendment, nor does the
proposed amendment cause existing equipment to be operated in a new
or different manner.
Since the proposed changes do not involve a change to the plant
design or operation, no new system interactions are created by this
change. The proposed amendment does not produce any parameters or
conditions that
[[Page 51229]]
could contribute to the initiation of accidents different from those
already evaluated in the Updated Safety Analysis Report.
The changes to the VFTP do not affect the assumed accident
performance of the ESF Ventilation systems, nor [sic] any plant
structure, system or component previously evaluated.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response:
The proposed amendment does not impact the ESF ventilation
systems performance, including the capability for each ESF
ventilation system to attain and maintain required air flow assumed
in the plant safety analysis.
The proposed amendment does not involve a significant reduction
in a margin of safety since the operability of the ESF ventilations
[sic] systems continues to be determined as required to support the
capability of the ESF ventilations [sic] systems to provide the
required ventilation, filtration and temperature control to mitigate
the consequences of an accident.
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 W. Jenkins, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Branch Chief: Daniel S. Collins.
Florida Power Corporation, et al., Docket No. 50-302, Crystal River
Unit 3 Nuclear Generating Plant, Citrus County, Florida
Date of amendment request: May 25, 2006.
Description of amendment request: The proposed license amendment
revises the requirements in the Crystal River Unit 3 Improved Technical
Specification related to steam generator tube integrity. The licensee
states that the changes are consistent with NRC-approved Technical
Specification (TS) Task Force (TSTF) Change Traveler, TSTF-449, ``Steam
Generator Tube Integrity,'' Revision 4. The availability of this
technical specification improvement was announced in the Federal
Register on May 6, 2005, as part of the consolidated line item
improvement process.
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 requires a SG [Steam Generator] Program that
includes performance criteria that will provide reasonable assurance
that the SG tubing will retain integrity over the full range of
operating conditions (including startup, operation in the power
range, hot standby, cooldown and all anticipated transients included
in the design specification). The SG performance criteria are based
on tube structural integrity, accident induced leakage, and
operational LEAKAGE.
A SGTR [steam generarator tube rupture] event is one of the
design basis accidents that are analyzed as part of a plant's
licensing basis. In the analysis of a SGTR event, a bounding primary
to secondary LEAKAGE rate equal to the operational LEAKAGE rate
limits in the licensing basis plus the LEAKAGE rate associated with
a double-ended rupture of a single tube is assumed.
For other design basis accidents such as MSLB [main steamline
break], rod ejection, and reactor coolant pump locked rotor the
tubes are assumed to retain their structural integrity (i.e., they
are assumed not to rupture). These analyses typically assume that
primary to secondary LEAKAGE for all SGs is 1 gallon per minute or
increases to 1 gallon per minute as a result of accident induced
stresses. The accident induced leakage criterion introduced by the
proposed changes accounts for tubes that may leak during design
basis accidents. The accident induced leakage criterion limits this
leakage to no more than the value assumed in the accident analysis.
The SG performance criteria proposed change to the TS identify
the standards against which tube integrity is to be measured.
Meeting the performance criteria provides reasonable assurance that
the SG tubing will remain capable of fulfilling its specific safety
function of maintaining reactor coolant pressure boundary integrity
throughout each operating cycle and in the unlikely event of a
design basis accident. The performance criteria are only a part of
the SG Program required by the proposed change to the TS. The
program, defined by NEI [Nuclear Energy Institute] 97-06, Steam
Generator Program Guidelines, includes a framework that incorporates
a balance of prevention, inspection, evaluation, repair, and leakage
monitoring. The proposed changes do not, therefore, significantly
increase the probability of an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT 1-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
Therefore, limits are included in the plant technical specifications
for operational leakage and for DOSE EQUIVALENT 1-131 in primary
coolant to ensure the plant is operated within its analyzed
condition. The typical analysis of the limiting design basis
accident assumes that primary to secondary leak rate after the
accident is 1 gallon per minute with no more than [500 gallons per
day or 720 gallons per day] in any one SG, and that the reactor
coolant activity levels of DOSE EQUIVALENT 1-131 are at the TS
values before the accident.
The proposed change does not affect the design of the SGs, their
method of operation, or primary coolant chemistry controls. The
proposed approach updates the current TSs and enhances the
requirements for SG inspections. The proposed change does not
adversely impact any other previously evaluated design basis
accident and is an improvement over the current TSs.
Therefore, the proposed change does not affect the consequences
of a SGTR accident and the probability of such an accident is
reduced. In addition, the proposed changes do not affect the
consequences of an MSLB, rod ejection, or a reactor coolant pump
locked rotor event, or other previously evaluated accident.
