[Federal Register Volume 68, Number 140 (Tuesday, July 22, 2003)]
[Notices]
[Pages 43382-43400]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-18084]
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NUCLEAR REGULATORY COMMISSION
Biweekly Notice; Applications and Amendments to Facility
Operating Licenses Involving No Significant Hazards Considerations
I. Background
Pursuant to Public Law 97-415, the U.S. Nuclear Regulatory
Commission (the Commission or NRC staff) is publishing this regular
biweekly notice. Public Law 97-415 revised section 189 of the Atomic
Energy Act of 1954, as amended (the Act), to require the Commission to
publish notice of any amendments issued, or proposed to be issued,
under a new provision of section 189 of the Act. This provision grants
the Commission the authority to issue and make immediately effective
any amendment to an operating license upon a determination by the
Commission that such amendment involves no significant hazards
consideration, notwithstanding the pendency before the Commission of a
request for a hearing from any person.
This biweekly notice includes all notices of amendments issued, or
proposed to be issued from, June 27, 2003, through July 10, 2003. The
last biweekly notice was published on July 8, 2003 (68 FR 40707).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve no significant hazards consideration. Under
the Commission's regulations in 10 CFR 50.92, this means that operation
of the facility in accordance with the proposed amendment would not:
(1) Involve a significant increase in the probability or consequences
of an accident previously evaluated; or (2) create the possibility of a
new or different kind of accident from any accident previously
evaluated; or (3) involve a significant reduction in a margin of
safety. The basis for this proposed determination for each amendment
request is shown below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of the 30-day notice period. However, should circumstances
change during the notice period such that failure to act in a timely
way would result, for example, in derating or shutdown of the facility,
the Commission may issue the license amendment before the expiration of
the 30-day notice period, provided that its final determination is that
the amendment involves no significant hazards consideration. The final
determination will consider all public and State comments received
before action is taken. Should the Commission take this action, it will
publish in the Federal Register a notice of issuance and provide for
opportunity for a hearing after issuance. The Commission expects that
the need to take this action will occur very infrequently.
Written comments may be submitted by mail to the Chief, Rules and
Directives Branch, Division of Administrative Services, Office of
Administration, U.S. Nuclear Regulatory Commission, Washington, DC
20555-0001, and should cite the publication date and page number of
this Federal Register notice. Written comments may also be delivered to
Room 6D22, Two White Flint North, 11545 Rockville Pike, Rockville,
Maryland, from 7:30 a.m. to 4:15 p.m. Federal workdays. Copies of
written comments received may be examined at the Commission's Public
Document Room (PDR), located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland. The
filing of requests for a hearing and petitions for leave to intervene
is discussed below.
By August 21, 2003, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and
[[Page 43383]]
any person whose interest may be affected by this proceeding and who
wishes to participate as a party in the proceeding must file a written
request for a hearing and a petition for leave to intervene. Requests
for a hearing and a petition for leave to intervene shall be filed in
accordance with the Commission's ``Rules of Practice for Domestic
Licensing Proceedings'' in 10 CFR part 2. Interested persons should
consult a current copy of 10 CFR 2.714, which is available at the
Commission's PDR, located at One White Flint North, Public File Area
01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/doc-collections/cfr/. If a request for a hearing or petition
for leave to intervene is filed by the above date, the Commission or an
Atomic Safety and Licensing Board, designated by the Commission or by
the Chairman of the Atomic Safety and Licensing Board Panel, will rule
on the request and/or petition; and the Secretary or the designated
Atomic Safety and Licensing Board will issue a notice of a hearing or
an appropriate order.
As required by 10 CFR 2.714, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following factors: (1) The nature of the petitioner's right under the
Act to be made a party to the proceeding; (2) the nature and extent of
the petitioner's property, financial, or other interest in the
proceeding; and (3) the possible effect of any order which may be
entered in the proceeding on the petitioner's interest. The petition
should also identify the specific aspect(s) of the subject matter of
the proceeding as to which petitioner wishes to intervene. Any person
who has filed a petition for leave to intervene or who has been
admitted as a party may amend the petition without requesting leave of
the Board up to 15 days prior to the first prehearing conference
scheduled in the proceeding, but such an amended petition must satisfy
the specificity requirements described above.
Not later than 15 days prior to the first prehearing conference
scheduled in the proceeding, a petitioner shall file a supplement to
the petition to intervene which must include a list of the contentions
which are sought to be litigated in the matter. Each contention must
consist of a specific statement of the issue of law or fact to be
raised or controverted. In addition, the petitioner shall provide a
brief explanation of the bases of the contention and a concise
statement of the alleged facts or expert opinion which support the
contention and on which the petitioner intends to rely in proving the
contention at the hearing. The petitioner must also provide references
to those specific sources and documents of which the petitioner is
aware and on which the petitioner intends to rely to establish those
facts or expert opinion. Petitioner must provide sufficient information
to show that a genuine dispute exists with the applicant on a material
issue of law or fact. Contentions shall be limited to matters within
the scope of the amendment under consideration. The contention must be
one which, if proven, would entitle the petitioner to relief. A
petitioner who fails to file such a supplement which satisfies these
requirements with respect to at least one contention will not be
permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing, including the opportunity to present evidence and cross-
examine witnesses.
If a hearing is requested, the Commission will make a final
determination on the issue of no significant hazards consideration. The
final determination will serve to decide when the hearing is held.
If the final determination is that the amendment request involves
no significant hazards consideration, the Commission may issue the
amendment and make it immediately effective, notwithstanding the
request for a hearing. Any hearing held would take place after issuance
of the amendment.
If the final determination is that the amendment request involves a
significant hazards consideration, any hearing held would take place
before the issuance of any amendment.
A request for a hearing or a petition for leave to intervene must
be filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff, or may be delivered to the Commission's PDR,
located at One White Flint North, Public File Area 01F21, 11555
Rockville Pike (first floor), Rockville, Maryland, by the above date.
Because of continuing disruptions in delivery of mail to United States
Government offices, it is requested that petitions for leave to
intervene and requests for hearing be transmitted to the Secretary of
the Commission either by means of facsimile transmission to 301-415-
1101 or by e-mail to [email protected]. A copy of the request for
hearing and petition for leave to intervene should also be sent to the
Office of the General Counsel, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and because of continuing disruptions in
delivery of mail to United States Government offices, it is requested
that copies be transmitted either by means of facsimile transmission to
301-415-3725 or by e-mail to [email protected]. A copy of the
request for hearing and petition for leave to intervene should also be
sent to the attorney for the licensee.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for a hearing will
not be entertained absent a determination by the Commission, the
presiding officer or the Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of
factors specified in 10 CFR 2.714(a)(1)(i)-(v) and 2.714(d).
For further details with respect to this action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the Agencywide
Documents Access and Management System's (ADAMS) Public Electronic
Reading Room on the Internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS or if there
are problems in accessing the documents located in ADAMS, contact the
NRC PDR Reference staff at 1-800-397-4209, 301-415-4737 or by e-mail to
[email protected].
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of amendment request: June 3, 2003.
Description of amendment request: Pursuant to title 10 of the Code
of Federal Regulations (10 CFR), Section 50.90, Duke Energy Corporation
requested an amendment to the McGuire Nuclear Station Facility
Operating Licenses and Technical Specifications (TS). The proposed
change would modify TS 3.6.14 to allow
[[Page 43384]]
a pressurizer hatch to be open for up to 6 hours, an increase from the
current TS limit of 1-hour. Conforming changes would also be made to
the associated Bases.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the change involve a significant increase in the
probability or consequences of an accident previously evaluated?
No. Implementation of this amendment will not involve a
significant increase in the probability or consequences of an
accident previously evaluated. Removal of the pressurizer enclosure
hatch will not cause an increase in the probability of an accident
which has been previously evaluated because the pressurizer
enclosure hatch is not an accident initiator.
The consequences of an accident which have been previously
evaluated will not be significantly increased by removal of the
pressurizer enclosure hatch. As discussed in the analysis contained
in the technical justification supporting this amendment request,
the new containment compression peak pressure will remain well below
the acceptance criteria. Additionally, the long term containment
peak pressure will not be adversely affected due to the delay time
in melting of the ice. The removal of the pressurizer enclosure
hatch itself has been previously evaluated in Modes 1 through 4 in
accordance with the analytical process described in NUREG-0612 and
the NRC's December 22, 1980 letter regarding the control of heavy
loads at nuclear plants. The changes proposed in this license
amendment request will have no adverse effect on the procedures used
for the handling of heavy loads (pressurizer enclosure hatch) at
McGuire nor on the generation of internal missiles as evaluated in
Section 3.5 of the McGuire Updated Final Safety Analysis Report.
2. Does the change create the possibility of a new or different
kind of accident from any accident previously evaluated?
No. Implementation of this amendment would not create the
possibility of a new or different kind of accident from any accident
previously evaluated. No new accident causal mechanisms are created
as a result of the NRC approval of this license amendment request.
As discussed above, extending the time that the pressurizer hatch is
allowed to be open does not create any new or different accidents
from those previously evaluated. Removal of the pressurizer
enclosure hatch to perform inspections or maintenance inside the
pressurizer cavity has been previously evaluated and determined to
be acceptable. The analysis contained in the technical justification
for this license amendment request provides results which conclude
that the containment compression peak pressure, and the long term
containment peak pressure are acceptable with the pressurizer
enclosure hatch open. This amendment does not impact any plant
systems that are accident initiators; therefore, no new accident
types are being created.
3. Does this change involve a significant reduction in a margin
of safety?
No. Implementation of this amendment would not involve a
significant reduction in a margin of safety. Margin of safety is
related to the confidence in the ability of the fission product
barriers to perform their design functions during and following an
accident situation. These barriers include the fuel cladding, the
reactor coolant system, and the containment system. The pressurizer
enclosure hatch and its performance have a direct impact on the
containment boundary, since peak containment pressure due to an
accident could be affected. However, the analysis supporting this
amendment request concludes that the containment compression peak
pressure and the long term containment peak pressure continue to be
acceptable with the increased open time for the hatch. Thus the
performance of the fission product barriers will not be
significantly impacted by implementation of this amendment and no
safety margin will be significantly impacted.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lisa F. Vaughn, Duke Energy Corporation,
422 South Church Street, Charlotte, North Carolina 28201-1006.
NRC Section Chief: John A. Nakoski.
Entergy Operations, Inc., Docket No. 50-313, Arkansas Nuclear One, Unit
No. 1, Pope County, Arkansas
Date of amendment request: June 30, 2003.
Description of amendment request: The proposed amendment would
revise the control room emergency ventilation system (CREVS)
surveillance requirement (SR) by modifying an existing SR related to
the makeup flow rate to show that it is applicable to the VSF-9 train
and by adding a new makeup flow rate SR that is applicable to the 2VSF-
9 train.
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 purpose of the CREVS is to provide airborne radiological
protection for operations from the control room for the design basis
loss of coolant accident fission product release and for a fuel
handling accident. The proposed change continues to assure that the
control room operator will be protected from the dose consequences
related to either of these accidents.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will establish appropriate outside air
makeup flow rates for the 2VSF-9 fan unit. This criterion has been
evaluated and determined to continue to provide protection to the
control room operator in accordance with General Design Criterion
19. The proposed change is not an accident initiator. No
modifications to the system are proposed which would create the
possibility of a new or different kind of accident.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change will establish the allowable makeup airflow
into the control room when the 2VSF-9 CREVS train is in operation.
Calculations have been performed which demonstrate that the proposed
flow criteria provides increased protection for the control room
operator.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: June 30, 2003.
