[Federal Register Volume 68, Number 236 (Tuesday, December 9, 2003)]
[Notices]
[Pages 68654-68675]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 03-30246]
-----------------------------------------------------------------------
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, November 14, through November 26. The last
biweekly notice was published on November 25, 2003 (68 FR 66131).
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 January 8, 2004, the licensee may file a request for a hearing
with respect to issuance of the amendment to the subject facility
operating license and any person whose interest may be affected by this
proceeding and who wishes to participate as a party in the proceeding
must file a written request for a hearing and a petition for leave to
intervene. Requests for a hearing and a petition for leave to intervene
shall be filed in accordance with the Commission's ``Rules of Practice
for Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested
persons should consult a current copy of 10 CFR 2.714, 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
[[Page 68655]]
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].
AmerGen Energy Company, LLC, Docket No. 50-289, Three Mile Island
Nuclear Station, Unit 1 (TMI-1), Dauphin County, Pennsylvania
Date of amendment request: August 6, 2003.
Description of amendment request: This amendment would revise the
Technical Specifications (TSs) to incorporate reference to the 10 CFR
50.55a, Codes and Standards, criteria for the inservice reactor
building tendon surveillance requirements, to incorporate an
administrative change to the TS Definition 1.22 to be consistent with
10 CFR 20.1003, as well as other administrative corrections from
previously issued TS amendments.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed revision to Technical Specification 4.4.2.1 and
associated Bases Section incorporates reference to the criteria of
10 CFR 50.55a, ``Codes and standards,'' in addition to the existing
criteria of Regulatory Guide 1.35. This change provides consistency
between the Technical Specification tendon surveillance program
criteria and the regulatory requirements specified in 10 CFR
50.55a(b)(2)(vi). These
[[Page 68656]]
regulatory requirements and the associated surveillance program
ensure that the reactor building tendon prestressing system is
capable of maintaining the structural integrity of the containment
during operating and accident conditions. The reactor building
prestressing system is not an initiator of any accident. Therefore,
this change is not related to the probability of any accident
previously evaluated. This change ensures that the containment
tendon surveillance program addresses the appropriate regulatory
criteria. This change does not result in any reduction in the
effectiveness of the existing surveillance program. The tendon
surveillance program will continue to ensure that the containment
structure is capable of performing its intended safety function in
the event of a design basis accident. Therefore, this change has no
affect on the consequences of an accident previously evaluated.
The proposed changes to Technical Specification Definition 1.22,
Technical Specification 3.1.6.6 and associated Bases, and Technical
Specification 3.24 Bases are only administrative changes or
corrections and have no affect on plant design or operations.
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 revision to Technical Specification 4.4.2.1 and
associated Bases Section incorporates references to the criteria of
10 CFR 50.55a, ``Codes and standards,'' in addition to the existing
criteria of Regulatory Guide 1.35. This change provides consistency
between the Technical Specification tendon surveillance program
criteria and the regulatory requirement specified in 10 CFR
50.55a(b)(2)(vi). The proposed Technical Specification change does
not result in any reduction in effectiveness of the existing tendon
surveillance program. The tendon surveillance program will continue
to satisfy the applicable Technical Specification and regulatory
required criteria, thus ensuring that the containment structure will
perform its design safety function. This change has no affect on the
design and operation of plant structures, systems, and components.
This change does not introduce any new accident precursors and does
not involve any alterations to plant configurations, which could
initiate a new or different kind of accident.
The proposed changes to Technical Specification Definition 1.22,
Technical Specification 3.1.6.6 and associated Bases, and Technical
Specification 3.24 Bases are only administrative changes or
corrections and have no affect on plant design or operations.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed revision to Technical Specification 4.4.2.1 and
associated Bases Section incorporates reference to the criteria of
10 CFR 50.55a, ``Codes and standards,'' in addition to the existing
criteria of Regulatory Guide 1.35. This change provides consistency
between the Technical Specification tendon surveillance program
criteria and the regulatory requirement specified in 10 CFR
50.55a(b)(2)(vi). The containment examination and inspection
requirements specified in 10 CFR 50.55a(b)(2)(vi) meet the same
standards as the criteria specified in Regulatory Guide 1.35. The
proposed Technical Specification change does not result in any
reduction in effectiveness of the existing tendon surveillance
program. The tendon surveillance program will continue to satisfy
the applicable Technical Specification and regulatory required
criteria, thus ensuring that the containment structure will perform
its design safety function in accordance with existing margins of
safety for containment integrity.
The proposed changes to Technical Specification Definition 1.22,
Technical Specification 3.1.6.6 and associated Bases, and Technical
Specification 3.24 Bases are only administrative changes or
corrections and have no affect on plant design or operations.
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: Edward J. Cullen, Jr., Esquire, Vice President,
General Counsel and Secretary, Exelon Generation Company, LLC, 300
Exelon Way, Kennett Square, PA 19348 NRC Section Chief: Richard J.
Laufer.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona
Date of amendments request: August 22, 2003.
Description of amendments request: The amendments would revise
three different sections in the Updated Final Safety Analysis Report
(UFSAR) for PVNGS [Palo Verde Nuclear Generating Station], Units 1, 2,
and 3. This request would revise the sections of the UFSAR which
describe the maximum fuel pin pressurization criteria used for fuel
handling accident safety analyses. This change is necessitated due to
the combination of higher core burnup designs, fuel which contains
erbia poison, and the recent introduction of ZIRLO cladded fuel to the
PVNGS reactors.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed change would revise sections of the PVNGS UFSAR,
which describe the maximum fuel pin pressurization criteria used for
fuel handling accident safety analyses.
No additional equipment is being added as a result of the
proposed change. None of the failure modes and effects analyses are
impacted by the proposed change since no structures, systems, or
components (SSCs) are being modified, system lineups remain the
same, and operator actions for fuel handling accident are not
changing. No manual actions are being substituted for automatic
actions. The SSCs relied upon to mitigate the event are not
changing. Specifically, the fuel building, BOPESFAS (Balance of
Plant-Engineered Safety Features Actuation System), radiation
monitor setpoints, etc. . . are not impacted. The methodology
changes will have no impact on the likelihood of a malfunction of
any SSCs.
No departures from the design or testing and performance
standards outlined in any 10 CFR [Part] 50, Appendix A, General
Design Criteria (GDC) will result from the proposed activity. The
proposed UFSAR changes will not make any SSCs more likely to fail
(no direct effects). Even with higher fuel pin pressures, the use of
ZIRLO cladding provides more margin to design stress limits (liftoff
pressure) than Zircaloy cladding. Regardless of the fuel type (and
hence cladding type), the design stress and code allowable limits
will not be exceeded. Palo Verde Nuclear Generating Station (PVNGS)
``Fuel Mishandling Accident Evaluation with ZIRLO Fuel Rods''
concluded that the analysis of record for fuel handling events
involving fuel assemblies containing ZIRLO cladding would remain
bounding. No physical changes to any SSCs will be performed as a
result of the proposed changes. In addition, system/equipment
redundancy requirements are maintained with the proposed UFSAR
changes.
Fuel handling accident analyses must ensure doses at the site
boundary and control room remains well within 10 CFR Part 100 and 10
CFR [Part] 50 Appendix A, GDC 19 exposure guideline. Restricting the
peak assembly average fuel pin pressure to <1200 psig will still
result in acceptable doses. Therefore, no indirect effects on SSCs
associated with dose limitations are impacted.
Consequences mean dose at the Exclusion Area Boundary (EAB), Low
Population Zone (LPZ), and Control Room; therefore, an increase in
consequences must involve an increase in radiological doses to the
public or to control room operators. No changes to the
[[Page 68657]]
dose exposure as a result of a fuel handling accident are proposed
for the methodology change and Regulatory Guide 1.25 deviation
requested. Therefore, there are no radiological consequence changes
for this event.
The fuel handling accident event does involve fuel barrier
failure and does involve consequences, however no changes to the
fuel handling dose calculation are required since the
decontamination factor will remain unchanged even with maximum fuel
pin pressure exceeding 1200 psig. Activities affecting on-site dose
consequences that may require prior NRC approval are those that
impede required actions inside or outside the control room to
mitigate the consequences of reactor accidents.
The proposed change does not modify any operator actions and
hence will not impede required actions inside or outside the control
room to mitigate the consequences of reactor accidents. The proposed
change will not prevent or degrade the effectiveness of actions
described or assumed in an accident discussed in the UFSAR. The
proposed change does alter assumptions previously made in evaluating
the radiological consequences of an accident described in the UFSAR,
however the altered assumption is a methodology change. If the
proposed methodology change were not applied, the calculated dose
would increase. The peak assembly average pin pressure concept would
allow the decontamination factor (DF) to remain the same and
therefore consequences would remain unchanged. The proposed change
does not play a direct role in mitigating the radiological
consequences of an accident described in the UFSAR. The radiological
consequences of the accident described in the UFSAR are bounding for
the proposed activity (e.g., the results of the UFSAR analysis bound
those that would be associated with the proposed change).
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 accident affected by the proposed change is the fuel
handling accident (UFSAR Section 15.7.4). The proposed change does
not involve any new equipment and does not operate any existing
equipment in a different or more severe manner than what has
previously been analyzed. PVNGS evaluations concluded all analyses
of record for fuel handling events involving fuel assemblies
containing ZIRLO cladding will remain bounding. The material
strength of ZIRLO is significantly higher than that for Zircaloy-4.
Since the allowable stresses for ZIRLO cladding are significantly
higher than for Zircaloy-4, the same number of fuel rods (or fewer)
will be damaged by the same accident scenarios as previously
evaluated. Regardless of the fuel type (and hence cladding type),
the design stress and code allowable limits will not be exceeded.
Slight changes in the maximum fuel pin pressure during fuel movement
will have no impact on the possibility of creating an accident of a
different type as long as the design pressure structural limits of
the fuel assembly are not approached. PVNGS calculation documents
minimum liftoff pressures will not be challenged regardless of the
fuel type or cladding type. Maintaining peak assembly average fuel
pin pressure below 1200 psig will not challenge the liftoff pressure
design basis limit for the cladding. The peak pin internal pressures
for the hot rods never exceed the clad liftoff pressure and
therefore the fuel pins will not be more likely to fail. Vendor
calculation shows that the ZIRLO cladding fuel design results in a
greater margin to the design pressure limit of the fuel cladding and
also documents liftoff pressures are not exceeded for PVNGS fuel
designs.