Criterion 2--The Proposed Change Does Not Create the Possibility of a
New or Different Kind of Accident from any Previously Evaluated
The proposed performance based requirements are an improvement
over the requirements imposed by the current technical
specifications. Implementation of the proposed SG Program will not
introduce any adverse changes to the plant design basis or
postulated accidents resulting from potential tube degradation. The
result of the implementation of the SG Program will be an
enhancement of SG tube performance. Primary to secondary LEAKAGE
that may be experienced during all plant conditions will be
monitored to ensure it remains within current accident analysis
assumptions.
The proposed change does not affect the design of the SGs, their
method of operation, or primary or secondary coolant chemistry
controls. In addition, the proposed change does not impact any other
plant system or component. The change enhances SG inspection
requirements.
Therefore, the proposed change does not create the possibility
of a new or different type of accident from any accident previously
evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
The SG tubes in pressurized water reactors are an integral part
of the reactor coolant pressure boundary and, as such, are relied
upon to maintain the primary system's pressure and inventory. As
part of the reactor coolant pressure boundary, the SG tubes are
unique in that they are also relied upon as a heat transfer surface
between the primary and secondary systems such that residual heat
can be removed from the primary system. In addition, the SG tubes
isolate the radioactive fission products in the primary coolant from
the secondary system. In summary, the safety function of an SG is
maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change does not affect tube design or operating environment. The
proposed change is expected to result in an improvement in the
[[Page 51230]]
tube integrity by implementing the SG Program to manage SG tube
inspection, assessment, repair, and plugging. The requirements
established by the SG Program are consistent with those in the
applicable design codes and standards and are an improvement over
the requirements in the current TSs.
For the above reasons, the margin of safety is not changed and
overall plant safety will be enhanced by the proposed change to the
TS.
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: L. Raghavan.
Nuclear Management Company, LLC, Docket No. 50-266, Point Beach Nuclear
Plant (PBNP), Unit 1, Town of Two Creeks, Manitowoc County, Wisconsin
Date of amendment request: July 11, 2006.
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) 5.5.8, ``Steam Generator (SG)
Program,'' to exclude the portion of the tube below 17 inches from the
top of the tubesheet from the SG tube inspection requirements for Unit
1 on a one-time basis for a single operating cycle. In addition,
administrative changes are proposed to correct a page number in the TS
table of contents and delete two blank pages in TS Section 5.
Basis for proposed no significant hazards consideration
determination: As required by 10CFR 50.91(a), the licensee has provided
its analysis of the issue of no significant hazards consideration which
is presented below.
1. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant increase in
the probability or consequences of any accident previously
evaluated.
The proposed change revises Technical Specification (TS) 5.5.8,
``Steam Generator (SG) Program'' to redefine the PBNP Unit 1 primary
pressure boundary for purposes of the SG tube inspection
requirements on a one-time basis for Unit 1 Refueling Outage 30 and
the subsequent operating cycle. The redefined primary pressure
boundary is relocated from the seal weld at the bottom of the SG
tube to the tube-to-tubesheet mechanical interface. The required
structural integrity margins of the SG tubes in this area are
unaffected by this change and will be maintained by the SG
tubesheet. SG tubes are hydraulically expanded into the tubesheet.
Steam generator tube rupture is constrained by the tubesheet for
tubes with cracks in the tubesheet. This constraint results from the
hydraulic expansion process which restricts further expansion of the
tube, thermal expansion mismatch between the tube and tubesheet and
from the differential pressure between the primary and secondary
side. Thermal expansion and differential pressure also restrain the
tube axially. For conservatism, hydraulic preload was not factored
into the analysis.
The proposed change continues to require that the SG Program
include performance criteria that will provide reasonable assurance
that the SG tubing will retain integrity over the full range of
operating conditions (including startup, operation in the power
range, hot standby, cooldown and all anticipated transients included
in the design specification).
The SG performance criteria are based on tube structural
integrity, accident induced leakage, and operational LEAKAGE. The
analysis shows that structural integrity retains acceptable safety
factors against burst under normal steady state full power operation
primary-to-secondary pressure differential and against burst applied
to the design basis accident primary-to-secondary pressure
differentials. The analysis also shows that accident induced leakage
is bound by twice the normal operating leakage and well below the
accident analysis assumption for each stream generator. The primary
to secondary operational LEAKAGE limit is not changed.