Description of amendment request: The proposed amendment would: (1)
Eliminate credit for the Boraflex neutron absorbing material used for
reactivity control in Region 1 of the spent fuel pool (SFP), (2) credit
a combination of soluble boron and several defined fuel loading
patterns within the storage racks to maintain SFP reactivity within the
effective neutron multiplication factor (Keff) limits of 10
CFR 50.68, (3) increase
[[Page 43385]]
the minimum boron concentration in the SFP to 2000 parts per million
(ppm), and (4) reduce the fresh fuel assembly initial enrichment to
less than or equal to 4.55 +/- 0.05 weight percent uranium-235 (U-235).
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 fuel handling accidents described below can be postulated to
increase reactivity. However, for these accident conditions, the
double contingency principle of ANS [American Nuclear Society]
N16.1-1975 is applied. This states that it is unnecessary to assume
two unlikely, independent, concurrent events to ensure protection
against a criticality accident. Thus, for accident conditions, the
presence of soluble boron in the storage pool water can be assumed
as a realistic initial condition since its absence would be a second
unlikely event.
Three types of drop accidents have been considered: a vertical
drop accident, a horizontal drop accident, and an inadvertent drop
of an assembly between the outside periphery of the rack and the
pool wall.
[sbull] A vertical drop directly upon a cell will cause damage
to the racks in the active fuel region. The proposed 2000 ppm
soluble boron concentration will ensure that Keff does
not exceed 0.95.
[sbull] A fuel assembly dropped on top of the rack that comes to
rest horizontally will not deform the rack structure such that
criticality assumptions are invalidated. The rack structure is such
that an assembly positioned horizontally on top of the rack results
in a minimum separation distance from the upper end of the active
fuel region of the stored assemblies. This distance is sufficient to
preclude interaction between the dropped assembly and the stored
fuel.
[sbull] An inadvertent drop of an assembly between the outside
periphery of the rack and the pool wall is bounded by the worst case
fuel misplacement accident condition.
The fuel assembly misplacement accident was considered for all
storage configurations. An assembly with high reactivity is assumed
to be placed in a storage location which requires a fuel assembly
with a lower reactivity. The presence of soluble boron in the pool
water assumed in the analysis has been shown to offset the worst
case reactivity effect of a misplaced fuel assembly for any
configuration. This soluble boron requirement is less than the
proposed 2000 ppm that will be required by the ANO-2 [Arkansas
Nuclear One, Unit No. 2] TS [Technical Specifications]. Thus, a five
percent subcriticality margin can be easily met for postulated
accidents, since any reactivity increase will be much less that the
negative worth of the dissolved boron.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change will define several acceptable 2 x 2 loading
patterns and acceptable interfaces between the patterns. In
addition, the proposed change will credit soluble boron to assure a
five percent subcriticality margin is maintained during normal
conditions and in the event of a postulated accident. The soluble
boron concentration assumed in the analyses for a postulated
accident is less than the proposed TS change of 2000 ppm. Thus, a
five percent subcriticality margin can easily be met for postulated
accidents, since any reactivity increase will be much less than the
negative worth of the dissolved boron.
No new or different types of fuel assembly drop scenarios are
created by the proposed change. The presence of soluble boron in the
SFP water assures a subcriticality margin is maintained in the event
of fuel assembly misplacement.
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.
With the presence of a nominal boron concentration, the fuel
storage patterns are designed to assure that fuel assemblies of less
than or equal to 4.55 +/- 0.05 weight percent U-235 enrichment when
loaded in accordance with the proposed loading patterns will be
maintained within a subcritical array with a five percent
subcritical margin (95% probability at the 95% confidence level).
This has been verified by criticality analyses.
Credit for soluble boron in the SFP water is permitted under
accident conditions as well as in non-accident conditions.
Criticality analyses have been performed to determine the required
boron concentration that would ensure a subcriticality margin of at
least five percent. By increasing the minimum boron concentration to
greater than 2000 ppm, the margin of safety currently defined by
taking credit for soluble boron will be maintained.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502
NRC Section Chief: Robert A. Gramm
Entergy Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: June 30, 2003.
Description of amendment request: The proposed amendment would (1)
reorganize the Arkansas Nuclear One, Unit No. 2 (ANO-2) Technical
Specifications (TSs) Section 6.0, Administrative Controls, (2) modify
the ANO-2 Facility Operating License, and actions and surveillance
requirements (SRs) of various other TSs, to support the reorganization
of Section 6.0, and (3) modify several actions and SRs that are related
to systems that are shared by ANO-2 and Arkansas Nuclear One, Unit No.
1 (ANO-1). These changes are being proposed so that the philosophy and
location of the TSs in Section 6.0 reflect the recently approved
conversion of the ANO-1 TSs to the Improved Technical Specifications
(ITS) and the subsequent amendments to the ANO-1 ITS. This amendment
request supersedes the previous application related to the revision of
TS Section 6.0 dated January 31, 2002, as supplemented on June 26 and
July 18, 2002. The January 31, 2002, application was previously noticed
in the Federal Register on March 19, 2002 (67 FR 12602).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
Administrative Changes
The proposed changes involve reformatting and rewording of the
existing TSs. The reformatting and rewording process involves no
technical changes to existing requirements. As such, the proposed
changes are administrative in nature and do not impact initiators of
analyzed events or assumed mitigation of accident or transient
events.
Less Restrictive--Administrative Deletion of Requirements
The proposed changes relocate requirements from the TSs to other
license basis documents which are under licensee control. The
documents containing the relocated requirements will be maintained
using the provisions of applicable regulatory requirements.
[[Page 43386]]
More Restrictive Changes
The proposed changes provide more stringent requirements for the
ANO-2 TSs. These more stringent requirements are not assumed to be
initiators of analyzed events and will not alter assumptions
relative to mitigation of accident or transient events. The more
stringent requirements are imposed to ensure process variables,
structures, systems, and components are maintained consistent with
the safety analyses and licensing basis and to provide greater
consistency with the ANO-1 TS and NUREG 1432.
Less Restrictive Changes
(1) A note will be added that allows three (3) hours to perform
the channel functional test on the control room radiation monitors
without entering the associated Actions.
The control room area radiation monitor is used to support
mitigation of the consequences of an accident; however, it is not
considered the initiator of any previously analyzed accident. Also,
the addition of the Note to allow time for testing reduces the
potential for initiation of a previously analyzed accident due to
reduced potential for shutdowns and startups due to incomplete or
missed surveillances. As such, the proposed revision to include an
allowance for testing does not significantly increase the
probability of any accident previously evaluated. This change does
not result in any hardware changes, but does allow operation for a
limited time with an inoperable monitor for the purposes of testing.
Since the capability of the control room area radiation monitor to
provide the required information continues to be verified, and the
time allowed for inoperability for testing is short, the change will
not reduce the capability of required equipment to mitigate the
event. Also, the consequences of an event occurring during the
proposed operation of the unit during the allowed inoperability for
testing are the same as the consequences of an event occurring while
operating under the current TS Actions. Therefore, this change does
not involve a significant increase in the consequences of any
accident previously evaluated.
(2) This change will allow the control room boundary to be
opened intermittently under administrative controls, and will allow
both trains of the CREVS [control room emergency ventilation system]
to be inoperable due to control room boundary inoperability for a
period of 24'hours.
Neither CREVS nor the control room boundary is the initiator of
any accident analyzed in the SAR [Safety Analysis Report].
Therefore, this change does not result in a significant increase in
the probability of an accident previously evaluated.
The CREVS and the control room boundary are intended to provide
a habitable environment for the control room operators in the event
of an accident that results in the release of radioactivity to the
environment. The allowance to open the control room boundary
intermittently is acceptable, because of the administrative controls
that will be implemented to ensure that the opening can be rapidly
closed when the need for control room isolation is indicated,
restoring the control room habitability envelope. Allowing both
CREVS trains to be inoperable for 24 hours due to an inoperable
control room boundary is acceptable because of the low probability
of an accident requiring control room isolation during any given 24
hour period, because entry into this condition is expected to be an
infrequent occurrence, and because preplanned compensatory measures
to protect the control room operators from potential hazards are
implemented. Therefore, this change will not result in a significant
increase in the probability [consequences] of an accident previously
evaluated.
(3) An allowance will be added to allow use of a ``simulated''
or ``actual'' test signal when testing the automatic isolation
feature of the control room air filtration system.
The phrase ``actual or simulated'' in reference to the automatic
initiation signal, has been added to the system functional test
surveillance test description. This does not impose a requirement to
create an ``actual'' signal, nor does it eliminate any restriction
on producing an ``actual'' signal. The proposed change does not
affect the procedures governing plant operations and the
acceptability of creating these signals; it simply would allow such
a signal to be utilized in evaluating the acceptance criteria for
the system functional test requirements. Therefore, the change does
not involve a significant increase in the probability of an accident
previously evaluated. Since the function of the system functional
test remains unaffected the change does not involve a significant
increase in the consequences of an accident previously evaluated.
(4) An allowance for the diesel fuel storage tanks to contain
less than 22,500 gallons of fuel for up to 48 hours as long as the
individual volume is greater than 17,446 gallons will be added. The
lower value when summed with the contents of the other tank ensures
six days of fuel oil is available. During the 48 hours, the diesel
generator is capable of performing its intended function. There is a
low probability that an event would occur for which the diesel
generator would be required during this short period of time when
the lower fuel oil volume is allowed.
The AC Sources are used to support mitigation of the
consequences of an accident and can be involved in the initiation of
the accident analyzed in SAR. Equipment powered by the AC Sources,
which may be considered as an initiator, continues to be assured of
electrical power. The proposed increased restoration time involves
parameters unrelated to initiating the failure of the AC Sources. As
such the proposed time allowance for restoration of limited levels
of readiness parameter degradation will not increase the probability
of any accident previously evaluated. The proposed changes allow
additional time for restoration of parameters that have been
identified as not immediately affecting the capability of the power
source to provide its required safety function. The identified
parameters are capable of being replenished during operation of the
diesel generators, and the short additional allowable action time
continues to provide adequate assurance of operable required
equipment. Therefore, this change does not involve a significant
increase in the probability of or the consequences of any accident
previously evaluated.
(5) Seven days will be allowed to restore the stored diesel fuel
oil total particulates to within the required limits prior to
declaring the associated diesel inoperable.
The testing of diesel generator fuel oil is not considered an
initiator, or a mitigating factor, in any previously evaluated
accident. The presence of particulates does not mean failure of the
fuel oil to burn properly in the diesel engine. In addition,
particulate concentration is unlikely to change significantly
between surveillance intervals (31 days). Therefore, the change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(6) An allowance for the person who is satisfying the
requirement of the radiation protection staff position and for the
person filling the Shift Technical Advisor (STA) position to be
vacant for not more than two hours in order to provide for
unexpected absences is being added. This is consistent with the
allowance permitted for the control room operator as reflected in
existing TSs.
This change does not result in any changes in hardware or
methods of operation. The change allowing the absence of the STA or
the radiation protection technician is not considered in the safety
analysis, and cannot initiate or affect the mitigation of an
accident in any way. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(7) The STA will be allowed to support the shift crew rather
than only the shift supervisor. This provides more flexibility and
does not dilute the function of the STA.
This change does not result in any changes in hardware or
methods of operation. The change in the support relationship between
the STA and the control room staff is not considered in the safety
analysis, and cannot initiate or affect the mitigation of an
accident in any way. Therefore, this change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(8) The Occupational Radiation Exposure Report will be submitted
by April 30 of each calendar year instead of prior to March 1.