The design function of the SSCs required to function during a
fuel handling accident is to provide protection to ensure fuel
damage is limited to 236 fuel pins (one fuel assembly) and ensuring
doses do not exceed established limits. These are indirect affects.
This change will not make a SSC more likely to fail (no direct
affects). In fact, ZIRLO cladding fuel is less likely to fail than
the original Zircaloy-4 cladding fuel. No physical changes to the
SSCs will be performed as a result of the proposed change. This
proposed change does not change the failure modes for the SSCs
required to operate for the fuel handling accident. The cladding
calculations document design stress or code allowable limits will
not be exceeded. Hence, system/equipment redundancy requirements are
maintained. Fuel handling accident analyses must ensure doses at the
site boundary remain within acceptable design limits. The cladding
calculations document fuel pin pressures do not exceed the design
pressure ratings for the fuel assembly. Therefore, no indirect
effects on SSCs associated fuel clad pressure boundary exist. None
of the failure modes and effects analyses are impacted by this
methodology change since no SSCs are being modified, system lineups
remain the same, and operator actions are not changing.
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 Nuclear Reactor Regulation (NRR) Safety Evaluation,
``Related to Task Interface Agreement 99-03 Regarding Potential
Nonconservative Assumptions for Fuel Handling Accident, McGuire
Nuclear Station, dated November 24, 1999,'' states in part, ``The
NRR staff has concluded that the increased rod pressures associated
with extended bumup fuel can be expected to decrease the value of
the iodine DF. However, the NRR staff believes that the iodine DF
value of 100 provided in Regulatory Guide 1.25 has sufficient margin
to compensate for the increases in rod gas pressures at current
allowable bumup levels and for the expected increases in gap release
fractions. Conservatisms in the assessment of the amount of fuel
damage provide additional margin. Design basis fuel handling
accidents are not considered to have a high risk significance. On
the basis of these findings, the staff concludes that there is
reasonable assurance that adequate protection of the public from the
effects of design basis fuel handling accidents involving fuel with
peak rod average bumups as high as 62 GWD/MTU will continue.''
To assess the margin of safety, the methodology specified in
Regulatory Guide 1.183, [``]Alternative Radiological Source Terms
for Evaluating Design Basis Accidents at Nuclear Power Reactors,['']
was evaluated. This regulatory guide suggests a DF of 200 for
iodine. This DF is well above the DF of 100 specified by Regulatory
Guide 1.25.
APS [Arizona Public Service] proposes that ample margin is
retained to justify the continued used [use] of an overall
decontamination factor of 100 at a peak assembly average fuel pin
pressure of 1200 psig.
Therefore, APS has concluded that the proposed license amendment
request does not involve a significant reduction in a margin of
safety.
Based on the above, APS concludes that the [activities
associated with] the proposed amendment(s) present no significant
hazards consideration under the standards set forth in 10 CFR 50.92
[``Issuance of Amendment,''] (c) and, accordingly, a finding of ``no
significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Stephen Dembek.
Arizona Public Service Company, et al., Docket Nos. STN 50-528, STN 50-
529, and STN 50-530, Palo Verde Nuclear Generating Station, Units 1, 2,
and 3, Maricopa County, Arizona.
Date of amendments request: September 17, 2003.
Description of amendments request: The amendments would revise
sections of the Technical Specifications (TS) to support replacement of
the part length control element assemblies (PLCEAs) with a new design
that contains neutron absorber over the entire control section of the
CEA. The replacements are referred to as part strength control element
assemblies (PSCEAs). Additionally, a change is proposed to TS 3.1.5--
``Control Element Assembly (CEA) Alignment,'' Condition B, to eliminate
a potential condition which could cause an unwarranted plant shutdown.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the
[[Page 68658]]
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 physical difference between the 4-finger full strength
control element assemblies (FSCEAs) and the PSCEAs involves using
Inconel rather than B4C (boron carbide) over 100% of the
active control section of each CEA finger. In addition, the PSCEAs
use Inconel tubing to encase solid Inconel slugs, which cover the
entire control section of the control element assembly (CEA). The
current PLCEAs (also have only 4-fingers) use solid Inconel rods for
only the lower half of each finger and B4C pellets in the
top 15 inches (10%) of the control section of the CEA. Although
failure of the solid Inconel region due to neutron fluence would be
less likely than a typical clad design, the differences in swelling
between the Inconel slugs encased by Inconel clad for the PSCEAs
will be minor and result in a minimal impact on clad integrity. With
the exception of the neutron absorber, the cladding design used for
the PSCEAs is similar to the cladding of the full strength CEAs
(FSCEAs). The geometry, cladding materials, and the spider assembly
that supports the CEA fingers are essentially the same for the 4-
finger FSCEAs and the PSCEAs. The principal difference results from
the Inconel slugs contained in the PSCEAs being heavier than the
B4C pellets used in the FSCEAs. Even though the weight of
a 4-finger PSCEA is greater than the weight of a 4-finger PLCEA or a
4-finger FSCEA, this weight difference is bounded by the 12-finger
FSCEAs which are operated by the same CEA drive mechanism system.
The PSCEAs use Inconel as a neutron absorber in the entire
control section of each CEA finger and will be operationally used
the same way as the PLCEAs. In particular, the insertion restraints
that are defined by the power dependent insertion limits (PDILs) for
the PLCEAs will remain the same for the PSCEAs. This existing
requirement will not result in any significant operational impact on
the PSCEAs since the solid Inconel cylinder in the bottom 50%
(operating range of the PDILs) of the PLCEAs has essentially the
same reactivity worth as that of the PSCEAs.
In addition, renaming the full length CEAs and part length CEAs
to full strength CEAs and part strength CEAs, respectively, and
providing definition for the PSCEAs will not impact the safe
operation of the plant. The terminology will be appropriately
changed in any related document, equipment tag, or indication on a
control panel.
The PLCEAs are not credited in the accident analyses for
accident mitigation. The PSCEA design eliminates an accident
scenario involving the insertion of a PLCEA past the PDIL, which
results in an axial shift in power due to the upper region of the
PLCEAs which has no neutron absorber. This condition will not occur
with the PSCEAs because they are filled with neutron absorber over
100% of the control section of each finger.
Concerning TS Limiting Condition for Operation (LCO) 3.1.5,
Condition B, proposed change; there are three position indicator
channels available for each CEA. Current TS Bases state that, ``At
least two of the following three CEA position indicator channels
shall be OPERABLE for each CEA.'' Additionally the TS Bases states,
``If only one CEA position indicator channel is OPERABLE, continued
operation in MODES 1 and 2 may continue, provided, within 6 hours,
at least two position indicator channels are returned to OPERABLE
status; or within 6 hours and once per 12 hours, verify that the CEA
group with the inoperable position indicators are either fully
withdrawn or fully inserted while maintaining the insertion limits
of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8.'' The TS Bases make no
restriction or condition limiting only one CEA within a subgroup to
having only one CEA position indication channel. Current analyses
already assume that more than one CEA in a subgroup could have only
one position indicator OPERABLE. Modifying the wording for Condition
B, of LCO 3.1.5, will not affect the likelihood or consequences of a
CEA drop, slip, ejection, or misalignment. This change will still
require at least one position indication channel be available for
each CEA.
Consequently, the proposed change does not involve a significant
increase in the probability or consequences of an accident.
2. The proposed change does not create the possibility of a new
or different kind of accident from any accident previously
evaluated.
The proposed changes do not introduce any new mode of plant
operation and the PSCEAs, like the PLCEAs, are not relied upon for
accident mitigation. The PSCEAs will be operated in exactly the same
manner in which the PLCEAs are operated. The existing operating
restrictions for the PLCEAs will apply to the PSCEAs. In particular,
the power dependent insertion limit (PDIL) restrictions identified
in the Core Operating Limits Report (COLR) will remain the same for
the PSCEAs. The PSCEA design uses Inconel over the entire control
section of each CEA finger, which will prevent the potential
undesired flux redistribution currently associated with the
misoperation of PLCEAs. Therefore, the analysis associated with the
undesired flux redistribution misoperation for the PLCEAs will be
eliminated from PVNGS [Palo Verde Nuclear Generating Station] safety
analyses. PSCEA misoperation events are bounded by the existing
PLCEA and FSCEA misoperation safety analyses.
In addition, renaming (within the Technical Specifications) the
``full length CEAs'' and ``part length CEAs'' to ``full strength
CEAs'' and ``part length or part strength CEAs,'' respectively, and
providing a definition for the PSCEAs will not impact the safe
operation of the plant. The terminology will be appropriately
changed in any related document, equipment tag, or indication on a
control panel.
Concerning TS LCO 3.1.5, Condition B proposed change, CEA
position indication channels have no control function and provide
input to the CEA Calculators (CEACs) and Core Protection Calculators
(CPCs) for generation of a penalty factor. This change will still
require at least one position indication channel be available for
each CEA. Allowing Condition `B' of LCO 3.1.5 to apply to more than
one CEA per group does not create the possibility of a different
type of malfunction than previously evaluated in the UFSAR [Updated
Final Safety Analysis Report].
Therefore, the proposed amendment will 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 design of the PSCEAs is very similar to the FSCEAs except
for the neutron absorber within each finger of a PSCEA. The PSCEAs
do not have as strong of a neutron absorber (Inconel) as that which
is contained in the FSCEAs (B4C). There is a weight
difference which results from the Inconel slugs contained in the
PSCEAs being heavier than the B4C pellets used in the
FSCEAs. Even though the weight of the 4-finger PSCEAs is greater
than the weight of the 4-finger PLCEAs, the CEA drive mechanism and
support components shall operate within their design bases.
Therefore, the PSCEAs can be considered adequate for safety-related
applications. Consequently, the differences in design between the
current PLCEAs and the PSCEAs do not adversely impact safe
operation.
The PLCEAs are not relied upon for shutdown margin or accident
mitigation and no new requirements will apply to the PSCEAs.
However, the design of the PSCEAs is effectively eliminating the
concern associated with the insertion of the PLCEAs past the PDILs
which could result in an undesirable shift in neutron flux to the
top of the core due to the region within the PLCEAs that do not have
neutron absorber. The PSCEAs have neutron absorber throughout their
entire control section, which prevents a neutron flux shift to the
top of the core if inserted past the PDIL, when compared to that of
the PLCEAs.