The planned inspection and supporting analysis provide
reasonable assurance that the SG tubing will remain capable of
fulfilling its specific safety function of maintaining reactor
coolant pressure boundary integrity throughout the operating cycle
and in the unlikely event of a design basis accident. The proposed
change does not, therefore, significantly increase the probability
of an accident previously evaluated.
The consequences of design basis accidents are, in part,
functions of the DOSE EQUIVALENT I-131 in the primary coolant and
the primary to secondary LEAKAGE rates resulting from an accident.
The plant technical specification limits for operational LEAKAGE and
for DOSE EQUIVALENT I-131 in primary coolant, which ensure the plant
is operated within its analyzed condition, are unaffected by the
proposed change. Therefore, the proposed change does not
significantly increase the consequences of any accident previously
evaluated.
The proposed change does not significantly affect the
probability of any event initiators. There will be no change to
normal plant operating parameters, engineered safety feature
actuation setpoints, accident mitigation capabilities, or accident
analysis assumptions or inputs.
Therefore, the probability or consequences of any accident
previously evaluated will not be significantly increased as a result
of the proposed change.
2. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a new or different kind
of accident from any accident previously evaluated.
Implementation of the proposed change will not introduce any
adverse changes to the plant design basis or postulated accidents
resulting from potential tube degradation. Primary to secondary
leakage that may be experienced during all plant conditions will
continue to be monitored to ensure it remains within current
accident analysis assumptions. The proposed change does not affect
the method of operation of the SGs, or primary or secondary coolant
chemistry controls. In addition, the proposed change does not impact
any other plant system or component.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the proposed change. Equipment important to safety will continue
to operate as designed. The changes do not result in any event
previously deemed incredible being made credible. The changes do not
result in adverse conditions or result in any increase in the
challenges to safety systems. Therefore, the proposed change does
not create the possibility of a new or different type of accident
from any accident previously evaluated.
3. Operation of the Point Beach Nuclear Plant in accordance with
the proposed amendments does not result in a significant reduction
in a margin of safety.
The steam generators (SGs) are an integral part of the reactor
coolant pressure boundary and, as such, are relied upon to maintain
the primary system's pressure and inventory. They are also relied
upon to remove residual heat from the primary system. The safety
function of an SG is maintained by ensuring the integrity of its
tubes. Steam generator tube integrity is a function of the design,
environment, and the physical condition of the tube. The proposed
change redefines the PBNP Unit 1 primary pressure boundary from the
tube end weld to 17 inches below the top of the tubesheet and
incorporates revisions to the inspection criteria for SG tube
inspection in the tubesheet. The SG operating environment is not
affected by the change. The proposed change maintains the required
structural margins of the SG tubes for both normal and accident
conditions.
For cracking located within the tubesheet, steam generator tube
rupture is constrained by the tubesheet. For circumferentially
oriented cracking, the associated analysis for the proposed change
validates that 17 inches of degradation free expanded tubing
provides the necessary resistance to tube pullout with applicable
safety factors applied.
The revised inspection criteria continue to verify SG tube
integrity. The safety function of the affected components will be
maintained with the redefined primary pressure boundary.
There are no new or significant changes to the initial
conditions contributing to accident severity or consequences. The
proposed amendment will not otherwise affect the plant protective
boundaries, will not cause a release of fission products to the
public, nor will it degrade the performance of any other structures,
systems or components (SSCs) important to safety. Therefore, the
requested change will not result in 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
[[Page 51231]]
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 Acting Branch Chief: Martin Murphy.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna
Steam Electric Station, Units 1 and 2 (SSES 1 and 2), Luzerne
County, Pennsylvania
Date of amendment request: October 13, 2005.
Description of amendment request: The proposed amendment would
modify the licensing bases of SSES 1 and 2 by adopting the
Alternative Source Term (AST) methodology which replaces the current
accident source term with an AST. The AST is characterized by the
composition and magnitude of the radioactive material, the chemical
and physical form of the radionuclides, and the timing of the
releases of these radionuclides. The exceptions would be that the
current Technical Information Document (TID)14844 accident source
term would remain the licensing basis for (1) equipment
qualification, (2) NUREG-0737 evaluations other than Control Room
Habitability Envelope (CRHE) doses, and (3) Final Safety Analysis
Report (FSAR) accidents not included in Regulatory Guide 1.183.
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.