This change does not result in any changes in hardware or
methods of operation. The change in date for submittal of ``after
the fact'' information is not considered in the safety analysis, and
cannot initiate or affect the mitigation of an accident in any way.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(9) An allowance is proposed that will revise the high radiation
areas to include additional previously approved methods for
implementation of alternatives to the ``control device'' or ``alarm
signal'' requirements of 10 CFR [Part] 20. These alternatives
provide adequate control of
[[Page 43387]]
personnel in high radiation areas as evidenced by NRC issuance of
NUREG-1432.
The controls for access to a high radiation area are not
considered as initiators, or as a mitigation factor, in any
previously evaluated accident. Therefore, the change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
(10) An allowance to require periodic testing of stored fuel for
the particulates only is proposed.
The testing of diesel generator fuel oil is not considered an
initiator or a mitigating factor in any previously evaluated
accident. Therefore, the change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(11) The removal of the requirement to notify the Vice
President, Operations ANO within 24 hours of violating a safety
limit.
Notification of the Vice President, Operations ANO when a safety
limit is violated is not considered an initiator or a mitigating
factor in any previously evaluated accident. Therefore, the change
does not involve a significant increase in the probability or
consequences of an accident previously evaluated.
(12) The Radioactive Effluent Release Report will be submitted
by May 1 of each calendar year instead of prior to March 1.
This change does not result in any changes in hardware or
methods of operation. The change in date for submittal of ``after
the fact'' information is not considered in the safety analysis, and
cannot initiate or affect the mitigation of an accident in any way.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
(13) A change to frequency of the integrated leak tests for each
system outside containment that could contain highly radioactive
fluids from ``at a frequency not to exceed refueling cycle
intervals'' to ``at least once per 18 months.''
Performance of the integrated leak tests for each system outside
containment that could contain highly radioactive fluids is not an
initiator or a mitigating factor in any previously evaluated
accident. Therefore, the change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
(14) A change that allows a 25% extension of the frequency in
accordance with SR 4.0.2 for the integrated leak tests of each
system outside containment that could contain highly radioactive
fluids.
The extension of the testing frequency, up to 25% of the test
interval, is not considered an initiator or a mitigating factor in
any previously evaluated accident. Therefore, the change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Administrative Changes
The proposed changes do not necessitate a physical alteration of
the plant (no new or different type of equipment will be installed)
or changes in parameters governing normal plant operations. The
proposed changes will not impose any different requirements.
Less Restrictive--Administrative Deletion of Requirements
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operations. The proposed changes will not impose any different
requirements and adequate control of the information will be
maintained.
More Restrictive Changes
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed changes do impose different requirements.
However, these changes do not impact the safety analysis and
licensing basis.
Less Restrictive Changes
(1) A note will be added that allows three (3) hours to perform
the channel functional test on the control room radiation monitors
without entering the associated Actions.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will still ensure proper
surveillances are required for the equipment considered in the
safety analysis. Thus, this change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
(2) This change will allow the control room boundary to be
opened intermittently under administrative controls, and will allow
both trains of the control room ventilation system (CREVS) to be
inoperable due to a control room boundary inoperability for a period
of 24 hours.
The proposed change does not necessitate a physical alteration
of the unit (no new or different type of equipment will be
installed) or changes in parameters governing normal unit operation.
Prompt and appropriate compensatory actions will still be taken in
the event of an accident. Thus, this change does not create the
possibility of a new or different kind of accident from any
previously evaluated.
(3) An allowance will be added to allow use of a ``simulated''
or ``actual'' test signal when testing the automatic isolation
feature of the control room air filtration system.
The possibility of a new or different kind of accident from any
accident previously evaluated is not created because the proposed
change introduces no new mode of plant operation and it does not
involve physical modification to the plant.
(4) An allowance for the diesel fuel storage tanks to contain
less than 22,500 gallons of fuel for up to 48 hours as long as the
individual volume is greater than 17,446 gallons will be added. The
lower value when summed with the contents of the other tank ensures
six days of fuel oil is available. During the 48 hours, the diesel
generator is capable of performing its intended function. There is a
low probability that an event would occur for which the diesel
generator would be required during this short period of time when
the lower fuel oil volume is allowed.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will continue to ensure operable
safety equipment is available. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(5) Seven days will be allowed to restore the stored diesel fuel
oil total particulates to within the required limits prior to
declaring the associated diesel inoperable.
No changes are proposed in the manipulation of the plant
structures, systems, or components, or in the design of the plant
structures, systems, or components. The presence of particulates
does not mean failure of the fuel oil to burn properly in the diesel
engine. In addition, particulate concentration is unlikely to change
significantly between surveillance intervals (31 days). Therefore,
the change does not create the possibility of a new or different
kind of accident from any accident previously evaluated.
(6) An allowance for the person who is satisfying the
requirement of the radiation protection staff position and for the
person filling the Shift Technical Advisor (STA) position to be
vacant for not more than two hours in order to provide for
unexpected absences is proposed. This is consistent with the
allowance permitted for the control room operator as reflected in
existing TSs.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will impact only the STA and
radiation protection staffing positions and does not directly impact
the operation of the plant. Thus, this change does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
(7) The STA will be allowed to support the shift crew rather
than only the shift supervisor. This provides more flexibility and
does not dilute the function of the STA.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will impact only the support
relationship the STA provides the control room staff and does not
directly impact the operation of the plant. Thus, this change does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
(8) The Occupational Radiation Exposure Report will be submitted
by April 30 of each calendar year instead of prior to March 1.
[[Page 43388]]
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will impact only the administrative
requirements for submittal of information and does not directly
impact the operation of the plant. Thus, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(9) An allowance is proposed that will revise the high radiation
areas to include additional previously approved methods for
implementation of alternates to the ``control device'' or ``alarm
signal'' requirements of 10 CFR [Part] 20. These alternatives
provide adequate control of personnel in high radiation areas as
evidenced by NRC issuance of NUREG-1432.
No changes are proposed in the manipulation of the plant
structures, systems, or components, or in the design of the plant
structures, systems, or components. Therefore, the change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(10) An allowance to require periodic testing of stored fuel for
the particulates only is proposed.
No changes are proposed in the manipulation of the plant
structures, systems, or components, or in the design of the plant
structures, systems, or components. Therefore, the change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
(11) The removal of the requirement to notify the Vice
President, Operations ANO within 24 hours of violating a safety
limit.
No changes are proposed that result in the manipulation or the
design of plant structures, systems, or components. Therefore, the
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(12) The Radioactive Effluent Release Report will be submitted
by May 1 of each calendar year instead of prior to March 1.
The proposed change does not necessitate a physical alteration
of the plant (no new or different type of equipment will be
installed) or changes in parameters governing normal plant
operation. The proposed change will impact only the administrative
requirements for submittal of information and does not directly
impact the operation of the plant. Thus, this change does not create
the possibility of a new or different kind of accident from any
accident previously evaluated.
(13) A change to frequency of the integrated leak tests for each
system outside containment that could contain highly radioactive
fluids from ``at a frequency not to exceed refueling cycle
intervals'' to ``at least once per 18 months.''
No changes are proposed that result in the manipulation or the
design of plant structures, systems, or components. Therefore, the
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
(14) A change that allows a 25% extension of the frequency in
accordance with SR 4.0.2 for the integrated leak tests of each
system outside containment that could contain highly radioactive
fluids.
No changes are proposed that result in the manipulation or the
design of plant structures, systems, or components. Therefore, the
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously evaluated.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
Administrative Changes
The proposed changes will not reduce the margin of safety
because they have no impact on any safety analysis assumptions. The
changes are administrative in nature.
Less Restrictive--Administrative Deletion of Requirements
The proposed changes will not reduce a margin of safety because
they have no impact on any safety analysis assumptions. In addition,
the requirements to be transposed from the TSs to other license
basis documents, which are under licensee control, are the same as
the existing TSs. The documents containing the relocated
requirements will be maintained using the provisions of applicable
regulatory requirements.
More Restrictive Changes
The imposition of more stringent requirements prevents a
reduction in the margin of plant safety by:
(a) increasing the scope of the specification to include
additional plant equipment,
(b) providing additional actions,
(c) decreasing restoration times, or
(d) imposing new surveillances.
The changes are consistent with the safety analysis and
licensing basis.
Less Restrictive Changes
(1) A note will be added that allows three (3) hours to perform
the channel functional test on the control room radiation monitors
without entering the associated Actions.
The margin of safety for the control room area radiation monitor
is based on availability and capability of the instrumentation to
provide the required information to the operator. The frequency is
based on unit operating experience that demonstrates channel failure
is rare, and on the use of less formal but more frequent checks of
channels during normal operational use of the displays associated
with the required channels. Therefore, the availability and
capability of the control room area radiation monitor continues to
be assured by the proposed Surveillance Requirements and this change
does not involve a significant reduction in a margin of safety.
(2) This change will allow the control room boundary to be
opened intermittently under administrative controls, and will allow
both trains of the control room ventilation system (CREVS) to be
inoperable due to control room boundary inoperability for a period
of 24 hours.
This change does not involve a significant reduction in a margin
of safety since: (1) administrative controls will be in place to
ensure that an open control room boundary can be rapidly closed when
a need for control room isolation is indicated; and (2) an
inoperable control room boundary that renders both trains of CREVS
inoperable is an infrequent occurrence, the probability of an
accident requiring control room isolation during any given 24 hour
period is low, and preplanned compensatory measures to protect the
control room operators from potential hazards are implemented.
(3) An allowance will be added to use a simulated or actual test
signal when testing the automatic isolation feature of the control
room air filtration system.
Use of an actual signal instead of the existing requirement
which limits use to a simulated signal, will not affect the
performance of the surveillance test. OPERABILITY is adequately
demonstrated in either case since the system itself can not
discriminate between ``actual'' or ``simulated'' signals. Therefore,
the change does not involve a significant reduction in a margin of
safety.
(4) An allowance for the diesel fuel storage tanks to contain
less than 22,500 gallons of fuel for up to 48 hours as long as the
individual volume is greater than 17,446 gallons. The lower value
when summed with the contents of the other tank ensures six days of
fuel oil is available. During the 48 hours, the diesel generator is
capable of performing its intended function. There is a low
probability that an event would occur for which the diesel generator
would be required during this short period of time when the lower
fuel oil volume is allowed.
The parameter limits provide substantial margin to the parameter
values that would be absolutely necessary for diesel generator
operability. When the parameters are less than their limits this
margin is reduced. However, the availability of AC Sources continues
to be assured since the allowed time for parameters to be less than
their limits is short and the allowed levels for the parameters are
adequate to provide the immediately needed power availability.
Further, the parameters can be restored to within limits during the
proposed time provided should they be required. Therefore, this
change does not result in a signification reduction in [a] margin of
safety.
(5) Seven days will be allowed to restore the stored diesel fuel
oil total particulates to within the required limits prior to
declaring the associated diesel inoperable.
The proposed change allows the stored diesel fuel oil total
particulates to be outside the required limits for seven days before
declaring the associated diesel inoperable. The presence of
particulates does not mean failure of the fuel oil to burn properly
in the diesel engine. In addition, particulate concentration is
unlikely to change
[[Page 43389]]
significantly between surveillance intervals (31 days). The seven
day allowance provides an appropriate backstop to ensure the
particulate level is restored to within limits in a reasonable time
period. Since the diesel is still capable of performing its function
the margin to safety is not reduced.
(6) An allowance for the person who is satisfying the
requirement of the radiation protection staff position and for the
person filling the Shift Technical Advisor (STA) position to be
vacant for not more than two hours in order to provide for
unexpected absences is proposed. This is consistent with the
allowance permitted for the control room operator as reflected in
existing TSs.
