[Federal Register Volume 66, Number 229 (Wednesday, November 28, 2001)]
[Notices]
[Pages 59495-59498]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-29585]
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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-333]
Entergy Nuclear Operations, Inc.; Notice of Consideration of
Issuance of Amendment to Facility Operating License and Opportunity for
a Hearing
The U.S. Nuclear Regulatory Commission (the Commission) is
considering issuance of an amendment to Facility Operating License No.
DPR-59, issued to Entergy Nuclear Operations, Inc. (ENO or the
licensee) for operation of the James A. FitzPatrick Nuclear Power
Plant, (FitzPatrick), located in Oswego County, New York.
The initial notice of consideration of issuance of amendment to
facility operating license and opportunity for hearing was originally
published in the Federal Register (64 FR 60854) on November 8, 1999,
and corrected in the Federal Register (64 FR 69574) on December 13,
1999. The information included in the supplemental letters indicates
the original notice, that included 13 proposed beyond-scope issues
(BSls) to the improved Technical Specifications (ITS) conversion, needs
to be expanded and revised to include a total of thirty one BSls and
requires re-notice in the Federal Register. This notice supercedes the
previous notice.
The proposed amendment, requested by the Power Authority of the
State of New York, the former licensee, in a letter dated March 31,
1999, as supplemented by letters dated May 20, June 1, July 14, October
14, 1999, February 11, April 4, April 13, June 30, July 31, September
12, September 13, and October 23, 2000, represents a full conversion
from the current Technical Specifications (CTS) to a set of ITS based
on NUREG-1433, ``Standard Technical Specifications (STS) for General
Electric Plants, BWR/4'' Revision 1, dated April 1995. On November 21,
2000, the Power Authority of the State of New York's (PASNY's)
ownership interest in FitzPatrick was transferred to Entergy Nuclear
FitzPatrick, LLC, to possess and use FitzPatrick and to Entergy Nuclear
Operations (ENO), Inc. to possess, use and operate FitzPatrick. By
letter dated January 26, 2001, ENO requested that the NRC continue to
review and act on all requests before the Commission which had been
submitted by PASNY before the transfer. ENO has supplemented the
original application with supplements by letter dated February 7,
February 20, May 31 and August 6, 2001. NUREG-1433 has been developed
by the Commission's staff through working groups composed of both NRC
staff members and industry representatives, and has been endorsed by
the staff as part of an industry-wide initiative to standardize and
improve the Technical Specifications (TSs) for nuclear power plants. As
part of this submittal, the licensee has applied the criteria contained
in the Commission's ``Final Policy Statement on Technical Specification
Improvements for Nuclear Power Reactors (Final Policy Statement),''
published in the Federal Register on July 22, 1993 (58 FR 39132), to
the CTS and using NUREG-1433 as a basis, proposed an ITS for
FitzPatrick. The criteria in the Final Policy Statement was
subsequently added to 10 CFR 50.36, ``Technical Specifications,'' in a
rule change that was published in the Federal Register on July 19, 1995
(60 FR 36953) and became effective on August 18, 1995.
The licensee has categorized the proposed changes to the CTS into
four general groupings. These groupings are characterized as
administrative changes, relocated changes, more restrictive changes and
less restrictive changes.
Administrative changes are those that involve restructuring,
renumbering, rewording interpretation and complex rearranging of
requirements and other
[[Page 59496]]
changes not affecting technical content or substantially revising an
operating requirement. The reformatting, renumbering and rewording
process reflects the attributes of NUREG-1433 and does not involve
technical changes to the CTS. The proposed changes include: (a)
Providing the appropriate numbers, etc., for NUREG-1433 bracketed
information (information that must be supplied on a plant-specific
basis, and which may change from plant to plant), (b) identifying
plant-specific wording for system names, etc., and (c) changing NUREG-
1433 section wording to conform to existing licensee practices. Such
changes are administrative in nature and do not impact initiators of
analyzed events or assumed mitigation of accident or transient events.
Relocated changes are those involving relocation of requirements
and surveillances for structures, systems, components, or variables
that do not meet the criteria for inclusion in TSs. Relocated changes
are those CTS requirements that do not satisfy or fall within any of
the four criteria specified in the 10 CFR 50.36(c)(2)(ii) and may be
relocated to appropriate licensee-controlled documents.
The licensee's application of the screening criteria is described
in the attachment of the licensee's March 31, 1999, submittal, which is
entitled, ``Application of NRC Selection Criteria to James A.
FitzPatrick Nuclear Power Plant Technical Specifications'' (Split
Report) in Volume 1 of the submittal. The affected structures, systems,
components or variables are not assumed to be initiators of analyzed
events and are not assumed to mitigate accident or transient events.