In addition, renaming the ``full length CEAs'' and ``part length
CEAs'' to ``full strength CEAs'' and ``part length or part strength
CEAs,'' respectively, and providing definition for the PSCEAs will
not impact the safe operation of the plant. The terminology will be
appropriately changed in any related document, equipment tag, or
indication on a control panel.
Concerning TS LCO 3.1.5, Condition B, proposed change, the
current licensing bases already consider having more than one CEA in
a CEA group with only one available position indication. The TS
Bases for LCO 3.1.5, Condition B state that, ``At least two of the
following three CEA position indicator channels shall be OPERABLE
for each CEA.'' Additionally the Bases states, ``If only one CEA
position indicator channel is OPERABLE, continued operation in MODES
1 and 2 may continue, provided, within 6 hours, at least two
position indicator channels are returned to OPERABLE status; or
within 6 hours and once per 12 hours, verify that the CEA group with
the inoperable position indicators are either fully
[[Page 68659]]
withdrawn or fully inserted while maintaining the insertion limits
of LCO 3.1.6, LCO 3.1.7 and LCO 3.1.8.'' The TS Bases make no
restriction or condition limiting only one CEA within a subgroup, to
having only one CEA position indication channel OPERABLE. Therefore,
modifying the wording for LCO 3.1.5, Condition B, does not involve a
significant reduction in the margin of safety since loss of
indication to more than one CEA is already considered in the
licensing bases.
Therefore, the proposed changes do not involve a significant
reduction in the margin of safety.
Based on the above, APS [Arizona Public Service] concludes that
the activities associated with the proposed amendment(s) present no
significant hazards consideration under the standards set forth in
10 CFR 50.92 [``Issuance of Amendment,''] (c) and, accordingly, a
finding of ``no significant hazards consideration'' is justified.
The NRC staff has reviewed the licensee's analysis and, based on
that review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
request for amendments involves no significant hazards consideration.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Stephen Dembek.
Arizona Public Service Company, et al. Docket Nos. STN 50-528, STN
50-529, and STN 50-530, Palo Verde Nuclear Generating Station,
Units 1, 2, and 3, Maricopa County, Arizona
Date of amendment request: October 7, 2003.
Description of amendment request: The licensee is proposing to
revise Technical Specification (TS) Section 5.5.6, ``Pre-Stressed
Concrete Containment Tendon Surveillance Program,'' for consistency
with the requirements of 10 CFR 50.55a(g)(4) for components
classified as Code Class CC. The proposed revision to TS 5.5.6 is to
indicate that the Containment Tendon Surveillance Program,
inspection frequencies, and acceptance criteria shall be in
accordance with Section XI, Subsection IWL of the American Society
of Mechanical Engineers Boiler and Pressure Vessel Code and the
applicable addenda as required by 10 CFR 50.55a, except where an
exemption or relief has been authorized by the NRC. The licensee has
also proposed to delete the provisions of Surveillance Requirement
3.0.2 from this specification.
In addition, the licensee is proposing to revise TS 5.5.16,
``Containment Leakage Rate Testing Program,'' to add exceptions to
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test
Program.''
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
The proposed changes would revise Technical Specification (TS)
Section 5.5.6, ``Pre-Stressed Concrete Containment Tendon
Surveillance Program,'' and Section 5.5.16, ``Containment Leakage
Rate Testing Program,'' for consistency with the requirements of 10
CFR 50.55a(g)(4) for components classified as Code Class CC. The
revised requirements do not affect the function of the containment
post-tensioning system components. The post-tensioning systems are
passive components whose failure modes could not act as accident
initiators or precursors. The improved inspections required by the
American Society of Mechanical Engineers (ASME) Code serve to
maintain containment response to accident conditions, by causing the
identification and repair of defects in the containment.
The proposed change affects the frequency of visual examinations
that will be performed for the concrete surfaces of the containment
for the purpose of the Containment Leakage Rate Testing Program. In
addition, the proposed change allows those examinations to be
performed during power operation as opposed to during a refueling
outage. The frequency of visual examinations of the concrete
surfaces of the containment and the mode of operation during which
those examinations are performed has no relationship to or adverse
impact on the probability of any of the initiating events assumed in
the accident analyses. The proposed change would allow visual
examinations that are performed pursuant to NRC approved ASME Code
Section XI requirements (except where relief has been granted by the
NRC) to meet the intent of visual examinations [as] required by
Regulatory Guide 1.163, ``Performance-Based Containment Leak-Test
Programs,'' without requiring additional visual examinations
pursuant to the Regulatory Guide. The intent of early detection of
deterioration will continue to be met by the more rigorous
requirements of the ASME Code[-]required visual examinations. As
such, the safety function of the containment as a fission product
barrier is maintained.
The proposed amendment does not impact any accident initiators,
analyzed events, or assumed mitigation of accident or transient
events. The proposed changes do not involve the addition or removal
of any equipment or any design changes to the facility.
Therefore, this proposed change does not represent 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 proposed change revises the Technical Specification
administrative controls programs for consistency with the
requirements of 10 CFR 50.55a(g)(4) for components classified as
Code Class CC. The function of the containment post-tensioning
system components are not altered by this change. The improved
inspections required by the American Society of Mechanical Engineers
(ASME) Code serve to maintain containment response to accident
conditions, by causing the identification and repair of defects in
the containment. In addition, the change affects the frequency of
visual examinations that will be performed for the concrete surface
containments. The proposed change also allows those examinations to
be performed during power operation as opposed to during a refueling
outage. Therefore, this change updates the Technical Specifications
to meet the current regulations and eliminates duplication of
requirements. The safety function of the containment as a fission
product barrier will be maintained.
Therefore, this proposed change does not create the possibility
of an accident of a different kind than previously evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The proposed change revises the improved Standard Technical
Specification administrative controls programs for consistency with
the requirements of 10 CFR 50.55a(g)(4) for components classified as
Code Class CC. The function of the containment post-tensioning
system components are not altered by this change. The change also
affects the frequency of visual examinations that will be performed
for the concrete surface containments. In addition, the proposed
change allows those examinations to be performed during power
operation as opposed to during a refueling outage. The change
ensures that containment integrity [will be maintained] and ensures
that the safety function of the containment as a fission product
barrier will be maintained.
Therefore, this 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
amendments request involves no significant hazards consideration.
Attorney for licensee: Kenneth C. Manne, Senior Attorney, Arizona
Public Service Company, P.O. Box 52034, Mail Station 7636, Phoenix,
Arizona 85072-2034.
NRC Section Chief: Stephen Dembek.
Dominion Nuclear Connecticut, Inc., Docket No. 50-245, Millstone Power
Station, Unit No. 1, New London County, Connecticut
Date of amendment request: September 18, 2003.
Description of amendment request: The licensee is proposing to
revise the Design Features Technical Specification 4.2, ``Fuel
Storage.'' The licensee's technical specification change implements the
following proposed changes:
(1) Eliminates all credit for Boraflex as a neutron absorber.
[[Page 68660]]
(2) Reduces the number of fuel assemblies allowed to be stored in
the spent fuel pool (SFP) from 3229 to 2959. The fuel will be
prohibited from being stored in 270 specific storage rack locations.
This is necessary to support the elimination of all credit for
Boraflex.
(3) Changes the required spent fuel pool keff to <=0.95.
This is necessary to support the elimination of all credit for
Boraflex.
(4) Eliminates the design features requirements on new fuel
storage, since Millstone Unit No. 1 (MP1) is a plant that has ceased
power operation and will no longer receive new fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed change does not involve a significant increase
in the probability or consequences of an accident previously
evaluated.
Accidents previously evaluated are the fuel handling accidents[,
as] described in the Decommissioned Safety Analysis Report (DSAR),
and a seismic event, which is considered as part of the spent fuel
rack design.
Since there are no changes to plant hardware, nor any changes in
how fuel is moved, there are no changes to the probability of a fuel
handling accident. The consequences of a fuel handling accident are
not affected, since none of the inputs to the fuel handling accident
is affected.
The proposed changes affect the criticality analysis of the
spent fuel storage racks. The spent fuel racks will continue to be
able to perform their design function, which is to maintain the
stored fuel in a sub-critical and cooled condition under all normal
and postulated accident conditions. There are no physical hardware
changes to the plant from these proposed changes. The revised
criticality analysis submitted with these proposed changes
demonstrates that fuel will be maintained in a sub-critical
condition during all normal and postulated accident conditions,
including the seismic event. Since there is no change in the ability
of the fuel storage racks to maintain a sub-critical condition due
to a seismic event, there is no change in the probability or
consequences of this accident.
Reducing the amount of fuel storage is a conservative action,
and the spent fuel racks were designed and licensed to allow empty,
partially filled, or completely full storage racks. Thus the fuel
racks will continue to be able to perform their design function to
maintain the fuel in a coolable condition.
The change to the new fuel storage racks is to delete the
Technical Specification requirements for the new fuel storage
keff limits. Since MP1 is a plant that has ceased power
operation and will no longer receive new fuel, there is no need for
these Technical Specification requirements. There are no new fuel
related accidents previously analyzed, therefore this change does
not involve a significant increase in the probability or
consequences of an accident previously evaluated.
In summary, the proposed changes do not involve an 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.
Since there are no changes to the plant equipment, there is no
possibility of a new or different kind of accident being initiated
or affected by equipment issues.
Reducing the number of fuel assemblies to be stored in the pool,
and discontinuing credit for Boraflex are conservative changes that
do not introduce any new or different kind of failure modes.
The changes made primarily affect the nuclear criticality
analysis and do not create a new or different kind of accident.
Changes in eliminating Boraflex credit, restricting fuel in certain
storage locations, and changing the allowable keff limit
are all impacts to the nuclear criticality analysis for the SFP. The
SFP criticality analysis is part of the basic design of the system
and is not an accident. The ability to maintain the SFP
keff less than or equal to 0.95, as well as within the 10
CFR Part 50 Appendix A, ``General Design Criteria for Nuclear Power
Plants,'' Criterion 62 ``Prevention of Criticality in Fuel Storage
and Handling'' (Reference 6) criteria of sub-critical, have been
evaluated. Criticality impacts are more appropriately discussed
under the margin of safety criterion.
