[Federal Register Volume 66, Number 157 (Tuesday, August 14, 2001)]
[Notices]
[Pages 42683-42687]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 01-20402]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-333]


Entergy Nuclear Operations; James A. Fitzpatrick Nuclear Power 
Plant Environmental Assessment and Finding of No Significant Impact

    The U.S. Nuclear Regulatory Commission (the NRC) is considering 
issuance of an amendment to Facility Operating License No. NPR-59, 
issued to Entergy Nuclear Operations (ENO or the licensee) for 
operation of the James A. FitzPatrick Nuclear Power Plant 
(FitzPatrick), located in Oswego County, New York. Therefore, as 
required by 10 CFR 51.21, the NRC is issuing this environmental 
assessment and finding of no significant impact. The original 
application was submitted by the Power Authority of the State of New 
York, (PASNY), and an Environmental Assessment (EA) and Finding of No 
Significant Impact (FONSI) was originally published in the Federal 
Register (64 FR 66509) on November 26, 1999.
    On November 21, 2000, PASNY's ownership interest in FitzPatrick was 
transferred to Entergy Nuclear FitzPatrick, LLC, to possess and use 
FitzPatrick and to ENO 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. As set forth below, PASNY and 
ENO submitted several supplements to the application. The information 
included in the supplemental letters indicates that the original 
notice, which included eleven proposed beyond-scope issues (BSIs) to 
the improved Technical Specifications (ITS) conversion, needs to be 
expanded (added 27 new BSIs) and revised to include a total of 38 BSIs. 
Accordingly, the NRC is issuing this EA and FONSI, which supercede the 
original EA and FONSI.

Environmental Assessment

Identification of the Proposed Action

    The proposed action would revise the existing, or current, 
Technical Specifications (TS) for FitzPatrick in

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their entirety based on the guidance provided in NUREG-1433, ``Standard 
Technical Specifications for General Electric Plants, BWR/4,'' Revision 
1, dated April 1995, and in the Commission's ``Final Policy Statement 
on Technical Specifications Improvements for Nuclear Power Reactors,'' 
published on July 22, 1993 (58 FR 39132). The proposed amendment is in 
accordance with the request by PASNY, 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. ENO has 
supplemented the original application by letter dated May 31, 2001.

The Need for the Proposed Action

    It has been recognized that nuclear safety in all nuclear power 
plants would benefit from the improvement and standardization of plant 
TS. The ``NRC Interim Policy Statement on Technical Specification 
Improvements for Nuclear Power Plants'' (52 FR 3788), contained 
proposed criteria for defining the scope of TS. Later, the Commission's 
``Final Policy Statement on Technical Specifications Improvements for 
Nuclear Power Reactors,'' published on July 22, 1993 (59 FR 39132), 
incorporated lessons learned since publication of the interim policy 
statement and formed the basis for revisions to 10 CFR 50.36, 
``Technical Specifications.'' The ``Final Rule'' (60 FR 36953) codified 
criteria for determining the content of TS. To facilitate the 
development of standard TS for nuclear power reactors, each power 
reactor vendor owners' group (OG) and the NRC staff developed standard 
TS. For FitzPatrick, the Improved Standard Technical Specifications 
(ISTS) are in NUREG-1433, Rev. 1. The NRC Committee to Review Generic 
Requirements (CRGR) reviewed the ISTS, made note of their safety 
merits, and indicated its support of the conversion by operating plants 
to the ISTS.