The requirements and surveillances for these affected structures,
systems, components, or variables will be relocated from the TSs to
administratively controlled documents such as the quality assurance
program, the final safety analysis report (FSAR), the ITS BASES, the
Technical Requirements Manual (TRM) that is incorporated by reference
in the FSAR, the Core Operating Limits Report (COLR), the Offsite Dose
Calculation Manual (ODCM), the Inservice Testing (IST) Program, or
other licensee-controlled documents. Changes made to these documents
will be made pursuant to 10 CFR 50.59 or other appropriate control
mechanisms, and may be made without prior NRC review and approval. In
addition the affected structures, systems, components, or variables are
addressed in existing surveillance procedures that are also subject to
10 CFR 50.59. These proposed changes will not impose or eliminate any
requirements.
More restrictive changes are those involving more stringent
requirements compared to the CTS for operation of the facility. These
more stringent requirements do not result in operation that will alter
assumptions relative to the mitigation of an accident or transient
event. The more restrictive requirements will not alter the operation
of process variables, structures, systems, and components described in
the safety analyses. For each requirement in the STS that is more
restrictive than the CTS that the licensee proposes to adopt in the
ITS, the licensee has provided an explanation as to why it has
concluded that adopting the more restrictive requirement is desirable
to ensure safe operation of the facility because of specific design
features of the plant.
Less restrictive changes are those where CTS requirements are
relaxed or eliminated, or new plant operational flexibility is
provided. The more significant ``less restrictive'' requirements are
justified on a case-by-case basis. When requirements have been shown to
provide little or no safety benefit, their removal from the TS may be
appropriate. In most cases, relaxations previously granted to
individual plants on a plant-specific basis were the result of (a)
generic NRC actions, (b) new NRC staff positions that have evolved from
technological advancements and operating experience, or (c) resolution
of the Owners Groups' comments on the Improved standard Technical
Specifications. Generic relaxations contained in NUREG-1433 were
reviewed by the staff and found to be acceptable because they are
consistent with current licensing practices and NRC regulations. The
licensee's design is being reviewed to determine if the specific design
basis and licensing basis are consistent with the technical basis for
the model requirements in NUREG-1433, thus providing a basis for the
ITS, or if relaxation of the requirements in the CTS is warranted based
on the justification provided by the licensee.
These administrative, relocated, more restrictive, and less
restrictive changes to the requirements of the CTS do not result in
operations that will alter assumptions relative to mitigation of an
analyzed accident or transient event.
In addition to the proposed changes solely involving the
conversion, there are also changes proposed that are different to the
requirements in both the CTS and the Standard Technical Specifications
(STS) NUREG-1433. These proposed beyond-scope issues to the its
conversion are as follows:
1. ITS 3.3.1.1, Reactor Protection System (RPS) Instrumentation
Function 5, reactor scram on main steam isolation valve (MSIV) closure.
The trip setting valve was changed from less than or equal to 10
percent (in the CTS) to less than or equal to 14 percent in the ITS.
2. ITS 3.3.1.1 changed the CTS allowable values for turbine stop
valve closure, the turbine control valve fast closure and the EHC oil
pressure low functions setpoints based on recent setpoint calculations.
3. ITS 3.3.3.1, Suppression Pool Water Temperature is modified by
footnote (c), which states : ``A channel requires 15 to 16 RTDs to be
OPERABLE.'' This results in a CTS change and a deviation from the STS.
4. ITS 3.3.4.1 changes the CTS and ISTS channel configuration from
2 channels per trip system to 4 channels in one trip system.
5. ITS 3.3.5.1 changed the CTS allowable values for CS pump flow,
LPCI pressure, LPCI pump flow, HPCI vessel water level high and HPCI
pump discharge flow low based on recent setpoint calculations.
6. ITS 3.3.5.1, Automatic Depressurization System (ADS) initiation
timer and the containment Spray (CS) and Low-Pressure Coolant Injection
(LPCI) pump start timer values were changed from the CTS and the ISTS
and tolerances relaxed to allow the extension of calibration Frequency
to 24 months in the ITS.
7. ITS 3.3.5.1 changed CTS Table 3.2-2 Item 9, Reactor Low
Pressure, LPCI and Core Spray Injection Valve Open Permissive of >450
psig to >410 psig in ITS Table 3.3.4.1-1 Functions 1.c and 2.c.
8. ITS 3.3.5.1 changed CTS Table 3.3-2, Item 5, Reactor Low Level
Containment spray Interlock trip level setting of >~0.0 inch to >~1.0
inch in ITS Table 3.3.5.1-1.