The change to the new fuel storage racks is to delete the
Technical Specification requirements for the new fuel storage
keff limits. Since MP1 is a plant that has ceased power
operation and will no longer receive new fuel, there is no need for
these Technical Specification requirements. Since Millstone 1
currently has no new fuel and new fuel cannot be received, this
change does not create the possibility of a new or different kind of
accident from any accident previously evaluated.
In summary, the proposed changes do not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. The proposed change does not involve a significant reduction
in a margin of safety.
The margin of safety relevant to the SFP is defined as (1) SFP
keff remains sub-critical by an acceptable margin, and
(2) the spent fuel in the SFP remains adequately cooled so that the
fission product barriers remain intact.
The industry and regulatory accepted value for [the] required
sub-criticality margin[s] in the SFP is to ensure that the
keff of the SFP remains <=0.95 under all normal and
postulated accident conditions. This is documented in the Standard
Review Plan, Regulatory Guide 1.13, and ANSI/ANS-57.2, ``American
National Standard Design Requirements for LWR Spent Fuel Storage
Facilities at Nuclear Power Plants.'' The current MP1 Technical
Specifications require a more conservative value of 0.90 for SFP
keff. The proposed Design Features Technical
Specification changes the maximum SFP keff from 0.90 to
0.95. This is not a significant reduction in the margin to [of]
safety since the proposed value of 0.95 is consistent with the
accepted regulatory guidance for [the] sub-criticality margin. The
proposed criticality analysis demonstrates that the SFP
keff remains <=0.95 on a 95/95 basis under all normal and
postulated accident conditions, thus the required margin of
criticality safety has been maintained.
The proposed changes conservatively reduce the amount of fuel
that can be stored, and therefore do not affect the SFP cooling
analysis. Therefore, the spent fuel in the SFP remains adequately
cooled so that the fission product barriers remain intact.
The removal of Technical Specification requirements for the new
fuel storage keff limits does not affect the margin of
safety since new fuel can no longer be received.
Therefore, based on the above, 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: Lilliam M. Cuoco, Esq., Senior Nuclear
Counsel, Dominion Nuclear Connecticut, Inc., Rope Ferry Road,
Waterford, Connecticut 06385.
NRC Section Chief: Stephen Dembek.
Duke Energy Corporation, Docket Nos. 50-269, 50-270, and 50-287, Oconee
Nuclear Station, Units 1, 2, and 3, Oconee County, South Carolina.
Date of amendment request: October 16, 2001; as supplemented by
letters dated May 20, September 12, and November 21, 2002; and January
27, September 22, and November 20, 2003.
Description of amendment request: The proposed amendments would
revise the Technical Specifications to incorporate changes resulting
from the use of an alternate source term and the implementation of
several plant modifications. Publications of the Proposed No
Significant Hazards Consideration Determination and Opportunity for
Hearing have already appeared in the Federal Register on January 22,
2002 (67FR2922) and October 14, 2003 (68FR59215). The November 20,
2003, submittal contained a revised No Significant Hazards
Consideration Determination.
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:
[[Page 68661]]
Standards for determining whether a license amendment involves
no significant hazards considerations are contained in
10CFR50.92(c). The TS [Technical Specification] changes and
modifications as proposed in this LAR [license amendment request]
have been evaluated in accordance with 10 CFR 50.92 and determined
not to involve any significant hazards considerations.
The proposed LAR includes (1) implementing the AST [alternate
source term] for accident analysis as described in Regulatory Guide
1.183; (2) removing the PRVS [penetration room ventilation system]
and relaxing the SFPVS [spent fuel pool ventilation system] TS
because they are no longer credited for Control Room and off-site
doses; (3) revising the CRVS [control room ventilation system] to
allow for a one time completion time extension on Conditions B and C
when entering the conditions to support implementation of the
Control Room intake/booster fan modification; (4) lowering the
Reactor Building leakage rate from 0.25 w%/day to 0.20 w%/day; (5)
revising the VFTP [ventilation filter testing program] radioactive
methyl iodide removal acceptance criterion for SFPVS and CRVS
Booster Fan trains; and (6) adoption of TSTF [Technical
Specification Task Force]-51.
Plant modifications are also being proposed in concert with the
proposed TS changes. They include relocating the existing Control
Room outside air intake from the roof of the Auxiliary Building to
the roof of the Turbine Building and installing dual intakes for
each Control Room; re-routing HPI [high-pressure injection]/LPI
[low-pressure injection] relief valve discharge back into the
Reactor Building and replacing the existing Caustic Addition system
with a passive system.
As a result of this evaluation, Duke has concluded:
1. The proposed amendment will not involve a significant
increase in the probability of consequences of an accident
previously evaluated.
The AST and those plant systems affected by implementing the
proposed changes to the TS are not assumed to initiate design basis
accidents. The AST does not affect the design or operations of the
facility. Rather, the AST is used to evaluate the consequences of a
postulated accident. The implementation of the AST has been
evaluated in the revisions to the analysis of the design basis
accidents for ONS [Oconee Nuclear Station]. Based on the results of
these analyses, it has been demonstrated that, with the requested
changes, the dose consequences of these events meet the acceptance
criteria of 10 CFR 50.67 and Regulatory Guide 1.183. Therefore, the
proposed amendment will not involve a significant increase in the
probability or consequences of an accident previously evaluated.
2. The proposed amendment will not create the possibility of a
new or different kind of accident from any accident previously
evaluated.
The AST and those plant systems affected by implementing the
proposed changes to the TS are not assumed to initiate design basis
accidents. The systems affected by the changes are used to mitigate
the consequences of an accident that has already occurred. The
proposed TS changes and modifications do not significantly affect
the mitigative function of these systems. Consequently, these
systems do not alter the nature of events postulated in the Safety
Analysis Report nor do they introduce any unique precursor
mechanisms. Therefore, the proposed amendment will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed amendment will not involve a significant
reduction in the margin of safety.
The implementation of the AST, proposed changes to the TS and
the implementation of the proposed modifications have been evaluated
in the revisions to the analysis of the consequences of the design
basis accidents for the ONS. Based on the results of these analyses,
it has been demonstrated that with the requested changes the dose
consequences of these events meet the acceptance criteria of 10 CFR
50.67 following the provisions of Regulatory Guide 1.183. Thus, the
proposed amendment will 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: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005,
NRC Section Chief: John A. Nakoski.
Duke Energy Corporation, Docket No. 50-270, Oconee Nuclear Station,
Unit 2, Oconee County, South Carolina
Date of amendment request: October 28, 2003.
Description of amendment request: The proposed amendment would
revise the licensing basis in the Updated Final Safety Analysis Report
to support installation of a passive low-pressure injection (LPI) cross
connect inside containment. The proposed changes would revise the
licensing basis for selected portions of the core flood and LPI piping
to allow exclusion of the dynamic effects associated with postulated
pipe rupture of that piping by application of leak-before-break
methodology. A similar amendment was approved for Unit 1 by NRC letter
dated September 29, 2003.
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:
Pursuant to 10 CFR 50.91, Duke Power Company (Duke) has made the
determination that this amendment request involves a No Significant
Hazards Consideration by applying the standards established by the
NRC regulations in 10 CFR 50.92. This ensures 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:
The proposed LAR [license amendment request] modifies the Unit 2
licensing basis to allow the dynamic effects associated with
postulated pipe rupture of selected portions of the Unit 2 LPI [low-
pressure injection]/Core Flood (CF) piping to be excluded from the
design basis. The proposed design allowances for these selected
portions of piping continue to allow the LPI system design to meet
GDC [General Design Criterion] 4 requirements related to
environmental and dynamic effects. The proposed LAR will continue to
ensure that ONS [Oconee Nuclear Station] can meet design basis
requirements associated with the LPI safety function. The addition
of the crossover line will enhance the ability of the control room
operator to mitigate the consequences of specific events for which
LPI is credited. Therefore, the proposed LAR does not involve a
significant increase in the probability or consequences of an
accident previously evaluated.
(2) Create the possibility of a new or different kind of
accident from any kind of accident previously evaluated:
The proposed LAR modifies the Unit 2 licensing basis to allow
the dynamic effects associated with postulated pipe rupture of
selected portions of the Unit 2 LPI/Core Flood (CF) piping to be
excluded from the design basis. The proposed design allowances for
these selected portions of piping continue to allow the LPI system
design to meet GDC 4 requirements related to environmental and
dynamic effects. The systems affected by the changes are used to
mitigate the consequences of an accident that has already occurred.
The proposed licensing basis change does not affect the mitigating
function of these systems. Consequently, these changes do not alter
the nature of events postulated in the Safety Analysis Report nor do
they introduce any unique precursor mechanisms. Therefore, the
proposed amendment will not create the possibility of a new or
different kind of accident from any accident previously evaluated.
(3) Involve a significant reduction in the margin of safety:
The proposed licensing basis change does not unfavorably affect
any plant safety limits, set points, or design parameters. The
change also do [SIC] not unfavorably affect the fuel, fuel cladding,
RCS [reactor coolant system], or containment integrity. Therefore,
the proposed licensing basis change, which adds new design
allowances associated with the passive LPI cross connect
modification, do [SIC] not involve a significant reduction in the
margin of safety.
[[Page 68662]]
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: Anne W. Cottington, Winston and Strawn, 1200
17th Street, NW., Washington, DC 20005.
NRC Section Chief: John A. Nakoski.
Energy Northwest, Docket No. 50-397, Columbia Generating Station,
Benton County, Washington
Date of amendment request: November 4, 2003.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
technical specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), part 50, Section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4 exceptions in individual TSs would be
eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect
the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated November 4, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Thomas C. Poindexter, Esq., Winston &
Strawn, 1400 L Street, NW., Washington, DC 20005-3502.
NRC Section Chief: Stephen Dembek.
Entergy Gulf States, Inc., and Entergy Operations, Inc., Docket No. 50-
458, River Bend Station, Unit 1, West Feliciana Parish, Louisiana
Date of amendment request: October 21, 2003.
Description of amendment request: The proposed change would remove
MODE restrictions that currently prevent performance of Surveillance
Requirements (SRs) 3.8.4.7 and 3.8.4.8 for the Division III DC
electrical power subsystem while in MODE 1, 2, or 3. These
surveillances verify that the battery capacity is adequate to perform
its required functions. The changes would allow the performance of SR
3.8.4.7 and SR 3.8.4.8 during normal plant operations rather than only
during refueling outages.