Description of the Proposed Change

    The proposed changes to the current TS (CTS) are based on NUREG-
1433, Revision 1, and on guidance provided by the Commission in the 
Final Policy Statement. The objective of the changes is to completely 
rewrite, reformat, and streamline the TS (i.e., to convert the CTS to 
Improved Technical Specifications (ITS)). Emphasis is placed on human 
factors principles to improve clarity and understanding of the TS. The 
Bases section of the ITS has been significantly expanded to clarify and 
better explain the purpose and foundation of each specification. In 
addition to NUREG-1433, Revision 1, portions of the CTS were also used 
as the basis for the development of the FitzPatrick ITS. Plant-specific 
issues (e.g., unique design features, requirements, and operating 
practices) were discussed with the licensee, and generic matters were 
discussed with General Electric and other OGs.
    The proposed changes from the ITS can be grouped into four 
categories. These groupings are characterized as administrative 
changes, relocation changes, more restrictive changes and less 
restrictive changes.
    1. Administrative changes are those that involve restructuring, 
renumbering, rewording, interpretation, and complex rearranging of 
requirements and other changes not affecting technical content or 
substantially revising an operating requirement. The reformatting, 
renumbering, and rewording process reflects the attributes of NUREG-
1433, Rev. 1, and does not involve technical changes to the ITS. 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.
    2. Relocation changes are those involving relocation of 
requirements and surveillances for structures, systems, components, or 
variables that do not meet the criteria for inclusion in TS. Relocated 
changes are those CTS requirements that do not satisfy or fall within 
any of the four criteria specified in 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 TS 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 NRC-approved control 
mechanisms, which provide appropriate procedural means to control 
changes by the licensee.
    3. 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 ISTS 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.
    4. 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 ISTS. Generic relaxations 
contained in NUREG-1433, Rev. 1 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, Rev. 1, thus providing a basis for the ITS, or if 
relaxation of the requirements in the ITS

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is warranted based on the justification provided by the licensee.
    These administrative, relocated, more restrictive, and less 
restrictive changes to the requirements of the ITS 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 differences to the 
requirements in both the CTS and the ISTS. These proposed beyond-scope 
issues to the ITS conversion are as follows:
    1. ITS 3.0.3, Limiting Condition for Operation (LCO) to be in MODE 
2 was changed to allow a 9-hour completion time.
    2. 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.
    3. ITS 3.3.1.1, Extending Required Action F.1 Completion Time from 
6 hours to 8 hours for consistency with Current Licensing Basis (CLB) 
and changing 3.0.3, which currently allows 8 hours to be in MODE 2 
after initiation of Action.
    4. 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.
    5. ITS 3.3.5.1, CS, LPCI, and ADS Logic System Functional Test 
(LSFT) frequency was extended from 18 months (in the CTS) to 24 months 
in the ITS.
    6. ITS 3.4.9, Reactor Coolant System (RCS) Pressure/Temperature 
(PT) 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 
percent flow for less than or equal to 30 minutes.
    7. ITS 3.5.1, Emergency Core Cooling System (ECCS)--Operating, 
High-Pressure Coolant Injection (HPCI) and LPCI pump flow rates in CTS 
were reduced to SAFER/GESTR-Loss-of-Coolant Accident (LOCA) flow rates 
in the ITS.
    8. ITS 3.5.2, ECCS-Shutdown, reduced Residual Heat Removal (RHR) 
LPCI pump flow rates in CTS to SAFER/GESTR-LOCA flow rates as in ITS 
3.5.1 for RHR LPCI pumps.
    9. 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.
    10. ITS 3.8.4, DC Sources--Operating (in CTS) was changed to allow 
8 hours to restore one inoperable source in the ITS.
    11. ITS 5.5, changed Standby Gas Treatment (SGT) and Control Room 
Emergency Ventilation Air Supply (CREVAS) system filter testing (in the 
CTS) from 6 months (or 12 months) to 24 months in the ITS for 
consistency with Regulatory Guide 1.52, Revision 2 or the fuel cycle 
length.
    12. ITS 3.3.5.01 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.
    13. 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.
    14. 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 
CTS 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.
    15. 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 CTS 
Table 3.3.5.1-1 Function 3.e to 14.5 inches which is ~6 inches above 
normal level.
    16. 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 psig to >~300 psig in ITS 
Table 3.3.5.1 Function 2.d.
    17. 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 dp in ITS Table 3.3.6.1-1 Function 3.a.
    18. ITS 3.3.6.1 changed the trip setpoint Allowable Value ``HPCI/
Reactor Core Isolation Cooling (RCIC) Steam Line Low Pressure'' in CTS 
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.
    19. 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.
    20. ITS 3.3.8.2 changed CTS 4.9.G.3 setpoint or Allowable Value of 
>~108V to >109.9V in its ITS SR 3.3.8.2.3.
    21. ITS 3.4.7 added an RHR Shutdown Cooling-Hot Shutdown 
specification to the ITS SPECIFICATION based on the current licensing 
basis.
    22. ITS 3.6.1.1 changed the location of the details requiring that 
the drywell and suppression chamber leakage rate limit shall be 
monitored via the suppression chamber 10 minute pressure transient of 
0.25 inches of water/minute to ITS B3.6.1.1 Bases--SR 3.6.1.1.2.
    23. ITS 3.6.1.3 modifies the ISTS criteria for the surveillance of 
Excess Flow Check valves (EFCV) to require that the EFCV be tested for 
proper operation to actuate to the isolation position on an actual or 
simulated instrument line break. This would be reflected in ITS SR 
3.6.1.3.8.
    24. ITS 3.6.1.7 modifies CTS 4.7.A.5 by addition of a new 
surveillance requirement (ITS SR 3.6.7.1). ITS SR 3.6.7.1, which is 
based on ISTS SR 3.6.1.8.1, will require verification that each 
suppression chamber-to-drywell vacuum breaker is closed every 14 days. 
The ITS SR 3.6.7.1 also deletes the ISTS SR 3.6.8.1 requirement in 
observing the vacuum breaker position by verifying that a differential 
pressure of [0.5] psid between the suppression chamber and the drywell 
is maintained for 1 hour without makeup.
    25. ITS 3.6.1.7 ACTION B changes the Completion Time to close the 
open vacuum breaker when one suppression chamber-to-drywell vacuum 
breaker is not closed to 12 hours instead of 2 hours as required by 
ISTS 3.6.1.8 ACTION B.
    26. ITS 3.6.1.9 modifies ISTS SR 3.6.1.7.1 RHR Containment Spray 
System by deleting the SR Note on system alignment in MODE 3, and adds 
the phrase ``or can be aligned to the correct position'' in ITS SR 
3.6.1.9.1. The details of the SR Note have been relocated to ITS 
B3.6.1.9 Bases--LCO.
    27. ITS 3.6.2.3 modifies ISTS B3.6.2.3--LCO by adding an insert 
that defines RHR Suppression Pool Cooling System OPERABILITY in MODE 3. 
The addition is for enhanced clarity or consistency with other Bases 
and is not in the ISTS.
    28. ITS 3.8.1 deletes the requirement that all core and containment 
cooling systems and shutdown cooling systems are OPERABLE in the CTS 
3.9.B.2 requirement that allows operation for 7 days with 2 offsite 
circuits inoperable, provided that all EDGs are OPERABLE and all core 
and containment cooling systems and shutdown cooling systems are 
OPERABLE. Instead, ITS 3.8.1 would add a requirement to declare 
required features inoperable when the redundant required features are