9. ITS 3.3.5.1 changed the trip setpoint Allowable Values in CTS
Table 3.2-2 for the Core Spray Pump Start Timer (item 11), the RHR LPCI
Pump Start Timer (item 12, and the Auto Blowdown Timer (item 13) in ITS
Table 3.3.5.1-1 Functions 1.d, 2.f, 4.b and 5.b to reflect values
corresponding to a 6-month to 24-month reduction in calibration
Frequency.
10. ITS 3.3.5.1 changed the trip setpoint Allowable Values in CTS
Table 3.2-1 for the suppression Chamber High Level (item 13) in ITS
Table 3.3.5.1-1 Function 3.e to 14.5 inches which is ~6 inches above
normal level.
11. ITS 3.3.5.1 changed the CTS Table 3.2-2 trip level setting for
Item 24, Reactor Low-pressure from 285 to 335
[[Page 59497]]
psig to >~300 psig in ITS Table 3.3.5.1 Function 2.d.
12. ITS 3.3.6.1 changed the Allowable Values in CTS Table 3.2-1 for
the HPCI Turbine steam Line High Flow to reflect values corresponding
to 160 to 161 inches of water differential pressure (dp) in ITS TABLE
3.3.6.1-1 Function 3.a.
13. ITS 3.3.6.1 changed the trip setpoint Allowable Value ``HPCI/
Reactor Core Isolation cooling (RCIC) Steam Line Low Pressure'' in ITS
Table 3.3.6.1-1 Function 3.b and 4.b to reflect values corresponding to
>60 and ~90 for HPCI and >61 and ~90 for RCIC.
14. ITS 3.3.6.1 changed the CTS allowable values of setpoint
temperatures for the RWCU, HPCI, and RCIC.
15. ITS 3.3.6.1 changed the CTS allowable values for the setpoints
for main steam line flow high, main steam tunnel area temperature high,
HPCI steam line flow high, HPCI turbine exhaust diaphragm pressure
high, HPCI steam line penetration (drywell entrance) area temperature
high, HPCI steam line torus room area temperature high, HPCI equipment
area temperature high, RHR heat exchanger A area temperature high,
reactor building (RB) southwest area of elevation 272 feet temperature
high, RCIC steam line flow high, RCIC steam supply line pressure low,
RCIC turbine exhaust diaphragm pressure high RCIC steam supply line
pressure low, RCIC turbine exhaust diaphragm pressure high, RCIC steam
line steam line penetration (drywell entrance) area temperature high,
RCIC steam line torus room area temperature high, RCIC equipment area
temperature high, RWCU suction line penetration area temperature high,
RWCU heat exchanger room area temperature high, RWCU pump area
temperature high (Pumps A and B), and SDC reactor pressure high to be
consistent with support setpoint calculations.
16. ITS 3.3.7.3 changed the LCO section of the Bases consistent
with the changes made to accommodate RAI 3.3.1.1-1.
17. ITS 3.3.8.1 safety analysis section of the Bases has been
changed to be consistent with changes made as a result of RAI 3.3.1-
1.SI
18. ITS 3.3.8.2 changed the Trip Level Settings for Loss of Offsite
Power (LOP) instrumentation listed in CTS Table 3.2.-2 to new ITS
Allowable Values listed in ITS Table 3.3.8.1-1.
19. ITS 3.3.8.2 changed CTS 4.9.G.3 setpoint or Allowable Value of
>~108V to >109.9V in its SR 3.3.8.2.3.
20. ITS 3.4.7 added an RHR Shutdown Cooling-Hot Shutdown
specification to the ITS SPECIFICATION based on current licensing basis
been restored to operable status within 30 days. ITS 3.3.3.1 ACTION B
specifies initiating action in accordance with ITS 5.5.6 which relates
to reporting requirements.
21. ITS 3.4.9, Reactor Coolant System (RCS) Pressure/Temperature
(P/T) Limits in CTS were changed to add a new alternate criteria in ITS
to allow idle recirculating pump (loop) start if the operating loop is
greater than 40 percent flow or if the idle loop is less than 40% flow
for less than or equal to 30 minutes.
22. ITS 3.5.1 and ITS 3.5.2, Emergency Core Cooling System (ECCS)-
Operating and Shutdown, High-Pressure Coolant Injection (HPCI) and
Residual Heat Removal (RHR) LPCI pump flow rates in CTS were reduced to
SAFER/GESTR-Loss-of-Coolant Accident (LOCA) flow rates in the ITS.
23. ITS 3.5.3 adds an additional requirement to ITS SR 3.5.3.3 that
requires the performance of the surveillance ``Once each startup prior
to exceeding 25% RTP.''