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 power supplied by the battery is used as a source of control
and motive power for the HPCS [High Pressure Core Spray] system
logic, HPCS diesel-generator set control and protection, and other
Division III related controls. The loads supplied by this system are
loads associated with Division III of the Emergency Core Cooling
System (ECCS).
The battery testing period is within the period of time that the
system will already be out of service for other planned maintenance.
The battery test does not increase unavailability of the supported
system or represent any change in risk above the current practice of
planned system maintenance outages as currently allowed by the TS
[Technical Specification]. Any risk associated with the testing of
the Division III batteries will be enveloped by the risk management
of the system outage.
The out of service condition is controlled and evaluated for
safety implications in accordance with 10 CFR 50.65 [``Requirements
for monitoring the effectiveness of maintenance at nuclear power
plants'']. The HPCS system reliability and availability are
monitored and evaluated in relationship to Maintenance Rule goals to
ensure that total outage times do not degrade operational safety
over time.
[[Page 68663]]
Therefore, the proposed change will have no effect on the
probability or consequences of any previously evaluated accident.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The request involves the testing of the HPCS battery on-line
while the system is already out of service. The testing will not add
additional out of service time. Testing during this period has no
influence on, nor does it contribute in any way to, the possibility
of a new or different kind of accident or malfunction from those
previously analyzed. The method of performing this test is not
changed. No new accident modes are created by testing during the
period when the system is already unavailable. Because the system is
already out of service, no safety-related equipment or safety
functions are altered as a result of this change.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The battery testing will be performed when the HPCS system is
already out of service for maintenance. The out of service condition
is controlled and evaluated for safety implications in accordance
with 10 CFR 50.65. The batteries are not expected to be unavailable
for more than 36 hours. This testing period is within the period of
time that the system will already be out of service for other
planned maintenance. Therefore, the battery test does not increase
unavailability of the supported system or represent any change in
risk above the current practice of planned system maintenance
outages as currently allowed by the TS. Timing of this test has no
effect on any fission product barrier.
Therefore, the propose 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: Mark Wetterhahn, Esq., Winston & Strawn,
1400 L Street, NW., Washington, DC 20005.
NRC Section Chief: Robert A. Gramm.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of amendment request: October 21, 2003.
Description of amendment request: The proposed amendment would
revise Technical Specification Section 5.5.7, ``Steam Generator (SG)
Tube Surveillance Program,'' to allow a one-time extension of the
frequency for examination of the SG tubes. Specifically, the amendment
would extend the examination, currently due no later than November 17,
2004, to June 17, 2006.
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.
There is no direct increase in SG leakage because the proposed
change does not alter the plant design. The scope of the inspection
performed during the first refueling outage subsequent to SG
replacement (last outage), exceeded the technical specification
requirements for the first two refueling outages combined, after
replacement. More tubes were inspected than were required by the
technical specifications. Indian Point 2 does not have an active SG
damage mechanism and will meet the current industry examination
guidelines without performing inspections during the next refueling
outage. The results of the Condition Monitoring Assessment
subsequent to the last outage, demonstrated that all performance
criteria were met during the last operating period. The results of
the aforementioned Operational Assessment show that all performance
criteria will be met over the proposed operating period.
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 not alter any plant design basis or
postulated accident resulting from potential SG tube degradation.
The scope of the inspections performed during the last (first after
SG replacement) refueling outage significantly exceeds the Technical
Specification requirements for the scope of the first two refueling
outages combined subsequent to SG replacement.
The proposed change does not affect the SG design, the method of
operation, or reactor coolant chemistry controls. No new equipment
is being introduced, and installed equipment is not being operated
in a new or different manner. The proposed change involves a one-
time extension of the SG tube inservice inspection frequency, and
therefore will not give rise to new failure modes. In addition, the
proposed change does not impact any other plant system or
components.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
SG tube integrity is a function of design, environmental, and
current physical condition. Extending the SG tube inservice
inspection frequency by one operating cycle will not alter the
function or design of the SGs. Inspections conducted prior to
placing the SGs into service (pre-service inspection) and inspection
during the first refueling outage following SG replacement,
demonstrate that the SGs do not have fabrication damage or an active
damage mechanism. The scope of those inspections significantly
exceeds those required by the technical specifications. These
inspection results were comparable to similar inspection results for
the same model SG installed at other plants, and subsequent
inspections at those plants provided results that support the
extension request. The improved design of the replacement SGs also
provides assurance that significant tube degradation is not likely
to occur over the proposed operating period.
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 Operations, Inc., Docket No. 50-368, Arkansas Nuclear One, Unit
No. 2, Pope County, Arkansas
Date of amendment request: June 30, 2003, as supplemented by letter
dated November 20, 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
[[Page 68664]]
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), and the June 30, 2003,
application was previously noticed in the Federal Register on July 22,
2003 (68 FR 43385).
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.
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'' 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.
[[Page 68665]]
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 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 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.
(14) A change that allows the OSRC [Onsite Safety Review
Committee] review of the desirability of maintaining a channel in
the bypassed condition to be at or before the next regularly
scheduled meeting.
The proposed change 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 any 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'' 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
[[Page 68666]]
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.
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 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.
(14) A change that allows the OSRC review of the desirability of
maintaining a channel in the bypassed condition to be at or before
the next regularly scheduled meeting.
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 exiting 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
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
[[Page 68667]]
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 significant 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 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 of 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 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.
(14) A change that allows the OSRC review of the desirability of
maintaining a channel in the bypassed condition to be at or before
the next regularly scheduled meeting.
The proposed change allows the OSRC review to occur earlier than
previously required if an OSRC meeting is called before the next
regularly scheduled meeting. Therefore, the proposed 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 Operations Inc., Docket No. 50-382, Waterford Steam Electric
Station, Unit 3, (Waterford 3) St. Charles Parish, Louisiana
Date of amendment request: October 22, 2003.
Description of amendment request: The licensee proposes to change
the existing pressure/temperature limits (P/T) from 16 to 32 effective
full power years (EFPY). In addition, the maximum heatup rate will be
changed to 60 [deg]F per hour and the maximum cooldown rate to 100
[deg]F per hour for all reactor coolant system temperatures. For
inservice hydrostatic pressure and leak testing, the maximum heatup and
cooldown rates will be changed to 60 [deg]F and 100 [deg]F,
respectively.
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 analyzed?
Response: No.
The probability of occurrence of an accident previously
evaluated for Waterford 3 is not altered by the proposed amendment
to the TSs [Technical Specifications]. The accidents currently
analyzed in the Waterford 3 Final Safety Analysis Report (FSAR)
remain the same considering the results of the proposed changes to
the P/T limits and the LTOP [low temperature overpressure] enable
temperature. The new P/T and LTOP enable temperature limits were
based on the NRC [Nuclear Regulatory Commission] accepted
methodologies along with the ASME [American Society of Mechanical
Engineers] Code [Boiler and Pressure Vessel Code] alternatives. The
proposed changes do not impact the integrity of the reactor coolant
pressure boundary (RCPB) (i.e., there is no change to the
[[Page 68668]]
operating pressure, materials, loadings, etc.). The proposed change
does not affect the probability nor consequences of any design basis
accident (DBA). The proposed P/T limit curves, maximum heatup and
cooldown rates, and the LTOP enable temperature are not considered
to be an initiator or contributor to any accident currently
evaluated in the Waterford 3 FSAR. The 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. During the spring 2002 Waterford 3 refueling outage, a reactor
vessel specimen capsule was withdrawn and analyzed 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 induces ``RTNDT was calculated using the
respective reactor vessel beltline materials copper and nickel
contents and neutron fluence applicable to 32 EFPY including an
estimated increase in flux due to proposed power uprates. The RTNDT
and, in turn, the operating limits for Waterford 3 were adjusted to
account for the effects of irradiation on the fracture toughness of
the reactor vessel materials. Therefore, new operating limits will
be established which are represented in the revised operating curves
for heatup/criticality, cooldown, and inservice hydrostatic testing
contained in the TSs.
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 LTOP enable temperature will
not create a new accident scenario. The requirements to have P/T
limits and LTOP protection are part of the licensing basis for
Waterford 3. The approach used to develop the new P/T limits and
LTOP enable temperature meets NRC and ASME regulations and
guidelines. The data analysis for the vessel specimen removed during
the last Waterford 3 refueling outage confirms 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 enable temperature in the TSs
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 Waterford 3 continues to operate within the
operating 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 material reports for copper and
nickel content. The application of ASME Code Case N-641 presents
alternative procedures for calculating P/T and LTOP temperatures in
lieu of that established for ASME Section XI, Appendix G-2215. The
Code alternative allows certain assumptions to be conservatively
reduced. However, the procedures allowed by Code Case N-641 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: N.S. Reynolds, Esquire, Winston & Strawn,
1400 L Street NW., Washington, DC 20005-3502.
NRC Section Chief: Robert A. Gramm.
Exelon Generation Company, LLC, Docket Nos. 50-352 and 50-353, Limerick
Generating Station, Unit Nos. 1 and 2, Montgomery County, Pennsylvania
Date of amendment request: September 8, 2003.
Description of amendment request: The proposed change allows entry
into a mode or other specified condition in the applicability of a
Technical Specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of Title 10 of the Code of Federal
Regulations (10 CFR), Part 50, Section 50.65(a)(4). Limiting Condition
for Operation (LCO) 3.0.4, exceptions in individual TSs, would be
eliminated, and Surveillance Requirement (SR) 3.0.4 revised to reflect
the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated September 8, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions
[[Page 68669]]
of the TS. The TS allow operation of the plant without the full
complement of equipment through the conditions for not meeting the
TS LCO. The risk associated with this allowance is managed by the
imposition of required actions that must be performed within the
prescribed completion times. The net effect of being in a TS
condition on the margin of safety is not considered significant. The
proposed change does not alter the required actions or completion
times of the TS. The proposed change allows TS conditions to be
entered, and the associated required actions and completion times to
be used in new circumstances. This use is predicated upon the
licensee's performance of a risk assessment and the management of
plant risk. The change also eliminates current allowances for
utilizing required actions and completion times in similar
circumstances, without assessing and managing risk. The net change
to the margin of safety is insignificant. Therefore, this change
does not involve a significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Mr. Edward Cullen, Vice President & General
Counsel, Exelon Generation Company, LLC, 2301 Market Street,
Philadelphia, PA 19101.