The margin of safety is not dependent on the presence of the STA
or the radiation protection technician. Therefore, this change does
not involve a significant reduction in a margin of safety.
(7) The STA will be allowed to support the shift crew rather
than only the shift supervisor. This provides more flexibility and
does not dilute the function of the STA.
The margin of safety is not dependent upon who the STA supports.
Therefore, this change does not involve a significant reduction in a
margin of safety.
(8) The Occupational Radiation Exposure Report will be submitted
by April 30 of each calendar year instead of prior to March 1.
The margin of safety is not dependent on the submittal of
information. Therefore, this change does not involve a significant
reduction in a margin of safety.
(9) An allowance is proposed that will revise the high radiation
areas to include additional previously approved methods for
implementation of alternatives to the ``control device'' or ``alarm
signal'' requirements of 10 CFR [Part] 20. These alternatives
provide adequate control of personnel in high radiation areas as
evidenced by NRC issuance of NUREG-1432.
The requirements for control of high radiation areas provide for
the use of alternates to the ``control device'' or ``alarm signal''
requirements of 10 CFR 20.1601. This change provides such
alternative methods for controlling access. These methods and
additional administrative requirements have been determined to
provide adequate controls to prevent unauthorized and inadvertent
access to such areas. Therefore, this change does not involve a
significant reduction in a margin of safety.
(10) An allowance to require periodic testing of stored fuel for
the particulates only is proposed.
The testing of stored diesel generator fuel oil is revised to
require the periodic testing of the stored fuel oil only for
particulates (replacing the periodic testing per ASTM-D975) once
every 31 days. The change reflects industry-standard acceptable DG
fuel oil testing programs. Over the storage life of ANO-2 DG fuel
oil, the properties tested by ASTM-D975 are not expected to change
and performing these tests once on the new fuel oil provides
adequate assurance of the proper initial quality of fuel oil. The
periodic testing for particulates monitors a parameter that reflects
degradation of fuel oil and can be trended to provide increased
confidence that the stored DG fuel oil will support DG operability.
Therefore, this change does not involve a significant reduction in a
margin of safety.
(11) The removal of the requirement to notify the Vice
President, Operations ANO within 24 hours of violating a safety
limit.
The margin of safety is not dependent upon notification of the
Vice President, Operations ANO upon the violation of a TS safety
limit. Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
(12) The Radioactive Effluent Release Report will be submitted
by May 1 of each calendar year instead of prior to March 1.
The margin of safety is not dependent on the submittal of
information. Therefore, this change does not involve a significant
reduction in a margin of safety.
(13) A change to frequency of the integrated leak tests for each
system outside containment that could contain highly radioactive
fluids from ``at a frequency not to exceed refueling cycle
intervals'' to ``at least once per 18 months.''
The current and proposed frequencies of this test are equivalent
for all practical purposes. Therefore, this change does not involve
a significant reduction in a margin of safety.
(14) A change that allows a 25% extension of the frequency in
accordance with SR 4.0.2 for the integrated leak tests of each
system outside containment that could contain highly radioactive
fluids.
The proposed allowance allows a possible increase in performance
interval. However, the test will still be performed at reasonable
intervals to ensure the intent of the surveillance is maintained.
Therefore, this change does not involve a significant reduction in a
margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Nicholas S. Reynolds, Esquire, Winston and
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Inc., Docket No. 50-286, Indian Point
Nuclear Generating Unit No. 3, Westchester County, New York
Date of amendment request: May 28, 2003, as supplemented on June
24, 2003
Description of amendment request: The proposed amendment would
revise Technical Specification (TS) Section 3.4.3, ``RCS Pressure and
Temperature (P/T) Limits,'' and Section 3.4.12, ``Low Temperature
Overpressure Protection (LTOP),'' to incorporate revised reactor
pressure vessel P/T limits and overpressure protection system limits to
allow operation up to 20 effective full-power years. Specifically, the
proposed amendment would revise TS Figures 3.4.3-1 to 3.4.3-3 and TS
Figures 3.4.12-1 to 3.4.12-4.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The probability of occurrence of an accident previously
evaluated for Indian Point 3 is not altered by the proposed
amendment to the technical specifications (TSs). The accidents
remain the same as currently analyzed in Final Safety Analysis
Report (FSAR) as a result of changes to the P/T and LTOP limits. The
new P/T and LTOP limits were based on the NRC [Nuclear Regulatory
Commission] approved, for Indian Point 3, Westinghouse/Combustion
Engineering methodology along with American Society of Mechanical
Engineers (ASME) Code (Boiler and Pressure Vessel Code) alternatives
including Code Case N-640. Code Case N-640 has been accepted for use
by the NRC but has not been incorporated into Reg. [Regulatory]
Guide 1.147, Rev. 12, at this time. An exemption is being submitted
separately for the use of Code Case N-640. The proposed changes do
not impact the integrity of the reactor coolant system pressure
boundary (RCPB) as a result of this change. In addition there is no
increase in the potential for the occurrence of a loss of coolant
accident. The probability of any design basis accident is not
affected by the change, nor are the consequences of any design basis
accident affected by the proposed change. The proposed P/T limit
curves and LTOP limits are not considered to be an initiator or
contributor to any accident currently evaluated in the Indian Point
3 FSAR. These new limits ensure the long term integrity of the RCPB.
Fracture toughness test data are obtained from material
specimens contained in capsules that are periodically withdrawn from
the reactor vessel. These data permit determination of the
conditions under which the vessel can be operated with adequate
safety margins against non-ductile fracture throughout its service
life. A new reactor vessel specimen was withdrawn at the most recent
refueling outage and will be analyzed over the next year to enhance
the database used to predict the fracture toughness requirements
using projected neutron fluence calculations. For each analyzed
transient and steady state condition, the allowable pressure is
determined as a function of reactor coolant temperature considering
postulated flaws in the reactor vessel beltline, inlet nozzle,
outlet nozzle, and closure head.
The predicted radiation induced [Delta]RTNDT (shift
in reference temperature nil-ductility
[[Page 43390]]
transition) was calculated using the respective reactor vessel
beltline copper and nickel contents and the neutron fluence
applicable to normal plant performance through the remainder of the
operating license, using the most up-to-date cross sections
methodologies, as documented in the recent Appendix K power uprate
report.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the P/T and the LTOP limits will not
create a new accident scenario. The requirements to have P/T and
LTOP protection are part of the licensing basis of Indian Point 3.
The proposed changes reflect the change in vessel material
properties acknowledged and managed by regulation and the best data
available in response to NRC Generic Letter 92-01, Revision 1. The
approach used meets NRC and ASME regulations and guidelines. The
Westinghouse/Combustion Engineering methodology has been approved
for use at Indian Point 3 by the NRC. Code Case N-640 has been found
acceptable by the NRC to be used at other nuclear plants. By
separate letter ENO [Entergy Nuclear Operations, Inc.] is requesting
an exemption to use Code Case N-640 because the Code Case has not
been incorporated in Regulatory Guide 1.147, Rev. 12, at this time.
The adjusted reference temperatures for fracture toughness are
consistent with that previously provided to the NRC [* * *] The data
analysis for the vessel specimen removed to date, confirm that the
vessel materials are responding as predicted.
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 existing P/T curves and LTOP limits in the technical
specifications are reaching their expiration period for the number
of years at effective full power operation. The revision of the P/T
limits and curves will ensure that Indian Point 3 continues to
operate within margins allowed by 10 CFR 50.60 and the ASME Code.
The material properties used in the analysis are based on results
established through Westinghouse/Combustion Engineering material
reports for copper and nickel content. The material properties were
evaluated in parallel using statistical methodology. The results are
consistent and for conservative purposes, the more restrictive
result is used. The application of Code Case N-640 presents
alternative procedures for calculating P/T and LTOP temperatures and
pressures in lieu of that established for ASME Section XI, Appendix
G-2215. This Code alternative allows certain assumptions to be
conservatively reduced. However, the procedures allowed by Code Case
N-640 still provide significant conservatism and ensure an adequate
margin of safety in the development of P/T operating and pressure
test limits to prevent non-ductile fractures.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John Fulton, Assistant General Counsel,
Entergy Nuclear Operations, Inc., 440 Hamilton Avenue, White Plains, NY
10601.
NRC Section Chief: Richard J. Laufer.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: August 16, 2002, as supplemented June 6,
2003.
Description of amendment request: The proposed amendment would add
a new Technical Specification (TS) requirement to the Pilgrim Nuclear
Power Station (Pilgrim) TSs consistent with Technical Specification
Task Force (TSTF)-358, Revision 5. TSTF-358 addresses modifications to
requirements for missed surveillances consistent with NUREG 1433,
Revision 2, ``Standard Technical Specification, General Electric
Plants, BWR/4'' (STS) surveillance requirement (SR) 3.0.3. The proposed
amendment to the Pilgrim TSs would be added as TS 4.0.3.
The U.S. Nuclear Regulatory Commission (NRC) staff issued a notice
of opportunity for comment in the Federal Register on June 14, 2001 (66
FR 32400), on possible amendments concerning missed surveillances,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process (CLIIP). The NRC staff subsequently issued a notice
of availability of the models for referencing in license amendment
applications in the Federal Register on September 28, 2001 (66 FR
49714). The licensee affirmed the applicability of the model NSHC
determination in its application dated August 16, 2002, as supplemented
on June 6, 2003.
In addition, the following statement would be added to the TS
definition of Limiting Condition for Operation (LCO): ``Failure to meet
a Surveillance, whether such failure is experienced during the
performance of the Surveillance or between performances of the
Surveillance, shall be failure to meet the LCO.'' The proposed
amendment would also make administrative changes to add new TS Sections
3.0, ``Limiting Condition for Operation (LCO) Applicability,'' and 4.0,
``Surveillance Requirement (SR) Applicability,'' into the Pilgrim TSs.
New TSs 3.0, 4.0.1, and 4.0.2 would be identified as ``Not Used.''
These changes are proposed to rectify the differences in the format and
terminology of the current Pilgrim TSs to the STS. The associated Bases
would also be implemented.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR), Section 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.
[CLIIP Changes]
The proposed change relaxes the time allowed to perform a missed
surveillance. The time between surveillances is not an initiator of
any accident previously evaluated. Consequently, the probability of
an accident previously evaluated is not significantly increased. The
equipment being tested is still required to be operable and capable
of performing the accident mitigation functions assumed in the
accident analysis. As a result, the consequences of any accident
previously evaluated are not significantly affected. Any reduction
in confidence that a standby system might fail to perform its safety
function due to a missed surveillance is small and would not, in the
absence of other unrelated failures, lead to an increase in
consequences beyond those estimated by existing analyses. The
addition of a requirement to assess and manage the risk introduced
by the missed surveillance will further minimize possible concerns.
Therefore, this change does not involve a significant increase in
the probability or consequences of an accident previously evaluated.
[Additional Changes]
The proposed change involves an addition to clarify the required
action when an SR is not met and new TS sections for consistency
with the STS. These additions do not involve technical changes to
the existing TSs. As such, these changes provide clarity and are
administrative in nature and do not affect initiators of analyzed
events or assumed mitigation of accident or transient events.
Therefore, these changes will not increase the probability or
consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility of
a New or Different Kind of Accident From Any Previously Evaluated.