24. ITS 3.5.3 divides the existing CTS 4.5.E.1.d surveillance
requirement that ``RCIC delivers at least 400 gpm against a system head
corresponding to a reactor vessel pressure of 1195 psig to 150 psig''
into two separate Surveillance Requirements: ITS SR 3.4.3.5 and ITS SR
3.5.3.6.
25. ITS 3.6.1.1 deletes the CTS 4.7.A.1 requirement to inspect the
interior surface of the drywell and suppression chamber above the water
line every 24 months based on the inspection being required by the
primary containment leakage rate testing program 3 times in 10 years.
26. ITS SR 3.6.1.1.1 changes the note in the ISTS markup that LPCI
and Core Spray air operated testable check valve leakage test failure
does not result in an ITS SR 3.6.1.1.1 failure.
27. ITS 3.6.1.3 changed CTS LPCI and CS testable check valve
testing per Primary Containment Leakage Rate Testing (PCLRT) program
(twice every 24 months).
28. ITS SR 3.6.1.7.1, SR 3.6.1.7.2, and B 3.6.1.7 changes the
frequency of performing a functional test of each required vacuum
breaker from 31 days as indicated in the ISTS to a new schedule in
accordance with the IST Program which is 92 days.
29. ITS SR 3.6.2.3.2 was changed to add the word ``required'' to
make it clearer that the SR is only applicable to the single RHR pump
in a subsystem rather than both pumps in a subsystem that are provided
by design.
30. ITS 3.8.1, AC Sources--Operating, Condition D for two reserve
circuits inoperable in CTS was changed to add new interim power
reduction to less than or equal to 45 percent with a 36-hour Completion
Time in the ITS.
31. ITS 3.8.4, DC Sources--Operating (in CTS) was changed to allow
8 hours to restore one inoperable source in the ITS.
Before issuance of the proposed license amendments, the Commission
will have made findings required by the Atomic Energy Act of 1954, as
amended (the Act) and the commission's regulations.
By December 28, 2001, 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 Public Document Room, located at One
White Flint North, 11555 Rockville Pike (first floor), Rockville,
Maryland, or electronically on the Internet at the NRC Web site http://www.nrc.gov/NRC/CFR/index.html. If there are problems in accessing the
document, contact the Public Document Room Reference staff at 1-800-
397-4209, 301-415-4737, or by e-mail to [email protected]. If a request for a
hearing or petition for leave to intervene is filed by the above date,
the Commission or 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 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
[[Page 59498]]
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 alist 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.
A request for a hearing and petition for leave to intervene must be
filed with the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemakings and
Adjudications Staff, or may be delivered to the Commission's Public
Document Room, located at One White Flint North, 11555 Rockville Pike
(first floor), Rockville, Maryland, by the above date. A copy of the
petition should also be sent to the Office of the General Counsel, U.S.
Nuclear Regulatory Commission, Washington, DC 20555-0001, and to Mr.
David E. Blabey, attorney for the licensee, 1633 Broadway, New York,
New York 10019.
Nontimely filings of petitions for leave to intervene, amended
petitions, supplemental petitions and/or requests for hearing will not
be entertained absent a determination by the commission, the presiding
officer or the presiding Atomic Safety and Licensing Board that the
petition and/or request should be granted based upon a balancing of the
factors specified in 10CAR 2.714(a)(1)(l)-(v) and 2.714(d).
If a request for a hearing is received, the Commission's staff may
issue the amendment after it completes its technical review and prior
to the completion of any required hearing if it publishes a further
notice for public comment of its proposed finding of no significant
hazards consideration in accordance with 10 CFR 50.91 and 50.92. For
further details with respect to the proposed action, see the licensee's
application dated March 31, 1999, as supplemented by letters dated May
20, June 1, July 14, October 14, 1999, February 11, April 4, April 13,
June 30, July 31, September 12, September 13, October 23, 2000,
February 7, February 20, May 31, and August 6, 2001. Documents may be
examined, and/or copied for a fee, at the NRC's Public Document room,
located at One White Flint North, 11555 Rockville Pike (first floor),
Rockville, Maryland. Publicly available records will be accessible
electronically 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. If you do not have access to ADAMS or
if there are problems in accessing the documents located in ADAMS,
contact the NRC Public Document Room (PDR) Reference staff at 1-800-
397-4209, 301-415-4737 or by email to [email protected].
For the Nuclear Regulatory Commission.
Dated at Rockville, Maryland, this 21st day of November 2001.
Guy S. Vissing,
Project Manager, Section 1, Project Directorate I, Division of
Licensing Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-29585 Filed 11-27-01; 8:45 am]
BILLING CODE 7590-01-P