NRC Section Chief: James W. Clifford.
Exelon Generation Company, LLC, and PSEG Nuclear LLC, Dockets Nos. 50-
277 and 50-278, Peach Bottom Atomic Power Station, Units 2 and 3, York
and Lancaster Counties, Pennsylvania
Date of application for amendments: September 26, 2003.
Description of amendment request: The proposed amendment would
modify the fire protection plan (FPP). The change to the FPP would
allow converting the existing carbon dioxide (CO2) fire suppression
systems, located in the cable spreading room (CSR) and each of the four
emergency diesel generator rooms, from automatic to manual actuation
systems.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed activity involves changing the actuation of the
carbon dioxide (CO2) fire suppression systems from
automatic to manual. With the exception of the Emergency Diesel
Generator (EDG) CO2 system itself, the proposed activity
does not result in any physical changes to safety-related
structures, systems, or components (SSCs), or the manner in which
safety-related SSCs are operated, maintained, modified, tested, or
inspected. The EDG CO2 system is safety related due to a
potential common mode effect on all four EDGs in the event of a
seismic event. Eliminating the automatic actuation function of the
EDG CO2 system will thereby eliminate a potential common
mode effect on the EDGs. The proposed activity does not degrade the
performance or increase the challenges of any safety-related SSCs
assumed to function in the accident analysis. As a result, the
proposed activity does not introduce any new accident initiators. In
addition, fires are not an accident that is previously evaluated.
Regardless, the proposed activity does not change the probability of
a fire occurring since fire ignition frequency is independent of the
method of fire suppression in the room. The consequences of the
proposed activity are bounded by the fire safe shutdown analysis,
which assumes fire damage throughout the affected fire area. The
fire safe shutdown analysis for each of the areas addressed by the
proposed activity demonstrates that safe shutdown can be
accomplished assuming that no fire suppression is available. In
addition, the removal of the automatic discharge capability of the
CO2 system in each of the EDG rooms significantly reduces
the potential for an inadvertent discharge to shutdown the EDG if
needed for non-fire accident conditions. Similarly, removal of the
automatic discharge feature in the CSR significantly reduces the
potential for an inadvertent discharge that would require (by
procedure) immediate shutdown of both units, and the potential
migration of CO2 into the main control room or other
areas. In the future, CO2 discharge will only occur as a
deliberate action to the most extreme fires, as one element of an
overall graded approach to fire fighting in the affected areas.
Therefore, changing the actuation of the CO2 fire
suppression systems from automatic to manual 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 activity involves changing the actuation of the
CO2 fire suppression systems from automatic to manual.
With the exception of the Emergency Diesel Generator (EDG)
CO2 system itself, the proposed activity does not result
in any physical changes to safety-related structures, systems, or
components (SSCs), or the manner in which safety-related SSCs are
operated, maintained, modified, tested, or inspected. The proposed
activity does not degrade the performance or increase the challenges
of any safety-related SSCs assumed to function in the accident
analysis. As a result, the proposed activity does not introduce nor
increase the number of failure mechanisms of a new or different type
than those previously evaluated. The fire safe shutdown analysis
assumes fire damage throughout the area consistent with a complete
lack of fire suppression capability. The elimination of the
potential for inadvertent actuation accomplished by changing the
CO2 systems from automatic to manual prevents the
CO2 systems from creating a challenge to existing
accidents.
Therefore, changing the actuation of the CO2 fire
suppression system from automatic to manual does not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed activity involves changing the actuation of the
CO2 fire suppression systems from automatic to manual.
With the exception of the Emergency Diesel Generator (EDG)
CO2 system itself, the proposed activity does not result
in any physical changes to safety-related structures, systems, or
components (SSCs), or the manner in which safety-related SSCs are
operated, maintained, modified, tested, or inspected. The proposed
activity does not degrade the performance or increase the challenges
of any safety-related SSCs assumed to function in the accident
analysis. The proposed activity does not impact plant safety since
the conclusions of the fire safe shutdown analysis remain unchanged.
Therefore, changing the actuation of the CO2 fire
suppression system from automatic to manual 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. Edward Cullen, Vice President and
General Counsel, Exelon Generation Company, LLC, 2301 Market Street,
S23-1, Philadelphia, PA 19101.
NRC Section Chief: James W. Clifford.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: August 25, 2003.
Description of amendment request: The proposed amendment would
allow extension of the current Emergency Diesel Generator (EDG)
Technical Specifications allowed outage time (AOT) from 72 hours to a
period of 14 days. This proposal would be supported by permanently
installing a non-safety-related supplemental emergency power system
(SEPS).
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), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
[[Page 68670]]
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
The proposed changes do not involve a change in the operational
limits or physical design of the electrical power systems,
particularly the emergency power systems. The proposed changes do
not change the function or operation of plant equipment or affect
the response of that equipment if called upon to operate. The
proposed AOT extensions to allow for additional operational
flexibility will not cause a significant increase in the probability
or consequences of an accident previously evaluated. In actuality,
the installation of the SEPS will have an overall net reduction in
core damage frequency. The AOT extensions will lessen the burden of
time pressure to quickly determine the cause of failure and perform
corrective actions without needing to place the plant in a transient
to shutdown because of a short allotted AOT.
A Probabilistic Risk Assessment (PRA) has been performed to
quantitatively assess the risk impact of an increase in the Allowed
Outage Time. The proposed change results in a significant decrease
in core damage frequency (CDF). Large Early Release Frequency (LERF)
is dominated by containment bypass and containment isolation
failures and remains relatively unchanged by the addition the SEPS
combined with a 14-day AOT.
Based on the above, the proposed changes will not significantly
increase the probability or consequences of an accident previously
evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed changes do not involve a change in the operational
limits or physical design of the electrical power systems,
particularly the emergency power systems. The proposed changes do
not change the function or operation of plant equipment or introduce
any new failure mechanisms. The SEPS and interfacing components with
the safety-related busses have been designed to ensure independence
and separation, particularly during faulted conditions. As such, no
new failure modes are being introduced. The plant equipment will
continue to respond per the design and analyses and there will not
be a malfunction of a new or different type introduced by the
proposed changes.
The proposed amendment extends the Allowed Outage Times for
restoring an inoperable EDG to OPERABLE status and extends the
period for operability verification of redundant features to allow
for minor repair prior to placing the plant in a shutdown transient.
The proposed amendment will not result in changes to the type of
corrective or preventive maintenance activities associated with the
EDGs. Plant operating procedures and the procedures used to respond
to abnormal or emergency conditions will be enhanced with the option
to use the SEPS when deemed necessary. Assumptions made in the
safety analysis related to EDG availability will also remain
unchanged. Performance of certain maintenance activities at power
requires an evaluation to assure plant safety is maintained or
enhanced, which would include evaluation for new or different plant
conditions. Therefore, the proposed change will not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. The proposed changes do not involve a significant reduction
in the margin of safety.
The proposed changes do not involve a change in the operational
limits. The proposed changes do not change the function or operation
of plant equipment or affect the response of that equipment if it is
called upon to operate. The performance capability of the emergency
diesel generators will not be affected. Installation of the SEPS
will have an overall net reduction in core damage frequency.
Emergency diesel generator reliability and availability will be
improved by implementation of the proposed changes. In addition,
administrative controls will ensure there are adequate compensatory
measures that can be and will be taken during extended EDG
maintenance activities to reduce overall risk. The results of the
PRA performed to quantitatively assess the risk impact of an
increase in the Allowed Outage Time indicate the proposed change
results in a significant decrease in core damage frequency (CDF) by
up to 30 percent. Therefore, the proposed changes do not involve a
significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Esquire, Florida Power & Light
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
FPL Energy Seabrook, LLC, Docket No. 50-443, Seabrook Station, Unit No.
1, Rockingham County, New Hampshire
Date of amendment request: October 6, 2003.
Description of amendment request: The proposed amendment would
revise the Seabrook Station licensing basis to implement the
alternative source term (AST) methodology of Regulatory Guide (RG)
1.183 through reanalysis of the radiological consequences of a number
of the Updated Final Safety Analysis Report Chapter 15 accidents.
Further, having revised the licensing basis, the amendment would also
revise the definition of dose equivalent I-131 in Technical
Specifications Section 1.12.
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), the licensee has provided its
analysis of the issue of no significant hazards consideration, which is
presented below:
1. The proposed changes do not involve a significant increase in
the probability or consequences of an accident previously evaluated.
Alternative source term calculations have been performed that
demonstrate the dose consequences remain below limits specified in
NRC [Nuclear Regulatory Commission] Regulatory Guide 1.183 (July
2000) and 10CFR50.67. The proposed change does not modify the
physical design or operation of the plant. The use of AST changes
only the regulatory assumptions regarding the analytical treatment
of the design basis accidents and has no direct effect on the
probability of the accident. AST has been utilized in the analysis
of the limiting design basis accidents listed above. The results of
the analyses, which include the proposed change to the Technical
Specifications, demonstrate that the dose consequences of these
limiting events are all within the regulatory limits. The proposed
Technical Specification change to the definition of dose equivalent
I-131 is consistent with the implementation of AST and the
requirements of RG 1.183 (July 2000).
Therefore, the proposed change does not involve a significant
increase the probability or consequences of an accident previously
evaluated.
2. The proposed changes do not create the possibility of a new
or different kind of accident from any previously evaluated.
The proposed change does not affect any plant structures,
systems, or components. The operation of plant systems and equipment
will not be affected by this proposed change. The alternative source
term and the dose equivalent I-131 definition change do not have the
capability to initiate accidents. 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 changes do not involve a significant reduction
in the margin of safety.
The proposed implementation of the alternative source term
methodology is consistent with NRC RG 1.183 (July 2000). The
Technical Specification change to the definition of dose equivalent
I131 is consistent with the implementation of AST and the
requirements of RG 1.183 (July 2000). Conservative methodologies,
per the guidance of RG 1.183 (July 2000), have been used in
performing the accident analyses. The radiological consequences of
these accidents are all within the regulatory acceptance criteria
associated with use of the alternative source term methodology.