[CLIIP Changes]
The proposed change does not involve a physical alteration of
the plant (no new or
[[Page 43391]]
different type of equipment will be installed) or a change in the
methods governing normal plant operation. A missed surveillance will
not, in and of itself, introduce new failure modes or effects and
any increased chance that a standby system might fail to perform its
safety function due to a missed surveillance would not, in the
absence of other unrelated failures, lead to an accident beyond
those previously evaluated. The addition of a requirement to assess
and manage the risk introduced by the missed surveillance will
further minimize possible concerns. Thus, this change does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
[Additional Changes]
The proposed change involves an addition to clarify the required
action when a SR is not met and new TS sections for consistency with
the STS. The changes do not involve physical alterations to the
plant (no new or different type of equipment will be installed) or
changes in methods governing normal plant operation. The changes
will not impose any new or different requirements or eliminate any
existing requirements. Therefore, these changes will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in the Margin of Safety
[CLIIP Changes]
The extended time allowed to perform a missed surveillance does
not result in a significant reduction in the margin of safety. As
supported by the historical data, the likely outcome of any
surveillance is verification that the LCO is met. Failure to perform
a surveillance within the prescribed frequency does not cause
equipment to become inoperable. The only effect of the additional
time allowed to perform a missed surveillance on the margin of
safety is the extension of the time until inoperable equipment is
discovered to be inoperable by the missed surveillance. However,
given the rare occurrence of inoperable equipment, and the rare
occurrence of a missed surveillance, a missed surveillance on
inoperable equipment would be very unlikely. This must be balanced
against the real risk of manipulating the plant equipment or
condition to perform the missed surveillance. In addition, parallel
trains and alternate equipment are typically available to perform
the safety function of the equipment not tested. Thus, there is
confidence that the equipment can perform its assumed safety
function.
Therefore, this change does not involve a significant reduction
in a margin of safety.
[Additional Changes]
The proposed change involves an addition to clarify the required
action when a SR is not met and new TS sections for consistency with
the STS. These additions do not involve technical changes to the
existing TSs. The changes will not reduce a margin of safety because
they have no impact on any safety analysis assumptions. Also, since
these changes provide clarity and are administrative in nature, no
question of safety is involved. Therefore, there will be no
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: J. M. Fulton, Esquire, Assistant General
Counsel, Pilgrim Nuclear Power Station, 600 Rocky Hill Road, Plymouth,
Massachusetts, 02360-5599.
NRC Section Chief: James W. Clifford.
FirstEnergy Nuclear Operating Company, Docket No. 50-346, Davis-Besse
Nuclear Power Station, Unit 1, Ottawa County, Ohio
Date of amendment request: December 4, 2001.
Description of amendment request: The proposed amendment would
change Technical Specification (TS) Section 6.9, ``Administrative
Controls--Reporting Requirements,'' to eliminate the requirement to
submit startup reports to the Nuclear Regulatory Commission. Under the
current provisions of TS Section 6.9, the Davis-Besse Nuclear Power
Station would be required to submit a startup report within 90 days.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensees have
provided their analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change is administrative in nature. As such, it
does not affect any accident initiators and does not affect
containment isolation, plant responses to accidents, or radiological
effluents. Therefore, the proposed change does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change is administrative in nature. As such, it
does not introduce any new or indifferent accident initiators.
Therefore, the proposed change does not create the possibility of a
new or different kind of accident from any previous evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change is administrative in nature and does not
reduce or adversely affect the capabilities of any plant structures,
systems, or components to perform their safety functions. 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: Mary E. O'Reilly, Attorney, FirstEnergy
Corporation, 76 South Main Street, Akron, OH 44308.
NRC Section Chief: Anthony J. Mendiola.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of amendment request: June 20, 2003.
Description of amendment request: The proposed amendment would
revise the Kewaunee Nuclear Power Plant (KNPP) Technical Specifications
(TSs) to allow a one-time extension of the interval between integrated
leakage rate tests from 10 years to 15 years.
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:
Proposed Power Level Changes
1. Does the proposed change involve a significant increase in
the probability of occurrence or consequences of an accident
previously evaluated?
Response: No
Probability of Occurrence of an Accident Previously Evaluated--
The proposed change to extend the [integrated leakage rate
tests] ILRT interval from 10 to 15 years does not affect any
accident initiators or precursors. The containment vessel function
is purely mitigative. There is no design basis accident that is
initiated by a failure of the containment leakage mitigation
function. The extension of the ILRT will not create any adverse
interactions with other systems that could result in initiation of a
design basis accident. Therefore, the probability of occurrence of
an accident previously evaluated is not significantly increased.
Consequences of an Accident Previously Evaluated--
The potential consequences of the proposed change have been
quantified by analyzing the changes in risk that would result from
extending the ILRT interval from
[[Page 43392]]
10 to 15 years. The increase in risk in terms of person rem per year
within 50 miles resulting from design basis accidents was estimated
to be of a magnitude that NUREG-1493 indicates is imperceptible. NMC
has also analyzed the increase in risk in terms of the frequency of
large early releases from accidents. The increase in the large early
release frequency resulting from the proposed extension was
determined to be within the guidelines published in Regulatory Guide
1.174. Additionally, the proposed change maintains defense-in-depth
by preserving a reasonable balance among prevention of core damage,
prevention of containment failure, and consequence mitigation. NMC
has determined that the increase in conditional containment failure
probability from reducing the ILRT frequency from 1 test per 10
years to 1 test per 15 years would be small. Continued containment
integrity is also assured by the history of successful ILRTs, and
that established programs for local leakage rate testing and in-
service inspections which are unaffected by the proposed change.
Therefore, the consequences of an accident previously analyzed are
not significantly increased.
In summary, the probability of occurrence and the consequences
of an accident previously evaluated are not significantly increased.
2. Does the proposed change create the possibility of a new or
different kind of accident from any previously evaluated?
Response: No.
The proposed change to extend the ILRT interval from 10 to 15
years does not create any new or different accident initiators or
precursors. The length of the ILRT interval does not affect the
manner in which any accident begins. The proposed change does not
create any new failure modes for the containment and does not affect
the interaction between the containment and any other system. Thus,
the proposed changes do 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 risk-based margins of safety associated with the containment
ILRT are those associated with the estimated person-rem per year,
the large early release frequency, and the conditional containment
failure probability. NMC has quantified the potential effect of the
proposed change on these parameters and determined that the effect
is not significant. The non-risk-based margins of safety associated
with the containment ILRT are those involved with its structural
integrity and leak tightness. The proposed change to extend the ILRT
interval from 10 to 15 years does not adversely affect either of
these attributes. The proposed change only affects the frequency at
which these attributes are verified. Therefore, the proposed change
does not involve a significant reduction in 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: John H. O'Neill, Jr., Esq., Shaw Pittman,
Potts & Trowbridge, 2300 N. Street, NW, Washington, DC 20037-1128.
NRC Section Chief: L. Raghavan.
Pacific Gas and Electric Company, Docket No. 50-323, Diablo Canyon
Nuclear Power Plant, Unit No. 2, San Luis Obispo County, California
Date of amendment requests: June 26, 2003.
Description of amendment requests: The proposed license amendment
would update the Diablo Canyon Power Plant (DCPP) Final Safety Analysis
Report Update to use a revised steam generator voltage-based repair
criteria probability of detection method for DCPP Unit 2 Cycle 12 using
plant-specific inspection results.
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 use of a revised steam generator (SG) voltage-based repair
criteria probability of detection (POD) method, the probability of
prior cycle detection (POPCD) method, to determine the beginning of
cycle (BOC) indication voltage distribution for the Diablo Canyon
Power Plant (DCPP) Unit 2 Cycle 12 operational assessment (OA) does
not increase the probability of an accident. Based on industry and
plant specific bobbin detection data for outside diameter stress
corrosion cracks (ODSCC) within the SG tube support plate (TSP)
region, large voltage bobbin indications which individually can
challenge structural or leakage integrity can be detected with near
100 percent certainty. Since large voltage ODSCC bobbin indications
within the SG TSP can be detected, they will not be left in service,
and therefore these indications should not be included in the
voltage distribution for the purpose of OAs. POPCD improves the
estimate of potentially undetected indications for OAs, but does not
directly affect the inspection results. Since large voltage
indications are detected, they will not result in an increase in the
probability of a steam generator tube rupture (SGTR) accident or an
increase in the consequences of a SGTR or main steam line break
(MSLB) accident.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The use of the POPCD method to determine the BOC voltage
distribution for the DCPP Unit 2 Cycle 12 OA concerns the SG tubes
and can only affect numerical predictions of probabilities for the
SGTR accident. Since the SGTR accident is already considered in the
Final Safety Analysis Report Update, there [is] no possibility to
create a design basis accident that has not been previously
evaluated.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The use of the POPCD method to determine the BOC voltage
distribution for the DCPP Unit 2 Cycle 12 OA does not involve a
significant reduction in a margin of safety. The applicable margin
of safety potentially impacted is the Technical Specification
5.6.10, ``Steam Generator Tube Inspection Report,'' projected end-
of-cycle leakage for a MSLB accident and the projected end-of-cycle
probability of burst. Based on industry and plant specific bobbin
detection data for ODSCC within the SG TSP region, large voltage
bobbin indications that can individually challenge structural or
leakage integrity can be detected with near 100 percent certainty
and will not be left in service. Therefore these indications should
not be included in the voltage distribution for the purpose of OAs.
Since these large voltage indications are detected, they will not
result in a significant increase in the actual end-of-cycle leakage
for a MSLB accident or the actual end-of-cycle probability of burst.
The POPCD approach to probability of detection considers the
potential for missing indications that might challenge structural or
leakage integrity by applying the POPCD data from successive
inspections. If a large indication was missed in one inspection, it
would continue to grow until finally detected in a later inspection.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment requests involve no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PPL Susquehanna, LLC, Docket Nos. 50-387 and 50-388, Susquehanna Steam
Electric Station, Units 1 and 2, Luzerne County, Pennsylvania
Date of amendment request: July 3, 2003.
Description of amendment request: The proposed amendments would
change the Technical Specifications (TSs) 3.8.1 for AC Sources--
Operating,
[[Page 43393]]
to extend, on a one-time basis, the allowable Completion Time for
Required Actions associated with one offsite circuit inoperable, from
72 hours to 10 days.
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 proposal would change the Technical Specifications for AC
Sources--Operating, to extend, on a one-time basis, the allowable
Completion Times for Required Actions for one offsite circuit
inoperable, from 72 hours to 10 days. The proposed change does not
involve a significant increase in the probability of an accident
previously evaluated because the probability increases are within
the guidance provided in Regulatory Guide 1.177.
The consequence[s] of losing offsite power have been evaluated
in the FSAR [Final Safety Analysis Report] and the Station Blackout
evaluation. Increasing the completion time for one offsite power
source from 72 hours to 10 days does not increase the consequences
of a LOOP [loss of offsite power] event nor change the evaluation of
LOOP events as stated in the FSAR or Station Blackout evaluation.
Therefore, this change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed
nor will there be changes in methods governing normal plant
operation).
Allowing the completion time for ST [startup transformer] No. 10
to increase from 72 hours to 10 days is a one-time change that will
allow continued operation of Unit 1 while replacing Startup
Transformer Number 10. The accident analyses affected by this
extension are the LOOP events that are discussed in the FSAR. The
potential for the loss of other plant systems or equipment to
mitigate the effects of an accident is not altered.
Thus, this change does not create the possibility of a new or
different kind of accident from any accident previously evaluated.
3. Does the proposed change amendment involve a significant
reduction in a margin of safety?
Response: No
The proposed change does not involve a significant reduction in
[a] margin of safety.