Adoption of the AST and pursuant TS [Technical Specification]
changes, changes to the TS's to address NRC Generic Letter 2003-01
(Reference 12.1) [see application dated October 13, 2005] and the
changes to the atmospheric dispersion factors, have no impact to the
initiation of DBAs [design basis accidents]. Once the occurrence of
an accident has been postulated, the new accident source term and
atmospheric dispersion factors are an input to analyses that
evaluate the radiological consequences. Some of the proposed changes
do affect the design or manner in which the facility is operated
following an accident; however, the proposed changes do not involve
a revision to the design or manner in which the facility is operated
that could increase in the probability of an accident previously
evaluated of a DBA discussed in Chapter 15 of the FSAR.
Therefore, the proposed change does not involve an increase in
the probability of an accident previously evaluated.
The structures, systems and components affected by the proposed
changes act as mitigators to the consequences of accidents. Based on
the revised analyses, the proposed changes do revise certain
performance requirements; however, 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
FSAR.
Plant-specific radiological analyses have been performed using
the AST methodology and new atmospheric dispersion factors. Based on
the results of these analyses, it had been demonstrated that the
CRHE dose consequences of the limiting events considered in the
analyses meet the regulatory guidance provided for use with the AST,
and the offsite doses are well within acceptable limits. This
guidance is presented in 10 CFR 50.67, RG [Regulatory Guide] 1.183,
and Standard Review Plan [SRP] Section 15.0.1.
Therefore, the proposed amendment does not result in a
significant increase in the consequences of any previously evaluated
accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Implementation of AST and the associated proposed TS changes and
new atmospheric dispersion factors do not alter or involve any
design basis accident initiators. These changes do not affect the
design function or mode of operation of structures, systems and
components in the facility prior to a postulated accident. Since
structures, systems and components are operated essentially no
differently after the AST implementation, no new failure modes are
created by this proposed change.
Licensing basis changes are proposed and justified to credit use
of the SLC [Standby Liquid Control] System to buffer suppression
pool pH to prevent iodine re-evolution following a postulated design
basis loss of coolant accident. There are new required manual
operator actions associated with the SLC System that are not
currently considered in the SSES design basis. Operator training
will be updated to reflect the new manual operator actions for the
pH control function of the of the SLC System as defined in the TS
Section 3.1.7. These changes are not significant because the
operators are already trained for the operation of the SLC System.
Procedural changes are mostly limited to the timing of SLC
initiation and termination. In addition, no new hardware changes are
necessary to use SLC in this new functional mode.
Licensing basis changes are proposed and justified for the
operation of the CREOASS [control room emergency outside air supply
system] to respond to NRC Generic Letter 2003-01 and TSTF [Technical
Specification Task Force] 448. No new hardware changes are necessary
to implement these changes. Since CREOASS will not be operated
differently as a result of these changes, no new failure modes are
created by these changes.
Therefore, the proposed license amendment will 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 results of the accident analyses revised in support of the
proposed change are subject to the acceptance criteria in 10 CFR
50.67. The analyzed events have been carefully selected, and the
analyses supporting these changes have been performed using approved
methodologies to ensure that analyzed events are bounding and safety
margin has not been reduced. The dose consequences of these limiting
events are within the acceptance criteria presented in 10 CFR 50.67,
RG 1.183, and SRP 15.0.1. Thus, by meeting the applicable regulatory
limits for AST, there is no significant reduction in a margin of
safety.
Changes to the SLC System to credit use of the Standby Liquid
Control (SLC) System to buffer suppression pool pH to prevent iodine
re-evolution and the CREOASS to address NRC Generic Letter 2003-01
and TSTF-448 [to] improve the margin of safety.
New offsite and Control Room atmospheric dispersion factors
(X/Qs) based on site specific meteorological data,
calculated in accordance with the guidance of RGs 1.145 and 1.194,
utilizes more recent data and improved calculational methodologies.
Therefore, because the proposed changes continue to result in
dose consequences within the applicable regulatory limits, the
changes are considered to not result in a significant reduction in a
margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Richard J. Laufer.
PSEG Nuclear LLC, Docket Nos. 50-272 and 50-311, Salem Nuclear
Generating Station, Unit Nos. 1 and 2, Salem County, New Jersey
Date of amendment request: June 7, 2006.
Description of amendment request: The proposed amendment would
delete Technical Specification (TS) 3/4.6.4, ``Combustible Gas
Control,'' TS 3/4.6.4.1, ``Hydrogen Analyzers,'' and TS 3/4.6.4.2,
``Electric Hydrogen Recombiners.'' The changes would be consistent with
NRC-approved TS Task Force (TSTF) Standard Technical Specification
(STS) Change Traveler number TSTF-447, Revision 1, ``Elimination of
Hydrogen Recombiners and Change to Hydrogen and Oxygen Monitors,'' as
part of the Consolidated Line Item Improvement Process (CLIIP).