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inoperable, and a requirement to reduce power to less than 45 percent 
or RTP. The 7-day completion time to restore both offsite circuits to 
OPERABLE status would remain unchanged.
    29. ITS 3.3.1.1 replaces the CTS 2.1.5, ``Main Steam Line Isolation 
Valve Closure Scram'' trip setting from 10 percent closure to 14 
percent closure in proposed ITS Table 3.3.1.1-1 Function 5, ``Main 
Steam Line Isolation Valve-Closure''.
    30. ITS 3.3.3.1 changes the CTS Table 3.2-8, Note k by a footnote 
(c) in ITS Table 3.3.3.1-1, Function 10, Suppression Pool Water 
Temperature operability, which states ``A channel requires 15 of 16 
RTDs to be OPERABLE.''
    31. ITS 3.3.3.1 relaxes the CTS Table 3.2-8 Note A requirement to 
be in cold shutdown within 24 hours when one or more of Items 15 
through 18 (ECCS or Primary containment cooling operating Parameters) 
PAM channel(s) have not 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.
    32. ITS 3.3.3.1 adds additional instrument requirements to the CTS 
Table 3.2-8, which includes a Reactor Vessel Water Level Function and 
for Drywell Water Level.
    33. ITS 3.3.3.2 relocates details in CTS Table 3.2-10 relating to 
Instrument and control functions of the Remote Shutdown System 
(including number of channels and divisions), which are unnecessary in 
the LCO, to the Technical Requirements Manual (TRM).
    34. 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.
    35. ITS 3.5.1 added several ACTIONS (ACTION A, B, C, E, G, H, I, 
and J) that neither conform to the CTS nor adopt the ISTS. These are 
new actions to the Core Spray systems, the low pressure coolant 
injection systems and the high pressure coolant injection systems.
    36. ITS 3.5.3 divides the existing CTS 4.5.E.1.d SR 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.
    37. ITS 3.5.3 adds an additional requirement to CTS SR 3.5.3.3 that 
requires the performance of the surveillance ``Once each startup prior 
to exceeding 25 percent RTP.''
    38. ITS 3.3.1.1 changed low function set points on the Allowable 
Values for Reactor Pressure, High Turbine Stop Valve Closure and 
Turbine Control Valve Fast Closure, EHC Oil Pressure in CTS 2.1.A.4, 
and CTS Table 3.1-1.