The proposed change allows, on a one-time basis, ST No. 10 to be
out of service for 7 days more than is allowed by Technical
Specifications. This increase in completion time for ST No. 10
results in a slight decrease in the margin of safety. Implementation
of the compensatory measures described in Section 4.0 mitigates the
increase in the core damage frequency and large early release
frequency during this time, such that the potential impact of
extending the completion time is small. Therefore, this one-time
exemption will not involve a significant reduction in safety margin.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Section Chief: Richard J. Laufer.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: June 2, 2003.
Description of amendment request: The amendment would increase the
value of the minimum fuel oil required in the storage tank for the
emergency diesel generators in Technical Specification (TS) 3.8.3,
``Diesel Fuel Oil, Lube Oil, and Starting Air.'' The licensee stated it
has implemented the change in the field. This was done because the
proposed new value is higher than the current value in the 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. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses since there are
no hardware changes. The design of the emergency diesel engine fuel
oil storage and transfer system and the function of the onsite
standby power sources will be unaffected. The only physical change
is to increase the [minimum] volume of fuel oil required to run the
emergency diesel generators at their continuous rating for 6 days.
This change has already been implemented in the field and is in the
conservative direction. The fuel oil storage and transfer system
will continue to function in a manner consistent with the plant
design basis. All design, material, and construction standards that
were applicable prior to this amendment request are maintained.
The probability and consequences of accidents previously
evaluated in the FSAR [(Callaway Final Safety Analysis Report)] are
not adversely affected because the change to the [minimum] volume of
fuel oil required is conservative and is consistent with the safety
analysis and licensing basis.
The proposed change will not affect the probability of any event
initiators. There will be no degradation in the performance of, or
an increase in the number of challenges imposed on, safety-related
equipment assumed to function during an accident situation. There
will be no change to normal plant operating parameters or accident
mitigation performance.
The proposed change will not alter any assumptions or change any
mitigation actions in the radiological consequence evaluations in
the FSAR.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no hardware changes nor are there any changes in the
method by which any safety-related plant system performs its safety
function. This amendment will not affect the normal method of plant
operation or change any operating parameters. The proposed change
does not induce a new mechanism that would result in a different
kind of accident from those previously analyzed. No performance
requirements or response time limits will be affected.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this amendment.
This amendment does not alter the performance of the emergency
diesel engine fuel oil storage and transfer system in [its] support
of the onsite standby power sources.
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 a margin of safety.
The proposed change does not eliminate any surveillances or
alter the frequency of surveillances required by the Technical
Specifications. The minimum volume of fuel oil required for a 6[-
]day supply as specified in the TS has already been increased in the
conservative direction. The safety analysis limits assumed in the
transient and accident analyses are unchanged. None of the
acceptance criteria for any accident analysis are changed.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on any margin of
safety. The radiological dose consequence acceptance criteria listed
in the [NRC] Standard Review Plan will continue to be met.
[[Page 43394]]
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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC. 20037
NRC Section Chief: Stephen Dembek
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: June 6, 2003
Description of amendment request: The amendment would modify
several surveillance requirements (SRs) in Technical Specifications
(TSs) 3.8.1 and 3.8.4 on alternating current and direct current
sources, respectively, for plant operation. The revised SRs would have
notes deleted or modified to allow the SRs to be performed, or
partially performed, in reactor modes that are currently not allowed by
the TSs. The current SRs are not allowed to be performed in Modes 1 and
2. Several of the current SRs also cannot be performed in Modes 3 and
4.
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 design of plant equipment is not being modified by the
proposed changes. In addition, the DGs [diesel generators] and their
associated emergency loads are accident mitigating features. As
such, testing of the DGs themselves is not associated with any
potential accident-initiating mechanism. Therefore, there will be no
significant impact on any accident probabilities by the approval of
the requested changes.
The changes include an increase in the online time that a DG
under test will be paralleled to the grid (for SRs 3.8.1.10 and
3.8.1.14) or unavailable due to testing (per SR 3.8.1.13). As such,
the ability of the tested DG to respond to a design basis accident
[(DBA)] could be adversely impacted by the proposed changes.
However, the impacts are not considered significant based, in part,
on the ability of the remaining DG to mitigate a DBA or provide safe
shutdown. With regard to SR 3.8.1.10 and SR 3.8.1.14, experience
shows that testing per these SRs typically does not perturb the
electrical distribution system. In addition, operating experience
and qualitative evaluation of the probability of the DG or bus loads
being adversely affected concurrent with or due to a significant
grid disturbance, while the DG is being tested, support the
conclusion that the proposed changes do not involve any significant
increase in the likelihood of a safety-related bus blackout or
damage to plant loads.
The SR changes that are consistent with TSTF [Technical
Specification Task Force]--283 have been approved by the NRC for
submittal by licensees. The on-line tests allowed by the TSTF are
only to be performed for the purpose of establishing OPERABILITY.
Performance of these SRs during restricted MODES will require an
assessment to assure plant safety is maintained or enhanced.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The capability to synchronize a DG to the offsite source (via
the associated plant bus) and test the DG in such a configuration is
a design feature of the DGs, including the test mode override in
response to a safety injection signal. Paralleling the DG for longer
periods of time during plant operation may slightly increase the
probability of incurring an adverse effect from the offsite source,
but this increase in probability is judged to be still quite small
and such a possibility is not a new or previously unrecognized
consideration.
The proposed changes would not require any new or different
accidents to be postulated since no changes are being made to the
plant that would introduce any new accident causal mechanisms. This
license amendment request does not impact any plant systems that are
potential accident initiators; nor does it have any significantly
adverse impact on any accident mitigating systems.
Therefore, the proposed change does not create the possibility
of a new or different accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed changes do not involve a significant reduction in
the margin of safety. The margin of safety is related to the
confidence in the ability of the fission product barriers to perform
their design [safety] functions during and following an accident
situation. These barriers include the fuel cladding, the reactor
coolant system, and the containment system. The proposed changes do
not directly affect these barriers, nor do they involve any
significantly adverse impact on the DGs which serve to support these
barriers in the event of an accident concurrent with a loss of
offsite power. The proposed changes to the testing requirements for
the plant DGs do not affect the OPERABILITY requirements for the
DGs, as verification of such OPERABILITY will continue to be
performed as required (except during different allowed MODES [of
operation]). The changes have an insignificant impact on DG
availability, as continued verification of OPERABILITY supports the
capability of the DGs to perform their required [safety] function of
providing emergency power to plant equipment that supports or
constitutes the fission product barriers. Only one DG is to be
tested at a time, so that the remaining DG will be available to
safely shut down the plant if required. Consequently, performance of
the fission product barriers will not be impacted by implementation
of the proposed amendment.
In addition, the proposed changes involve no changes to [safety]
setpoints or limits established or assumed by the accident
analys[e]s. On this and the above basis, no safety margins will be
impacted.
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: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: June 27, 2003.
Description of amendment request: The amendment would revise the
technical specifications (TSs) in two parts. It would: (1) Revise the
definition of dose equivalent radioiodine 131 (I-131) by adding the
phrase ``or those derived from the data provided in International
Commission on Radiological Protection Publication 30 (ICRP 30), `Limits
for Intakes of Radionuclides by Workers,' 1979'' to the current
definition, and (2) increase the maximum allowed closure time of each
main feedwater isolation valve (MFIV) from 5 seconds to 15 seconds in
Surveillance Requirement 3.7.3.1. A plant modification would replace
the electro-hydraulic MFIV actuators with system-medium actuators to
improve MFIV reliability and reduce maintenance requirements. The MFIV
stroke time would be increased. A plant modification would also replace
swing check valves in each auxiliary feedwater (AFW) motor-driven pump
discharge line with an automatic recirculation control (ARC) check
valve to reduce the potential for vibration and increase AFW flow
margin.
[[Page 43395]]
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.
MFIV Actuator Replacement and Increased MFIV Stroke Time
[* * *], the increase in MFIV stroke time does not adversely
impact the NSSS [nuclear steam system supplier] design transients
evaluated for the Callaway Plant. The increase in MFIV stroke time
will result in a slightly longer normal post trip cool down.
Although the plant post trip cool down is expected to be slightly
longer for the increased MFIV stroke time, the plant response does
not significantly deviate from its current evaluated response
following a normal reactor trip.
Evaluations assessing the impact of the change in MFIV actuators
and the increase in MFIV stroke time on LOCA [loss-of-coolant
accident] mass and energy releases; main steamline break mass and
energy releases; LOCA and LOCA[-]related transients; non-LOCA
transients; LOCA hydraulic forces[;] and steam releases used for
radiological consequence calculations were also performed. The
increase in isolation time and change in MFIV actuators either do
not provide an adverse impact or have no impact. Except for the SGTR
[steam generator tube rupture)] with overfill accident, the results
presented in the FSAR [Callaway Final Safety Analysis Report] remain
valid. The increase in MFIV stroke time was evaluated for impact on
the SGTR with overfill accident. [* * *], the results from the re-
analysis of the SGTR with overfill accident confirm that there is no
significant increase in the probability or consequences of an
accident previously evaluated.
The replacement of the existing electro-hydraulic MFIV actuators
with system-medium actuators and the increase in MFIV stroke time
from 5 seconds to 15 seconds will not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
MDAFP [Motor Driven Auxiliary Feedwater Pump] ARC Valve and
Increased Maximum AFW Flow
The replacement of existing MDAFP discharge check valves with
the ARC valves results in increased maximum AFW flow to the steam
generators [(SGs)]. In many accident scenarios the increase in AFW
flow to the SGs is beneficial to mitigation of the event. The
evaluations [* * *] demonstrate that in those accident scenarios
where maximum AFW flow is limiting, except for the SGTR with
overfill accident, the increase in AFW flow remains bounded by FSAR
analyses. The increase in maximum AFW flow was evaluated for impact
on the SGTR with overfill accident. [* * *], the results from the
re-analysis of the SGTR with overfill accident confirm that there is
no significant increase in the probability or consequences of an
accident previously evaluated. The AFW system is not the initiator
of any accident and there is no possibility of a significant
increase in the probability of an accident or malfunction previously
evaluated.
Use of the ARC valve is an enhancement and the associated
increase in the maximum AFW flow will not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
Use of Revised Methods in Re-Analysis of SGTR With Overfill
The re[-]analysis of the design basis accident for SGTR with
overfill does not significantly increase the probability or
consequences of an accident previously evaluated. The re-analysis of
an accident is not an initiator [of an accident]. The SGTR accident
is classified as an ANS [American Nuclear Society] Condition IV
Event, Limiting Faults, and is only postulated and not expected to
occur. The re[-]analysis activity being evaluated does not change
the ANS classification for this design basis event. The re-analysis
does provide dose consequences that are minimal increases to the
doses in the Analysis of Record.
However, the doses remain well below regulatory limits. In
support of this methodology the proposed TS definition for DOSE
EQUIVALENT I-131 will allow the use of ICRP 30 based DCFs [dose
conversion factors]. Section 4.1.2 and [* * *] Appendix E of
Regulatory Guide 1.195 find acceptable and recommend the [proposed]
method revisions.