The proposed changes continue to ensure that the doses at the
exclusion area and low population zone boundaries and in the Control
Room are within the corresponding regulatory limits of RG 1.183
(July 2000) and 10CFR50.67. The margin of safety for the
radiological consequences of these accidents is considered to be
that provided by meeting the applicable regulatory limits, which are
set at or below the 10CFR50.67 limits. An acceptable margin of
safety is inherent in these limits.
[[Page 68671]]
Therefore, the proposed change does not involve a significant
reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: M.S. Ross, Esquire, Florida Power & Light
Company, P.O. Box 14000, Juno Beach, FL 33408-0420.
NRC Section Chief: James W. Clifford.
Pacific Gas and Electric Company, Docket Nos. 50-275 and 50-323, Diablo
Canyon Nuclear Power Plant, Unit Nos. 1 and 2, San Luis Obispo County,
California
Date of amendment requests: September 26, 2003.
Description of amendment requests: The proposed change allows entry
into a mode or other specified condition in the applicability of a
Technical Specification (TS), while in a condition statement and the
associated required actions of the TS, provided the licensee performs a
risk assessment and manages risk consistent with the program in place
for complying with the requirements of 10 CFR 50.65(a)(4). Limiting
Condition for Operation (LCO) 3.0.4 exceptions in individual TS would
be eliminated, and Surveillance Requirement (SR) 3.0.4 revised to
reflect the LCO 3.0.4 allowance.
This change was proposed by the industry's Technical Specification
Task Force (TSTF) and is designated TSTF-359. The NRC staff issued a
notice of opportunity for comment in the Federal Register on August 2,
2002 (67 FR 50475), on possible amendments concerning TSTF-359,
including a model safety evaluation and model no significant hazards
consideration (NSHC) determination, using the consolidated line item
improvement process. The NRC staff subsequently issued a notice of
availability of the models for referencing in license amendment
applications in the Federal Register on April 4, 2003 (68 FR 16579).
The licensee affirmed the applicability of the following NSHC
determination in its application dated September 26, 2003.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), an analysis of the issue
of no significant hazards consideration is presented below:
Criterion 1--The Proposed Change Does Not Involve a Significant
Increase in the Probability or Consequences of an Accident
Previously Evaluated.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. Being in a
TS condition and the associated required actions is not an initiator
of any accident previously evaluated. Therefore, the probability of
an accident previously evaluated is not significantly increased. The
consequences of an accident while relying on required actions as
allowed by proposed LCO 3.0.4, are no different than the
consequences of an accident while entering and relying on the
required actions while starting in a condition of applicability of
the TS. Therefore, the consequences of an accident previously
evaluated are not significantly affected by this change. The
addition of a requirement to assess and manage the risk introduced
by this change will further minimize possible concerns. Therefore,
this change does not involve a significant increase in the
probability or consequences of an accident previously evaluated.
Criterion 2--The Proposed Change Does Not Create the Possibility
of a New or Different Kind of Accident from any Previously
Evaluated.
The proposed change does not involve a physical alteration of
the plant (no new or different type of equipment will be installed).
Entering into a mode or other specified condition in the
applicability of a TS, while in a TS condition statement and the
associated required actions of the TS, will not introduce new
failure modes or effects and will not, in the absence of other
unrelated failures, lead to an accident whose consequences exceed
the consequences of accidents previously evaluated. The addition of
a requirement to assess and manage the risk introduced by this
change will further minimize possible concerns. Thus, this change
does not create the possibility of a new or different kind of
accident from an accident previously evaluated.
Criterion 3--The Proposed Change Does Not Involve a Significant
Reduction in a Margin of Safety.
The proposed change allows entry into a mode or other specified
condition in the applicability of a TS, while in a TS condition
statement and the associated required actions of the TS. The TS
allow operation of the plant without the full complement of
equipment through the conditions for not meeting the TS LCO. The
risk associated with this allowance is managed by the imposition of
required actions that must be performed within the prescribed
completion times. The net effect of being in a TS condition on the
margin of safety is not considered significant. The proposed change
does not alter the required actions or completion times of the TS.
The proposed change allows TS conditions to be entered, and the
associated required actions and completion times to be used in new
circumstances. This use is predicated upon the licensee's
performance of a risk assessment and the management of plant risk.
The change also eliminates current allowances for utilizing required
actions and completion times in similar circumstances, without
assessing and managing risk. The net change to the margin of safety
is insignificant. Therefore, this change does not involve a
significant reduction in a margin of safety.
The NRC staff proposes to determine that the amendment request
involves no significant hazards consideration.
Attorney for licensee: Richard F. Locke, Esq., Pacific Gas and
Electric Company, P.O. Box 7442, San Francisco, California 94120.
NRC Section Chief: Stephen Dembek.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 23, 2003.
Description of amendment request: The proposed amendment would
delete the surveillance requirements associated with the Emergency
Diesel Generator lockout features.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed changes to the Technical Specifications (TS) 3/
4.8.1.1, AC Sources--Operating, would delete an unnecessary
surveillance. The probability of occurrence or the consequences for
an accident or malfunction of equipment is not increased by the
proposed changes. In addition, the proposed changes do not alter the
way any structure, system or component (SSC) functions, do not
modify the manner in which the plant is operated, and do not
significantly alter equipment out-of-service time. Deleting the
surveillance of equipment protection does not change the probability
or consequences of any accident and dose consequences are
unaffected. No changes to the design of structures, systems, or
components (SSC) are made and there are no effects on accident
mitigation.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The possibility of a new or different kind of accident from any
accident or malfunction in the Hope Creek Updated Final Safety
Analysis Report (UFSAR) is not created. The Emergency Diesel
Generators are accident mitigation equipment and cannot initiate an
accident. The proposed changes to the TS do not change the design
function or operation of any SSCs. The TS, as amended, would
[[Page 68672]]
continue to provide assurance of EDG operability.
Therefore, 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 proposed changes are procedural in nature and make no
changes that affect the ability of plant SSCs to perform their
design basis accident functions. In addition, the proposed changes
do not change the margin of safety since no SSCs are changed. The
results of accident analysis remain unchanged by the proposed
changes to TS.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
PSEG Nuclear LLC, Docket No. 50-354, Hope Creek Generating Station,
Salem County, New Jersey
Date of amendment request: October 24, 2003.
Description of amendment request: The proposed change to Technical
Specifications will revise surveillance requirements associated with
reactor protection system instrumentation, control rod block
instrumentation, source range monitors, and power distribution limits,
to minimize unnecessary testing.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed amendment would revise the Technical Specification
(TS) Surveillance Requirements (SRs) for certain Reactor Protection
System and Control Rod Block Instrumentation, the source range
monitors and power distribution limits, consistent with NUREG-1433,
``Standard Technical Specifications (STS) General Electric Plants,
BWR [Boiling Water Reactor]/4,'' Revision 2. No changes are being
made to any instrumentation setpoints or plant components. The
revised SRs continue to assure that the necessary quality of systems
and components is maintained, that facility operation will be within
safety limits, and that the Limiting Conditions for Operation will
be met.
Since the proposed changes do not affect any accident initiator
and since the associated equipment will remain capable of performing
its design function, the proposed change does not involve a
significant increase in the probability or radiological 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 do not change the design function or
operation of any plant equipment. No new failure mechanisms,
malfunctions, or accident initiators are being introduced by the
proposed changes. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
No changes are being made to any plant instrumentation setpoints
or to the required level of redundancy. No changes are being made to
any power distribution limits.
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: Jeffrie J. Keenan, Esquire, Nuclear Business
Unit--N21, P.O. Box 236, Hancocks Bridge, NJ 08038.
NRC Section Chief: James W. Clifford.
Virginia Electric and Power Company, Docket Nos. 50-338 and 50-339,
North Anna Power Station, Units No. 1 and No. 2, Louisa County,
Virginia
Date of amendment request: September 12, 2003.
Description of amendment request: The proposed amendments would
revise Technical Specification (TS) Sections 1.1, 3.7.10, 3.7.12,
3.7.13, 3.7.14, 3.9.4, 5.5.2, and 5.5.10, and the associated Bases
Sections to implement an alternate source term at North Anna Power
Station, Units 1 and 2. The proposed changes would implement NUREG-
1465, ``Accident Source Terms for Light-Water Nuclear Power Plants,''
dated February 1995, as the design-basis source term, achieve a
consistent design basis for all accident dose assessments, increase
operational flexibility by allowing for increased emergency core
cooling system leakage and unfiltered control room in-leakage, and
eliminate the surveillance requirement to test the bottled air flow
rate.
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:
We have reviewed the proposed TS changes relative to the
requirements of 10 CFR 50.92 and determined that a significant hazards
consideration is not involved. Specifically, operation of North Anna
Power Station with the proposed changes will not:
1. Involve a significant increase in the probability or
consequences of an accident previously evaluated.
The proposed amendment does not involve a significant increase
in the probability or consequence of an accident previously
analyzed. The North Anna MCR/ESGR [main control room/emergency
switchgear room] EVS [emergency ventilation system], PREACS [pump
room exhaust air cleanup system], and MCE [MCR]/ESGR Bottled Air
systems only function following the initiation of a design basis
radiological accident. Therefore, the changes to these
specifications, the definition of currently irradiated fuel, and the
increase [of] the depressurization time of [the] containment
following a design basis LOCA [loss-of-coolant accident] will not
increase the probability of any previously analyzed accident. These
systems are not initiators of any design bases accident.
Revised dose calculations, which take into account the changes
proposed by this [these] amendment[s] and the use of the alternative
source term[,] have been performed for the North Anna design basis
radiological accidents. The results of these revised calculations
indicate that public and control room doses will not exceed the
limits specified in 10 CFR 50.67 and Regulatory Guide 1.183. There
is not a significant increase in predicted dose consequences for any
of the analyzed accidents. Therefore, the proposed changes do not
involve a significant increase in the consequences of any previously
analyzed accident.
2. Create the possibility of a new or different kind of accident
from any accident previously evaluated.
The implementation of the proposed changes does not create the
possibility of an accident of a different type than was previously
evaluated in the UFSAR [Updated Final Safety Analysis Report].