The NRC staff issued a notice of availability of ``Model
Application Concerning Technical Specification Improvement To Eliminate
Hydrogen Recombiner Requirement, and Relax the Hydrogen and Oxygen
Monitor Requirements for Light Water Reactors
[[Page 51232]]
Using the Consolidated Line Item Improvement Process (CLIIP)'', in the
Federal Register on September 25, 2003 (68 FR 55416). The notice
included a model safety evaluation (SE), a model no significant hazards
consideration (NSHC) determination, and a model application.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee presented
an analysis of NSHC by endorsing the model NSHC determination for TSTF-
447 which was published in the Federal Register on September 25, 2003
(68 FR 55416) as follows:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The revised 10 CFR 50.44 no longer defines a design-basis loss-
of-coolant accident (LOCA) hydrogen release, and eliminates
requirements for hydrogen control systems to mitigate such a
release. The installation of hydrogen recombiners and/or vent and
purge systems required by 10 CFR 50.44(b)(3) was intended to address
the limited quantity and rate of hydrogen generation that was
postulated from a design-basis LOCA. The Commission has found that
this hydrogen release is not risk-significant because the design
basis LOCA hydrogen release does not contribute to the conditional
probability of a large release up to approximately 24 hours after
the onset of core damage. In addition, these systems were
ineffective at mitigating hydrogen releases from risk-significant
accident sequences that could threaten containment integrity.
With the elimination of the design basis LOCA hydrogen release,
hydrogen monitors are no longer required to mitigate design-basis
accidents and, therefore, the hydrogen monitors do not meet the
definition of a safety-related component as defined in 10 CFR 50.2.
RG [Regulatory Guide] 1.97 Category 1, is intended for key variables
that most directly indicate the accomplishment of a safety function
for design-basis accident events. The hydrogen monitors no longer
meet the definition of Category 1 in RG 1.97. As part of the
rulemaking to revise 10 CFR 50.44 the Commission found that Category
3, as defined in RG 1.97, is an appropriate categorization for the
hydrogen monitors because the monitors are required to diagnose the
course of beyond design-basis accidents.
The regulatory requirements for the hydrogen monitors can be
relaxed without degrading the plant emergency response. The
emergency response, in this sense, refers to the methodologies used
in ascertaining the condition of the reactor core, mitigating the
consequences of an accident, assessing and projecting offsite
releases of radioactivity, and establishing protective action
recommendations to be communicated to offsite authorities.
Classification of the hydrogen monitors as Category 3, and removal
of the hydrogen monitors from TS will not prevent an accident
management strategy through the use of the SAMGs [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.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c)
are satisfied. Therefore, the NRC staff proposes to determine that
the amendment request involves no significant hazards consideration.
Attorney for licensee: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Acting Branch Chief: Brooke D. Poole.
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
[[Page 51233]]
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].
Detroit Edison Company, Docket No. 50-341, Fermi 2, Monroe County,
Michigan
Date of application for amendment: January 31, 2006, as
supplemented by letter dated July 12, 2006.
Brief description of amendment: The amendment changed the technical
specifications to address issues related to an inconsistency that was
inadvertently introduced during conversion to improved technical
specifications when ``1 per room'' replaced ``2'' as the required
channels per trip system for the reactor water cleanup area ventilation
differential temperature--high isolation function.
Date of issuance: August 7, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days.
Amendment No.: 173.
Facility Operating License No. NPF-43: Amendment revised the
Technical Specifications, Surveillance Requirements, and License.
Date of initial notice in Federal Register: March 14, 2006 (71 FR
13171). The July 12, 2006, supplement contained clarifying information
and did not change the NRC staff's initial proposed finding of no
significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 7, 2006.
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: September 13, 2005.
Brief description of amendment: The amendment revised the Millstone
Power Station, Unit No. 3 Technical Specification (TSs) temperature
requirement for the reactivity control system rod drop time test.
Date of issuance: August 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment No.: 231.
Facility Operating License No. NPF-49: The amendment revised the
TSs.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61656).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 15, 2006.
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: May 24, 2005, revised by letter
dated May 2, 2006.
Brief description of amendment: The Technical Specification
amendment deleted the requirements for NRC approval of the engineering
evaluation justifying continued reactor operation with safety relief
valve (SRV) discharge pipe temperature exceeding the limit.
Date of issuance: August 4, 2006.
Effective date: As of the date of issuance, and shall be
implemented within 90 days.