In summary, using the proposed revised methods for the re-
analysis of the SGTR with overfill does not result in a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
MFIV Actuator Replacement and Increased MFIV Stroke Time
The change in MFIV actuators and associated increase in MFIV
stroke time will not prevent the main feedwater or auxiliary
feedwater systems from performing their safety functions. The
proposed increase will not affect the normal method of plant
operation. No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of the increase. Although the modification does alter the design of
the MFIV actuators, it does not prevent the main feedwater or AFW
systems from performing their safety functions.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
MDAFP ARC Valve and Increased Maximum AFW Flow
The new MDAFP ARC valve and associated increase in the maximum
AFW flow [* * *] will not prevent the AFW system from performing its
safety function. The proposed increase in AFW system flow margin
will not effect the normal method of plant operation. No new
accident scenarios, transient precursors, failure mechanisms, or
limiting single failures are introduced as a result of the increase
in AFW system flow margin. Although the modification alters the
design of the MDAFP discharge check valves, it does not prevent the
AFW system from performing its safety functions.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
Use of Revised Methods in Re-Analysis of SGTR With Overfill
The revision to the Technical Specifications to allow the use of
ICRP 30[-]based DCFs is based on methodologies found acceptable to
the NRC and recommended for use as described in Section 4.1.2 of
Regulatory Guide 1.195. The re [-]analysis of the design basis
accident for SGTR with overfill and the use of recommended analysis
methods acceptable to the NRC does not introduce the possibility of
a new accident. Accident re-analysis is not an initiator of any
accident and no new failure modes are introduced. In summary, there
is no increase in the possibility of an accident of a different
type.
Therefore, the proposed changes do not create a new or different
kind of accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
MFIV Actuator Replacement and Increased MFIV Stroke Time
The replacement of the MFIV actuator and the associated increase
in the MFIV stroke time does not affect the manner in which safety
limits or limiting safety system settings are determined, nor will
there be any adverse effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
significant impact on the overpower limit, departure from nucleate
boiling ratio limits, heat flux hot channel factor (FQ),
nuclear enthalpy rise hot channel factor (F-delta-H), loss[-]of[-
]coolant accident peak cladding temperature (LOCA PCT), peak local
power density, or any other margin of safety. The radiological dose
consequence acceptance criteria listed in the [NRC] Standard Review
Plan will continue to be met.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
MDAFP ARC Valve and Increased Maximum AFW Flow
The use of the MDAFP ARC valve and the associated increase in
AFW system flow margin does not affect the manner in which safety
limits or limiting safety system settings are determined nor will
there be any adverse effect on those plant systems necessary to
assure the accomplishment of protection functions. There will be no
significant impact on the overpower limit, departure from nucleate
boiling ratio limits, heat flux hot channel factor (FQ),
nuclear enthalpy rise hot channel factor (F-delta-H), loss[-]of[-
]coolant accident peak cladding temperature (LOCA PCT), peak local
power density, or any other margin of safety. The radiological dose
[[Page 43396]]
consequence acceptance criteria listed in the Standard Review Plan
will continue to be met.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
Use of Revised Methods in Re-Analysis of SGTR With Overfill
Use of revised methods in the re-analysis for the SGTR with
overfill accident does not affect the manner in which safety limits
or limiting safety system settings are determined nor will there be
any adverse effect on those plant systems necessary to assure the
accomplishment of protection functions. There is no significant
impact on the overpower limit, departure from nucleate boiling ratio
limits, heat flux hot channel factor (FQ), nuclear
enthalpy rise hot channel factor (F-delta-H), loss[-]of[-]coolant
accident peak cladding temperature (LOCA PCT), peak local power
density, or any other margin of safety. The radiological dose
consequence acceptance criteria listed in the Standard Review Plan
will continue to be met. The re-analysis of the SGTR with overfill
confirms that both the thermal-hydraulic and radiological
consequences are within the regulatory requirements and does not
result in a significant reduction in a margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: June 27, 2003.
Description of amendment request: The amendment would revise the
Technical Specifications (TSs) to (1) extend the allowed outage time
(AOT) or required action completion time (CT) for an inoperable diesel
generator (DG) by adding the phrase ``OR 108 hours once per cycle for
each DG'' to the completion time for Required Action B.4 in TS 3.8.1,
``AC Sources--Operating,'' and (2) delete the second CT given in
certain required actions in TS 3.6.6, ``Containment Spray and Cooling
Systems''; TS 3.7.5, ``Auxiliary Feedwater (AFW) System''; TS 3.8.1;
and TS 3.8.9, ``Distribution System--Operating,'' of the TSs. The
second part would also delete Example 1.3-3, delete text referring to
this example, and re-number the remaining examples in TS 1.3,
``Completion Times.''
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.
DG AOT/CT Extension
The proposed change to extend the DG AOT/CT from 72 hours to 108
hours for planned, on-line maintenance does not affect the design of
the DGs, the operational characteristics or function of the DGs, the
interfaces between the DGs and other plant systems, or the
reliability of the DGs. The DGs mitigate the consequences of
previously evaluated accidents including loss[-] of[-]offsite power,
but as such are not themselves initiators of any previously
evaluated accidents. DG allowed outage time is thus not associated
with any initiating condition for accidents previously evaluated.
The consequences of an accident are independent of the time the DGs
are out of service as long as adequate DG availability is assured.
The proposed changes will not result in a significant decrease in DG
availability, so assumptions regarding DG availability are not
impacted. Since the DGs will continue to be capable of performing
their accident mitigation function as assumed in the accident
analysis, the consequences of accidents previously analyzed are
unchanged with respect to the proposed changes.
In addition, to fully evaluate the effect of the proposed DG
completion time extension, probablistic risk assessment methods and
a deterministic analysis were utilized. The results of the analyses
show no significant increase in core damage frequency or large early
release frequency.
Elimination of Second Completion Times
Similar to the above change, the changes to eliminate the
``second'' Completion Times from the affected Technical
Specifications [(i.e., the specific TS sections being changed)] do
not affect the design, operational characteristics, or intended
functions of the equipment addressed by those Technical
Specifications. With no direct effects on that equipment (or any
other plant equipment or features), allowed equipment outage times
are not associated with any initiating condition for any accident
previously evaluated, and therefore would not affect the probability
of such accidents. Further, eliminating these Completion Times is
not expected to have an adverse effect on the availability of the
applicable systems or components because equipment availability
performance criteria required for conformance to the Maintenance
Rule impose an equivalent or acceptable level of control and
management of equipment availability regardless of such Completion
Times. As noted above, the consequences of evaluated accidents are
independent of mitigating equipment allowed outage times as long as
adequate availability of the equipment is ensured. Since elimination
of the second Completion Times has no significant impact on
equipment availability (in light of continued, required conformance
to the Maintenance Rule), the consequences of accidents previously
evaluated are unchanged.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
None of the proposed changes, i.e., neither the DG AOT extension
nor the elimination of [the] second Completion Times, involve a
change in the design, configuration, or operational characteristics
of the plant. No physical alteration of the plant is involved, as no
new or different type of equipment is to be installed. The changes
do not alter any assumptions made in the safety analyses, and no
alteration in the procedures which ensure that the plant remains
within analyzed limits is being proposed. As such, no new failure
modes or mechanisms that could cause a new or different kind of
accident from any previously evaluated are being introduced.
Therefore, the proposed changes do not create the possibility of
a new or different accident from any accident previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed DG AOT extension and elimination of second
Completion Times do not alter the manner in which safety limits or
limiting safety system settings are determined. The safety analysis
acceptance criteria are not impacted by [these] change[s], and the
proposed changes will not permit plant operation in a configuration
[that is] outside the design basis.
Further, with regard to plant risk, the risk assessment
performed for the DG AOT extension determined that the quantifiable
increase in plant risk is acceptably small. Likewise, for the
elimination of [the] second Completion Times, it may be assumed that
this change also involves little or no increase in risk on the basis
that required, continued compliance with the Maintenance Rule
provides adequate controls for maintaining equipment availability
regardless of the second Completion Times [proposed to be
eliminated].
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
[[Page 43397]]
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Union Electric Company, Docket No. 50-483, Callaway Plant, Unit 1,
Callaway County, Missouri
Date of application request: June 27, 2003.
Description of amendment request: The licensee is proposing to
amend the operating license for the Callaway Plant to allow plant
modifications in order to facilitate maintenance on the replacement
steam generators (SGs) to be installed in Refueling Outage (RO) 14
(Fall 2005). The proposed modifications (1) replace the existing sludge
lance platforms with new platforms to provide a larger platform area
around each SG, and (2) cut a permanent access opening through the
secondary shield wall to improve access to the sludge lance platforms.
They are to be done in RO 13 (Spring 2004). Dynamic effects associated
with large reactor coolant system (RCS) branch line ruptures are to be
excluded using a proposed leak-before-break (LBB) methodology. The
amendment would authorize changes to the Callaway licensing basis to be
added to the Callaway Final Safety Analysis Report (FSAR). There are no
proposed changes to the Technical Specifications. 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.
Overall protection system performance will remain within the
bounds of the previously performed accident analyses. The design of
the protection systems will be unaffected. The reactor protection
system and engineered safety feature actuation system will continue
to function in a manner consistent with the plant design basis. All
design, material, and construction standards that were applicable
prior to the request are maintained.
Neither the currently intact ``C'' SG cubicle secondary shield
wall, nor the proposed configuration that provides a permanent
access opening, create accident initiation mechanisms that would
increase the probability of an accident. There will be no change to
normal plant operating parameters or accident mitigation
performance.
The proposed amendment will not alter any assumptions or change
any mitigation actions in the radiological consequence evaluations
in the FSAR.
Therefore, the proposed amendment does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
There are no changes in the method by which any safety-related
plant system performs its safety function. This amendment will not
affect the normal method of power operation or change any operating
parameters. No performance requirements will be affected, but SG
maintenance access will be greatly improved.
No new accident scenarios, transient precursors, failure
mechanisms, or limiting single failures are introduced as a result
of this amendment. There will be no adverse effect or challenges
imposed on any safety-related system as a result of this amendment.
Presence of a permanent access opening in the ``C'' loop SG
secondary shield wall does not, of itself, create the possibility of
a new accident since the secondary shield walls are not used for
missile protection and the high-energy line breaks (greater than 10-
inches in diameter) that would generate missiles will be removed
from the structural design basis after NRC's review and acceptance
of the LBB topical reports.
The proposed amendment does not alter the design or performance
of the 7300 Process Protection System, Nuclear Instrumentation
System, or Solid State Protection System used in the plant
protection systems.
Therefore, the proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
There will be no effect on the manner in which safety limits or
limiting safety system settings are determined nor will there be any
effect on those plant systems necessary to assure the accomplishment
of protection functions. There will be no impact on the overpower
limit, departure from nucleate boiling ratio (DNBR) limits, heat
flux hot channel factor (FQ), nuclear enthalpy rise hot
channel factor (F[Delta]H), loss[-]of[-]coolant accident peak
cladding temperature (LOCA PCT), or peak local power density. The
LBB margins discussed in NUREG-1061 Volume 3 are satisfied. The
radiological dose consequence acceptance criteria listed in the
[NRC] Standard Review Plan will continue to be met. The secondary
shield walls are not fission product barriers. They provide
radiation shielding to maintain occupational exposure ALARA [as low
as is reasonably achievable] and provide structural support to
primary coolant SSCs [structures, systems, and components].
The proposed amendment does not eliminate any surveillances or
alter the Frequency of surveillances required by the Technical
Specifications. The nominal Reactor Trip System (RTS) and Engineered
Safety Features Actuation System (ESFAS) trip setpoints (TS Bases
Tables B 3.3.1-1 and B 3.3.2-1), RTS and ESFAS allowable values (TS
Tables 3.3.1-1 and 3.3.2-1), and the safety analysis limits assumed
in the transient and accident analyses (FSAR Table 15.0-4) are
unchanged. None of the acceptance criteria for any accident analysis
is changed.