Although the proposed changes could affect the operation of the MRC
[MCR]/ESGR EVS following a design basis radiological accident, none
of these changes can initiate a new or different kind of accident
since they are only related to system capabilities that provide
protection from accidents that have already occurred. These changes
do not alter the nature of events postulated in the UFSAR nor do
they introduce any unique precursor mechanisms.
[[Page 68673]]
Therefore, the proposed changes do not create the possibility of a
new or different kind of accident from those previously analyzed.
3. Involve a significant reduction in the margin of safety.
The implementation of the proposed changes does not reduce the
margin of safety. The proposed changes for the MCR/ESGR EVS, PREACS,
and MCE [MCR]/ESGR Bottled Air System do not affect the ability of
these systems to perform their intended safety functions to maintain
dose less than the required limits during design basis radiological
events. The revised dose calculations also indicate that the change
to the containment depressurization times will continue to maintain
the dose to the public and control room operators less than the
required limits.
The radiological analysis results, when compared with the
revised TEDE [total effective dose equivalent] acceptance criteria,
meet the applicable limits. These acceptance criteria have been
developed for application to analyses performed with alternative
source terms. These acceptance criteria have been developed for the
purpose of use in design basis accident analyses such that meeting
the stated limits demonstrates adequate protection of public health
and safety. It is thus concluded that the margin of safety will not
be reduced by the implementation of the changes.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 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 e-mail to [email protected].
AmerGen Energy Company, LLC, et al., Docket No. 50-219, Oyster Creek
Nuclear Generating Station, Ocean County, New Jersey
Date of application for amendment: June 2, 2003.
Brief description of amendment: The amendment revised the Technical
Specifications, Sections 3.7.B.1 and 3.7.C.2. Section 3.7.B.1 required
that the reactor may remain in operation ``for a period not to exceed 7
days in any 30 day period if a startup transformer is out of service.''
Section 3.7.C.2 required that the reactor may be in operation ``for a
period not to exceed 7 days in any 30 day period if a diesel generator
is out of service.'' The amendment deleted the phrase ``in any 30 day
period'' from these two sections.
Date of Issuance: November 24, 2003.
Effective date: November 24, 2003 and shall be implemented within
30 days of issuance.
Amendment No.: 239.
Facility Operating License No. DPR-16: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: July 8, 2003 (68 FR
40709).
The Commission's related evaluation of this amendment is contained
in a Safety Evaluation dated November 24, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Docket No. 50-247, Indian Point Nuclear
Generating Unit No. 2, Westchester County, New York
Date of application for amendment: March 27, 2002, as supplemented
on May 30, July 10, October 10, October 28, November 26, and December
18, 2002, and on January 6, January 27, February 26, April 8, May 19,
June 23, June 26, July 15, August 6, September 11, October 8, and
October 14, 2003.
Brief description of amendment: This amendment converts the current
Technical Specifications (TS) to a set of Improved TS based on NUREG-
1431, Revision 2, ``Standard Technical Specifications for Westinghouse
Plants,'' Revision 2, dated June 2001.
Date of issuance: November 21, 2003.
Effective date: As of the date of issuance to be implemented within
60 days.
Amendment No.: 238.
Facility Operating License No. DPR-26: Amendment replaced the
current Technical Specifications (TSs) with the Improved TSs in their
entirety and revised the license.
Date of initial notice in Federal Register: September 26, 2003 (68
FR 55660).
The supplemental letters that were received subsequent to the
issuance of the Federal Register notice provided clarifying information
that did not change the no significant hazards consideration.
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 21, 2003.
No significant hazards consideration comments received: No.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of application for amendment: March 19, 2003.
Brief description of amendment: This amendment deletes Technical
Specification (TS) 5.5.3, ``Post Accident Sampling,'' and thereby
eliminates the requirements to have and maintain the post accident
sampling system at the Pilgrim Nuclear Power Station.
[[Page 68674]]
Date of issuance: November 14, 2003.
Effective date: As of the date of issuance, and shall be
implemented within 60 days.
Amendment No.: 204.
Facility Operating License No. DPR-35: Amendment revised the TSs.
Date of initial notice in Federal Register: June 10, 2003 (68 FR
34663).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 14, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. STN 50-454 and STN 50-455,
Byron Station, Unit Nos. 1 and 2, Ogle County, Illinois
Docket Nos. STN 50-456 and STN 50-457, Braidwood Station, Unit Nos. 1
and 2, Will County, Illinois
Date of application for amendments: October 16, 2002, as
supplemented by letters dated June 20, 2003 and October 14, 2003.
Brief description of amendments: The amendments revise the
completion time of Required Action A.1 of Technical Specification
3.8.7, ``Inverters-Operating,'' from the current 24 hours to 7 days for
one inoperable instrument bus inverter. This provides greater
operational flexibility for online maintenance of an instrument bus
inverter with the potential to reduce the duration of refueling
outages.
Date of issuance: November 19, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment Nos.: 135/135, 129/129.
Facility Operating License Nos. NPF-37, NPF-66, NPF-72 and NPF-77:
The amendments revised the Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75874).
The supplemental letters contained clarifying information and did
not change the initial no significant hazards consideration
determination and did not expand the scope of the original Federal
Register notice.
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 2003. No significant hazards
consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: October 24, 2002 and as
supplemented by letter dated June 20, 2003.
Brief description of amendments: The amendments revise Technical
Specification 5.5.13, ``Primary Containment Leakage Rate Testing
Program,'' to reflect a one-time deferral of the primary containment
Type A test to no later than June 13, 2009 for Unit 1 and no later than
December 7, 2008 for Unit 2.
Date of issuance: November 19, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 162, 148.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: December 10, 2002 (67
FR 75876).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 19, 2003.
No significant hazards consideration comments received: No.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of application for amendments: May 19, 2003.
Brief description of amendments: The amendments revise Appendix A,
Technical Specifications (TS), of Facility Operating License Nos. NPF-
11 and NPF-18. Specifically, the change will decrease the frequency
associated with TS Surveillance Requirement (SR) 3.7.7.1 for Turbine
Bypass Valve (BPV) testing from 7 to 31 days. The change is consistent
with the testing frequency contained in NUREG-1434, ``Standard
Technical Specifications General Electric Plants, BWR/6,'' Revision 2,
dated June 2001, for BPV testing. The 7-day frequency associated with
SR 3.7.7.1 was established in the LaSalle County Station (LSCS) TS
during conversion to improved Standard Technical Specifications (STS)
format due to the testing frequency contained in the LSCS custom TS and
the difficulties experienced with other Electro-Hydraulic Control (EHC)
system valves to consistently pass their surveillance tests. LSCS has
recently re-evaluated the performance of these valves and has
determined that the current performance of these valves supports
decreasing the testing frequency of the BPVs from 7 to 31 days.
Date of issuance: November 13, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days.
Amendment Nos.: 163/148.
Facility Operating License Nos. NPF-11 and NPF-18: The amendments
revised the Technical Specifications.
Date of initial notice in Federal Register: June 24, 2003 (68 FR
37577).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 13, 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: December 19, 2002, as
supplemented July 25, 2003.
Brief description of amendment: The proposed amendment would revise
the Kewaunee technical specifications to change the Nuclear Regulatory
Commission reporting requirements for the discovery of defective or
degraded steam generator tubes so that the requirements are aligned
with 10 CFR 50.72 and 10 CFR 50.73.
Date of issuance: November 20, 2003.
Effective date: As of the date of issuance and shall be implemented
within 30 days.
Amendment No.: 171.
Facility Operating License No. DPR-43: Amendment revised the
Technical Specifications.
Date of initial notice in Federal Register: January 21, 2003 (68 FR
2807).
The supplemental letter 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 November 20, 2003.
No significant hazards consideration comments received: No.
Southern Nuclear Operating Company, Inc., Docket Nos. 50-348 and 50-
364, Joseph M. Farley Nuclear Plant, Units 1 and 2, Houston County,
Alabama Southern Nuclear Operating Company, Inc., et al., Docket Nos.
50-424 and 50-425, Vogtle Electric Generating Plant, Units 1 and 2,
Burke County, Georgia
Date of amendments request: September 2, 2003.
Brief Description of amendments: The amendments extend from 1 hour
to 24 hours the completion time for Condition B of Technical
Specification 3.5.1, which defines requirements for the restoration of
an emergency core cooling system accumulator when it has been declared
inoperable for a reason other than boron concentration.
[[Page 68675]]
Date of issuance: November 18, 2003.
Effective date: As of the date of issuance and shall be implemented
within 60 days from the date of issuance.
Amendment Nos.: 162, 155, 129, & 107.
Facility Operating License Nos. NPF-2 and NPF-8: Amendments revise
the Technical Specifications.
Date of initial notice in Federal Register: October 14, 2003 (68 FR
59220).
The Commission's related evaluation of the amendments is contained
in a Safety Evaluation dated November 18, 2003.
No significant hazards consideration comments received: No.
Tennessee Valley Authority, Docket No. 50-390, Watts Bar Nuclear Plant,
Unit 1, Rhea County, Tennessee
Date of application for amendment: December 19, 2002.
Brief description of amendment: The amendment consists of changes
to Technical Specification (TS) 5.0, ``Administrative Controls,'' to
incorporate three approved TS Task Force (TSTF) changes: TSTF-258,
Revision 4, ``Changes to Section 5.0, Administrative Controls''; TSTF-
299, Revision 0, ``Administrative Controls Program 5.5.2.b Test
Interval and Exception''; and TSTF-308, Revision 1, ``Determination of
Cumulative and Projected Dose Contributions in the Radioactive Effluent
Controls Program.'' In addition, two editorial changes are incorporated
to update personnel titles and clarify required staffing levels.
Date of issuance: November 13, 2003.
Effective date: As of the date of issuance and shall be implemented
within 45 days.
Amendment No.: 49.
Facility Operating License No. NPF-90: Amendment revised the TSs.
Date of initial notice in Federal Register: April 1, 2003 (68 FR
15764).
The Commission's related evaluation of the amendment is contained
in a Safety Evaluation dated November 13, 2003.
No significant hazards consideration comments received: No.
For the Nuclear Regulatory Commission
Dated at Rockville, Maryland, this 1st day of December 2003.
Ledyard B. Marsh,
Director, Division of Licensing Project Management, Office of Nuclear
Reactor Regulation.
[FR Doc. 03-30246 Filed 12-8-03; 8:45 am]
BILLING CODE 7590-01-P