Amendment No.: 222.
Facility Operating License No. DPR-35: The amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: August 16, 2005 (70 FR
48205).
The licensee originally requested for deletion of TS 3.6.D.3 in
their submittal dated May 24, 2005. The NRC staff found this
unacceptable. Therefore, the licensee revised the original application
by letter dated May 2, 2006, reducing the scope of the application as
originally noticed. Hence, there is no change to the NRC staff's
original proposed no significant hazards consideration determination.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 4, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: July 21, 2005, as supplemented on
February 15, May 3, and June 2, 2006.
Brief description of amendment: The amendment modified the
Waterford 3 Technical Specification (TS) to revise the existing steam
generator (SG) tube surveillance program to be consistent with TS Task
Force (TSTF) Change TSTF-449, ``Steam Generator Tube Integrity,''
Revision 4, and the model safety evaluation prepared by the NRC and
published in the Federal Register notice on March 2, 2005 (70 FR
10298). In this regard, the scope of the application includes changes
to the definition of leakage, changes to the primary-to-secondary
leakage requirements, changes to the SG tube surveillance program,
changes to the SG reporting requirements, and associated changes to the
TS Bases.
Date of issuance: July 31, 2006.
Effective date: As of the date of issuance and shall be implemented
90 days from the date of issuance.
Amendment No.: 204.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications and the Facility Operating License.
Date of initial notice in Federal Register: October 25, 2005 (70 FR
61659). The February 15, May 3, and June 2, 2006, supplemental letters
contained 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 finding of no significant
hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 31, 2006.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, St. Charles Parish, Louisiana
Date of amendment request: October 25, 2005.
Brief description of amendment: The amendment modified the
Waterford 3 Technical Specification 3.1.1.4, ``Minimum Temperature for
Criticality,'' to raise the minimum temperature for criticality from
the current value of >=520 [deg]F to >=533 [deg]F. Changes were also
proposed to the associated Action statement to reflect the increase in
temperature and to replace the current statement in Surveillance
Requirement 4.1.1.4 with wording consistent with NUREG-1432, ``Standard
Technical Specifications--Combustion Engineering Plants.''
Date of issuance: July 31, 2006.
Effective date: As of the date of issuance and shall be implemented
30 days from the date of issuance.
Amendment No.: 205.
Facility Operating License No. NPF-38: The amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: December 6, 2005 (70 FR
72672).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 31, 2006.
No significant hazards consideration comments received: No.
[[Page 51234]]
FirstEnergy Nuclear Operating Company, et al., Docket No. 50-346,
Davis-Besse Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of application for amendment: May 5, 2004, as supplemented by
letters dated January 17, October 10, and November 2, 2005 and May 30,
2006.
Brief description of amendment: This amendment revised the
technical specifications (TSs) for instrumentation setpoints, allowable
values, and calibration requirements based on updated calculations and
reviews, and add a definition of ``annual'' frequency for use in the
TSs.
Date of issuance: August 9, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 275.
Facility Operating License No. NPF-3: Amendment revised the
Technical Specifications and License.
Date of initial notice in Federal Register: June 8, 2004 (69 FR
32074). The January 17, October 10, and November 2, 2005 and May 30,
2006, supplements contained clarifying information and did not change
the NRC staff's initial proposed finding of no significant hazards
consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated August 9, 2006.
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: January 28, 2005, as
supplemented by letter dated January 12, 2006.
Brief description of amendments: The amendment revised Technical
Specifications (Tss) 1.1, ``Definitions,'' 3.4, ``Reactor Coolant
System [RCS],'' and 5.7, ``Reporting Requirements,'' to relocate the
RCS pressure-temperature curves and limits from the TSs to a licensee-
controlled document identified as the Pressure and Temperature Limit
Report.
Date of issuance: July 13, 2006.
Effective date: As of the date of issuance and shall be implemented
within 60 days of issuance.
Amendment Nos.: Unit 2-203; Unit 3-195.
Facility Operating License Nos. NPF-10 and NPF-15: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: March 1, 2005 (70 FR
9996). The January 12, 2006, supplemental letter 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.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 13, 2006.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-424 and 50-
425, Vogtle Electric Generating Plant, Units 1 and 2, Burke County,
Georgia
Date of application for amendments: September 19, 2005, as
supplemented by letter dated June 9, 2006.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) Limiting Conditions for Operation 3.3.1,
``Reactor Trip System Instrumentation,'' and TS Surveillance
Requirement (SR) 3.2.4.2, ``Quadrant Power Tilt Ratio (QPTR).'' The
amendments revise TS 3.3.1, Condition D and the note in SR 3.2.4.2 to
clarify when a flux map for QTPR is required.