Therefore, the proposed amendment does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: John O'Neill, Esq., Shaw, Pittman, Potts &
Trowbridge, 2300 N Street, NW., Washington, DC 20037.
NRC Section Chief: Stephen Dembek.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and 2, Louisa County, Virginia
Date of amendment request: March 28, 2002, as supplemented by
letters dated May 13, June 19, and November 15, 2002, and May 6, May 9,
May 27, and June 11 (2 letters), 2003. This notice supersedes the
notice that was published on May 14, 2002 (67 FR 34496).
Description of amendment request: The proposed amendments would
permit Virginia Electric and Power Company to replace the existing
Westinghouse fuel with Framatome ANP Advanced Mark-BW fuel at North
Anna Power Station, Units 1 and 2. This submittal was accompanied by
requested exemptions from the requirements of 10 CFR 50.44 and 10 CFR
50.46. These exemptions will be processed separately.
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 probability of occurrence or the consequences of an
accident previously evaluated is not significantly increased.
The proposed methodology has been generically reviewed and
approved for use by the NRC for determining core operating limits
prior to its use by Dominion. Analyzed events are assumed to be
initiated by the failure of plant structures, systems, or
components. The core operating limits developed in accordance with
the new methodologies will be bounded by any limitations in the NRC
safety evaluation report (SER) for the new methodologies.
Application of the topical reports associated with the new
methodologies will demonstrate that the integrity of the fuel will
be maintained during normal operations and that design requirements
will continue to be met. The proposed changes do not involve
physical changes to any plant structure, system, or component.
Therefore, the
[[Page 43398]]
probability of occurrence of any accident previously evaluated 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 fuel 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 changes do not affect the
performance of any equipment used to mitigate the consequences of an
analyzed accident. As a result, no analysis assumptions are violated
and there are no adverse effects on the factors that contribute to
offsite or onsite dose resulting from an accident. The proposed
changes do not affect setpoints that initiate protective or
mitigative actions. The proposed changes ensure that plant
structures, systems, and components are maintained consistent with
the safety analysis and licensing basis. Based on this evaluation,
there is no significant increase in the consequences of a previously
analyzed event.
2. The possibility for a new or different type of accident from
any accident previously evaluated is not created.
The proposed changes do not involve any physical alteration of
plant systems, structures, or components, other than allowing for
fuel design in accordance with NRC-approved methodologies. The
proposed methodologies continue to meet applicable criteria for
LBLOCA [large-break loss-of-coolant accident] and SBLOCA [small-
break loss-of-coolant accident] analyses. No new or different
equipment is being installed. No installed equipment is being
operated in a different manner. There is no alteration 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. There are no
changes in the methods governing normal plant operation, nor are the
methods utilized in response to plant transients changed. Therefore,
the possibility for a new or different kind of accident from any
accident previously evaluated is not created.
3. The margin of safety is not significantly reduced.
The margin of safety is established through the design of the
plant structures, systems, and components, through the parameters
within which the plant is operated, through the establishment of
setpoints for the actuation of equipment relied upon to respond to
an event, and through margins contained within safety analyses. The
proposed changes in the methodologies used in the LBLOCA and SBLOCA
analyses do not impact the condition or performance of structures,
systems, setpoints, and components relied upon for accident
mitigation. The proposed changes in the analysis methodologies
comply with the requirements of 10 CFR 50.46 paragraph (a)(1)(i)
(i.e., not exceeding a peak cladding temperature of 2200[deg]F for
[SB] LOCA and a high probability that peak cladding temperature will
remain below 2200[deg]F for [LB] LOCA). Therefore, the margin of
safety as defined in the Bases to the North Anna Units 1 and 2
Technical Specifications is not significantly reduced.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Ms. Lillian M. Cuoco, Esq., Senior Counsel,
Dominion Resources Services, Inc., Millstone Power Station, Building
475, 5th Floor, Rope Ferry Road, Rt. 156, Waterford, Connecticut 06385.
NRC Section Chief: John A. Nakoski.
Notice of Issuance of Amendments to Facility Operating Licenses
During the period since publication of the last biweekly notice,
the Commission has issued the following amendments. The Commission has
determined for each of these amendments that the application complies
with the standards and requirements of the Atomic Energy Act of 1954,
as amended (the Act), and the Commission's rules and regulations. The
Commission has made appropriate findings as required by the Act and the
Commission's rules and regulations in 10 CFR Chapter I, which are set
forth in the license amendment.
Notice of Consideration of Issuance of Amendment to Facility
Operating License, Proposed No Significant Hazards Consideration
Determination, and Opportunity for A Hearing in connection with these
actions was published in the Federal Register as indicated.
Unless otherwise indicated, the Commission has determined that
these amendments satisfy the criteria for categorical exclusion in
accordance with 10 CFR 51.22. Therefore, pursuant to 10 CFR 51.22(b),
no environmental impact statement or environmental assessment need be
prepared for these amendments. If the Commission has prepared an
environmental assessment under the special circumstances provision in
10 CFR 51.12(b) and has made a determination based on that assessment,
it is so indicated.
For further details with respect to the action see (1) the
applications for amendment, (2) the amendment, and (3) the Commission's
related letter, Safety Evaluation and/or Environmental Assessment as
indicated. All of these items are available for public inspection at
the Commission's Public Document Room, located at One White Flint
North, Public File Area 01F21, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible from
the Agencywide Documents Access and Management Systems (ADAMS) Public
Electronic Reading Room on the internet at the NRC web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units Nos.
1, 2, and 3, Maricopa County, Arizona
Date of application for amendments: April 15, 2003.
Brief description of amendments: The amendments revise Sections
2.2, ``SL [Safety Limits] Violations,'' for reporting such violations
to positions in the plant organization; 5.2.1, ``Onsite and Offsite
Organization,'' for the position responsible for overall safe plant
operation; and 5.5.1, ``Offsite Dose Calculation Manual (ODCM),'' to
replace the positions of Vice President, Nuclear Production, and
Director, Site Chemistry, with other positions in the plant
organization. Also, there would be the format change of adding the
title of Section 2.2 near the top of TS page 2.0-2.
Date of issuance: June 26, 2003.
Effective date: June 26, 2003, and shall be implemented within 30
days of the date of issuance.
Amendment Nos.: Unit 1-146, Unit 2-146, Unit 3-146.
Facility Operating License Nos. NPF-41, NPF-51, and NPF-74: The
amendments revised the Technical Specifications.
Date of initial notice in Federal Register: May 27, 2003 (68 FR
28845).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 26, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, et al., Docket Nos. 50-413 and 50-414, Catawba
Nuclear Station, Units 1 and 2, York County, South Carolina
Date of application for amendments: August 26, 2002, as
supplemented by letter dated June 18, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications regarding the Diesel Fuel Oil Testing Program.
Date of issuance: July 10, 2003.
Effective date: As of the date of issuance and shall be implemented
[[Page 43399]]
within 90 days from the date of issuance.
Amendment Nos.: 206 & 200.
Facility Operating License Nos. NPF-35 and NPF-52: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66008).
The supplement dated June 18, 2003, provided clarifying information
that did not change the scope of the August 26, 2002, application nor
the initial proposed no significant hazards consideration
determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 10, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: December 12, 2002, as
supplemented by letters dated March 27 and April 23, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications (TS) regarding the reactor vessel pressure-
temperature limit curves and revise the low-temperature overpressure
protection limits. The licensee also requested that a change be made to
TS Table 3.3.2-1, Footnote (c) to correct what was claimed to be an
editorial error. This request was not supported by sufficient
information and, accordingly, is denied.
Date of issuance: July 3, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 214 & 195.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: January 7, 2003 (68 FR
801).
The supplement dated March 27 and April 23, 2003, provided
clarifying information that did not change the scope of the December
12, 2002, application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 3, 2003.
No significant hazards consideration comments received: No.
Duke Energy Corporation, Docket Nos. 50-369 and 50-370, McGuire Nuclear
Station, Units 1 and 2, Mecklenburg County, North Carolina
Date of application for amendments: August 26, 2002, as
supplemented by letter dated June 18, 2003.
Brief description of amendments: The amendments revise the
Technical Specifications regarding the Diesel Fuel Oil Testing Program.
Date of issuance: July 10, 2003.
Effective date: As of the date of issuance and shall be implemented
within 90 days from the date of issuance.
Amendment Nos.: 215 & 195.
Facility Operating License Nos. NPF-9 and NPF-17: Amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: October 29, 2002 (67 FR
66008).
The supplement dated August 26, 2002, provided clarifying
information that did not change the scope of the June 18, 2003,
application nor the initial proposed no significant hazards
consideration determination.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated July 10, 2003.
No significant hazards consideration comments received: No.
Entergy Operations, Inc., System Energy Resources, Inc., South
Mississippi Electric Power Association, and Entergy Mississippi, Inc.,
Docket No. 50-416, Grand Gulf Nuclear Station, Unit 1, Claiborne
County, Mississippi
Date of application for amendment: March 19, 2003.
Brief description of amendment: The amendment deletes Technical
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and License
Condition 2(C)(33)(c) from Facility Operating License NPF-29, thereby
eliminating the requirement to have and maintain the post-accident
sampling system at Grand Gulf Nuclear Station, Unit 1. The amendment
also addresses related changes to TS 5.5.2, ``Primary Coolant Sources
Outside Containment.''
Date of issuance: June 30, 2003.
Effective date: As of the date of issuance and shall be implemented
within 120 days of issuance.
Amendment No: 158.
Facility Operating License No. NPF-29: The amendment revises the
Technical Specifications and deletes License Condition 2(C)(33)(c).
Date of initial notice in Federal Register: May 13, 2003 (68 FR
25652).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated June 30, 2003.
No significant hazards consideration comments received: No.
Nuclear Management Company, LLC, Docket No. 50-305, Kewaunee Nuclear
Power Plant, Kewaunee County, Wisconsin
Date of application for amendment: January 13, 2003, supplements
dated February 27, March 6, March 14, April 30, June 9, and June 30,
2003.
Brief description of amendment: The proposed amendment would revise
the Kewaunee Nuclear Power Plant Facility Operating License and
Technical Specifications to increase the licensed rated power by 1.4
percent from 1650 megawatts thermal to 1673 megawatts thermal using
measurement uncertainty recapture.
Date of issuance: July 8, 2003.
Effective date: As of the date of issuance and shall be implemented
within 120 days.
Amendment No.: 168.
Facility Operating License No. DPR-43: Amendment revised the
Facility Operating License and Technical Specifications.
Date of initial notice in Federal Register: February 4, 2003 (68 FR
5679).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated July 8, 2003.
No significant hazards consideration comments received: No.
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 application for amendments: January 29, 2003.
Brief description of amendments: The amendments revise Salem, Unit
No. 1, Technical Specifications (TSs) Section 3/4.7.6, and Salem, Unit
No. 2, TSs 3/4.2.2, 3/4.7.6, and Table 3.3-6. These changes are
administrative and editorial in nature, and correct errors made during
the implementation of previously-approved TS changes.
Date of issuance: June 26, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment Nos.: 258 and 239.
Facility Operating License Nos. DPR-70 and DPR-75: The amendments
revised the TSs.
Date of initial notice in Federal Register: April 15, 2003 (68 FR
18284).
[[Page 43400]]
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated June 26, 2003.
No significant hazards consideration comments received: No.
Dated at Rockville, Maryland, this 11th day of July 2003.
For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Acting Director, Division of Licensing Project Management, Office of
Nuclear Reactor Regulation.
[FR Doc. 03-18084 Filed 7-21-03; 8:45 am]
BILLING CODE 7590-01-P