Environmental Impacts of the Alternatives to the Proposed Action

    The NRC has completed its evaluation of the proposed conversion of 
the CTS to the ITS for FitzPatrick, including the beyond scope issues 
discussed above. Changes which were administrative in nature have been 
found to have no effect on the technical content of the TS. The 
increased clarity and understanding these changes bring to the TS are 
expected to improve the operators' control of FitzPatrick in normal and 
accident conditions.
    Relocation of the requirements from the ITS to other licensee-
controlled documents does not change the requirements themselves. 
Future changes to these requirements may be made by the licensee under 
10 CFR 50.59 and other NRC-approved control mechanisms, which will 
ensure continued maintenance of adequate requirements. All such 
relocations have been found consistent with the guidelines of NUREG-
1433, Rev.1, and the Commissions's Final Policy Statement.
    Changes involving more restrictive requirements have been found to 
enhance plant safety.
    Changes involving less restrictive requirements have been reviewed 
individually. When requirements have been shown to provide little or no 
safety benefit, or to place an unnecessary burden on the licensee, 
their removal from the TS was justified. In most cases, the relaxations 
previously granted to individual plants on a plant-specific basis were 
the result of generic action, or of agreements reached during 
discussions with the owners groups, and found to be acceptable for the 
plant. Generic relaxations contained in NUREG-1433, Revision 1, have 
been reviewed by the NRC staff and found to be acceptable.
    In summary, the proposed revisions to the TS were found to provide 
control of plant operations such that reasonable assurance will be 
provided that the health and safety of the public will be adequately 
protected.
    The proposed action will not significantly increase the probability 
or consequences of accidents, no changes are being made in the types of 
any effluents that may be released off site, and there is no 
significant increase in occupational or public radiation exposure. 
Therefore, there are no significant radiological environmental impacts 
associated with the proposed action.
    With regard to potential nonradiological impacts, the proposed 
action involves features located entirely within the restricted area 
for the plant defined in 10 CFR Part 20 and does not have the potential 
to affect any historic sites. It does not affect nonradiological plant 
effluents and have no other environmental impact. It does not increase 
any discharge limit for the plant. Therefore, there are no significant 
nonradiological environmental impacts associated with the proposed 
action.
    Accordingly, the NRC concludes that there are no significant 
environmental impacts associated with the proposed action.

Alternatives to the Proposed Action

    As an alternative to the proposed action, the staff considered 
denial of the proposed action (i.e., the ``no-action'' alternative). 
Denial of the application would result in no change in the current 
environmental impacts. The environmental impacts of the proposed action 
and alternative action are similar.

Alternative Use of Resources

    This action does not involve the use of any resource not previously 
considered in the FES for FitzPatrick.

Agencies and Persons Consulted

    On June 27, 2001, the staff consulted with the New York State 
official, Mr. Jack Spath, of the New York Energy and Research 
Authority, regarding the environmental impact of the proposed 
amendment. The State official had no comments.

Finding of No Significant Impact

    On the basis of the environmental assessment, the NRC concludes 
that the proposed amendment will not have a significant effect on the 
quality of the human environment. Accordingly, the NRC has determined 
not to prepare an environmental impact statement for the proposed 
action.
    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, and May 31, 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

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NRC web site, http://www.nrc.gov/NRC/ADAMS/index.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].

    Dated at Rockville, Maryland this 7th day of August 2001.

    For the Nuclear Regulatory Commission.
Richard P. Correia,
Acting Chief, Section 1, Project Directorate 1, Division of Licensing 
Project Management, Office of Nuclear Reactor Regulation.
[FR Doc. 01-20402 Filed 8-13-01; 8:45 am]
BILLING CODE 7590-01-P