Date of issuance: August 15, 2006.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 143 and 123.
Facility Operating License Nos. NPF-68 and NPF-81: Amendments
revised the licenses and the technical specifications.
Date of initial notice in Federal Register: March 14, 2005 (71 FR
13179). The supplement dated June 9, 2006, provided clarifying
information that did not change the scope of the September 19, 2005,
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 August 15, 2006.
No significant hazards consideration comments received: No.
TXU Generation Company LP, Docket Nos. 50-445 and 50-446, Comanche Peak
Steam Electric Station, Unit Nos. 1 and 2, Somervell County, Texas
Date of amendment request: August 10, 2005.
Brief description of amendments: The amendments revised the
Technical Specifications (TSs) to relocate diesel fuel oil testing
methods from TS 5.5.13 to a licensee-controlled document, provided
clarifications, and corrected the format and typographical errors.
Date of issuance: July 28, 2006.
Effective date: As of the date of issuance and shall be implemented
within 120 days from the date of issuance.
Amendment Nos.: 127 and 127.
Facility Operating License Nos. NPF-87 and NPF-89: The amendments
revised the Technical Specifications and the Facility Operating
Licenses.
Date of initial notice in Federal Register: November 8, 2005 (70 FR
67754).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 28, 2006.
No significant hazards consideration comments received: No.
Virginia Electric and Power Company, et al., Docket Nos. 50-280 and 50-
281, Surry Power Station, Units 1 and 2, Surry County, Virginia
Date of application for amendments: September 13, 2005, as
supplemented on April 7 and May 23, 2006.
Brief Description of amendments: These amendments revised Technical
Specification 5.1, ``Site,'' to redefine the exclusion area boundary as
the site boundary.
Date of issuance: August 10, 2006.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 249, 248.
Renewed Facility Operating License Nos. DPR-32 and DPR-37:
Amendments changed the licenses and the technical specifications.
Date of initial notice in Federal Register: January 3, 2006 (71 FR
156). The April 7 and May 23, 2006, supplements contained clarifying
information only and did not change the initial proposed no significant
hazards consideration determination or expand the scope of the initial
application.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated August 10, 2006.
No significant hazards consideration comments received: No.
Notice of Issuance of Amendments to Facility Operating Licenses and
Final Determination of No Significant Hazards Consideration and
Opportunity for a Hearing (Exigent Public Announcement or Emergency
Circumstances)
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application for the
amendment complies with the
[[Page 51235]]
standards and requirements of the Atomic Energy Act of 1954, as amended
(the Act), and the Commission's rules and regulations. The Commission
has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Because of exigent or emergency circumstances associated with the
date the amendment was needed, there was not time for the Commission to
publish, for public comment before issuance, its usual Notice of
Consideration of Issuance of Amendment, Proposed No Significant Hazards
Consideration Determination, and Opportunity for a Hearing.
For exigent circumstances, the Commission has either issued a
Federal Register notice providing opportunity for public comment or has
used local media to provide notice to the public in the area
surrounding a licensee's facility of the licensee's application and of
the Commission's proposed determination of no significant hazards
consideration. The Commission has provided a reasonable opportunity for
the public to comment, using its best efforts to make available to the
public means of communication for the public to respond quickly, and in
the case of telephone comments, the comments have been recorded or
transcribed as appropriate and the licensee has been informed of the
public comments.
In circumstances where failure to act in a timely way would have
resulted, for example, in derating or shutdown of a nuclear power plant
or in prevention of either resumption of operation or of increase in
power output up to the plant's licensed power level, the Commission may
not have had an opportunity to provide for public comment on its no
significant hazards consideration determination. In such case, the
license amendment has been issued without opportunity for comment. If
there has been some time for public comment but less than 30 days, the
Commission may provide an opportunity for public comment. If comments
have been requested, it is so stated. In either event, the State has
been consulted by telephone whenever possible.
Under its regulations, the Commission may issue and make an
amendment immediately effective, notwithstanding the pendency before it
of a request for a hearing from any person, in advance of the holding
and completion of any required hearing, where it has determined that no
significant hazards consideration is involved.
The Commission has applied the standards of 10 CFR 50.92 and has
made a final determination that the amendment involves no significant
hazards consideration. The basis for this determination is contained in
the documents related to this action. Accordingly, the amendments have
been issued and made effective as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
application for amendment, (2) the amendment to Facility Operating
License, and (3) the Commission's related letter, Safety Evaluation
and/or Environmental Assessment, as indicated. All of these items are
available for public inspection at the Commission's Public Document
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 System's (ADAMS) Public Electronic Reading Room
on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there are problems
in accessing the documents located in ADAMS, contact the PDR Reference
staff at 1 (800) 397-4209, (301) 415-4737 or by e-mail to [email protected].
The Commission is also offering an opportunity for a hearing with
respect to the issuance of the amendment. 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, and electronically on the Internet at the NRC Web site,
http://www.nrc.gov/reading-rm/doc-collections/cfr/. If there are
problems in accessing the document, contact the PDR Reference staff at
1 (800) 397-4209, (301) 415-4737, or by e-mail to [email protected]. If a
request for a hearing or petition for leave to intervene is filed by
the above date, the Commission or 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 identify 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
intends to rely in proving the contention at the hearing. The
petitioner must also provide references to those specific sources and
documents of which the petitioner is aware and on which the petitioner
intends to rely to establish those facts or expert opinion. The
petition must include sufficient information to show that a genuine
dispute exists with the applicant on a
[[Page 51236]]
material issue of law or fact.\1\ Contentions shall be limited to
matters within the scope of the amendment under consideration. The
contention must be one which, if proven, would entitle the petitioner
to relief. A petitioner/requestor who fails to satisfy these
requirements with respect to at least one contention will not be
permitted to participate as a party.
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\1\ To the extent that the applications contain attachments and
supporting documents that are not publicly available because they
are asserted to contain safeguards or proprietary information,
petitioners desiring access to this information should contact the
applicant or applicant's counsel and discuss the need for a
protective order.
---------------------------------------------------------------------------
Each contention shall be given a separate numeric or alpha
designation within one of the following groups:
1. Technical--primarily concerns/issues relating to technical and/
or health and safety matters discussed or referenced in the
applications.
2. Environmental--primarily concerns/issues relating to matters
discussed or referenced in the environmental analysis for the
applications.
3. Miscellaneous--does not fall into one of the categories outlined
above.
As specified in 10 CFR 2.309, if two or more petitioners/requestors
seek to co-sponsor a contention, the petitioners/requestors shall
jointly designate a representative who shall have the authority to act
for the petitioners/requestors with respect to that contention. If a
petitioner/requestor seeks to adopt the contention of another
sponsoring petitioner/requestor, the petitioner/requestor who seeks to
adopt the contention must either agree that the sponsoring petitioner/
requestor shall act as the representative with respect to that
contention, or jointly designate with the sponsoring petitioner/
requestor a representative who shall have the authority to act for the
petitioners/requestors with respect to that contention.
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. Since the Commission has made a final determination that the
amendment involves no significant hazards consideration, if a hearing
is requested, it will not stay the effectiveness of the amendment. Any
hearing held would take place while the amendment is in effect.
A request for a hearing or a petition for leave to intervene must
be filed 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 or 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).
Entergy Nuclear Operations, Inc., Docket No. 50-247, Indian Point
Nuclear Generating Unit No. 2, Westchester County, New York
Date of amendment request: July 26, 2006.
Description of amendment request: The amendment revised Function 6
[Containment Water Level (Containment Sump)] of Table 3.3.3-1 (``Post
Accident Monitoring Instrumentation''), referenced in the Technical
Specification (TS) Limiting Condition for Operation (LCO) 3.3.3, ``Post
Accident Monitoring Instrumentation.'' The revision changed Function 6
to specify 2 required channels for the Containment Sump water level
instrumentation instead of 3 channels.
Date of issuance: July 28, 2006.
Effective date: As of its date of issuance, and shall be
implemented prior to the expiration of the current 7-day allowed outage
time for inoperable containment sump water level channels, which was
entered on July 24, 2006.
Amendment No.: 249.
Facility Operating License No. DPR-26: The amendment revised the TS
and License.
Public comments requested as to proposed no significant hazards
consideration (NSHC): No. The Commission's related evaluation of the
amendment, finding of emergency circumstances, state consultation, and
final NSHC determination are contained in a Safety Evaluation dated
July 28, 2006.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Branch Chief: Richard J. Laufer.
Dated at Rockville, Maryland, this 21st day of August 2006.
For the Nuclear Regulatory Commission.
Cornelius F. Holden,
Deputy Director, Division of Operating Reactor Licensing, Office of
Nuclear Reactor Regulation.
[FR Doc. 06-7137 Filed 8-28-06; 8:45 am]
BILLING CODE 7590-01-P