Text

Draft of Safety Evaluation
	ML021540301
Person / Time
Site:	Prairie Island
Issue date:	06/03/2002
From:	; NRC/NRR/DLPM
To:	;
References
	TAC MB0695, TAC MB0696
	Download: ML021540301 (37)
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. TO FACILITY OPERATING LICENSE NO. DPR-42 AND AMENDMENT NO. TO FACILITY OPERATION LICENSE NO. DPR-60 NUCLEAR MANAGEMENT COMPANY, LLC PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 DOCKET NOS. 50-282 AND 50-306

1.0 INTRODUCTION

By application dated December 11, 2000, as supplemented by letters dated March 6, June 5, July 3, August 13, August 29, October 15, November 12, and December 12, 2001, and January 25, January 31, February 14, February 15, February 16, March 6, April 11, May 10, May 30, June xx, and June xx, 2002, Nuclear Management Company, LLC (the licensee),

requested changes to the Technical Specifications (TSs) for the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2. The proposed amendments would convert the current TS (CTS) to improved TS (ITS). The Commissions proposed action for the ITS conversion was published in the Federal Register on June xx, 2002 ( FR ).

PINGP has been operating with TS issued with the original Facility Operating Licenses on April 5, 1974 (for Unit 1), and October 29, 1974 (for Unit 2), as amended. The proposed conversion to the ITS is based upon:

NUREG-1431, "Standard Technical Specifications for Westinghouse Plants," Revision 1, dated April 1995;

The current PINGP CTS;

"Final Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (Final Policy Statement), published on July 22, 1993 (58 FR 39132); and

10 CFR 50.36, Technical Specifications, as amended July 19, 1995 (60 FR 36953).

Hereinafter, the proposed improved TS for PINGP are referred to as the ITS, the current TS are referred to as the CTS, and the improved standard TS, such as in NUREG-1431, are referred to as the STS. The corresponding Bases are ITS Bases, CTS Bases, and STS Bases, respectively. For convenience, a list of acronyms used in this Safety Evaluation (SE) is provided in Attachment 1 to this SE.

In addition to basing the ITS on the STS, the Final Policy Statement, and the requirements in 10 CFR 50.36, the licensee retained portions of the CTS as a basis for the ITS. Several post-application letters of request for additional information (RAI) and a series of telephone conference calls as well as meetings were required during the course of this review. These RAIs as well as the meetings and conference calls served to clarify the ITS with respect to the guidance in the Final Policy Statement and the STS. In addition, based on these discussions, the licensee also proposed matters of a generic nature that were not in the STS. The staff requested that the licensee submit such generic issues as proposed changes to the STS through the NRC/Nuclear Energy Institutes Technical Specifications Task Force (TSTF).

These generic issues were considered for specific applications in the PINGP ITS. Consistent with the Final Policy Statement, the licensee proposed transferring some CTS requirements to licensee-controlled documents (such as the PINGP Updated Safety Analysis Report (USAR),

for which changes to the documents by the licensee are controlled by a regulation such as 10 CFR 50.59 and may be changed without prior NRC approval). NRC-controlled documents, such as the TS, may not be changed by the licensee without prior NRC approval. In addition, human factors principles were emphasized to add clarity to the CTS requirements being retained in the ITS, and to define more clearly the appropriate scope of the ITS. Further, significant changes were proposed to the CTS Bases to make each ITS requirement clearer and easier to understand.

The overall objective of the proposed amendments, consistent with the Final Policy Statement, is to rewrite, reformat, and streamline the TS for PINGP, while still satisfying the requirements of 10 CFR 50.36. During its review, the staff relied on the Final Policy Statement and the STS as guidance for acceptance of CTS changes. This SE provides a summary basis for the staffs conclusion that the licensee can develop ITS based on STS, as modified by plant-specific changes, and that the use of the ITS is acceptable for continued operation of PINGP. This SE also explains the staffs conclusion that the ITS, which are based on the STS as modified by plant-specific changes, are consistent with the PINGP current licensing basis and the requirements of 10 CFR 50.36.

The license conditions included in the conversion amendment will make enforceable the following aspects of the conversion: (1) the relocation of requirements from the CTS and (2) the implementation schedule for new and revised surveillance requirements (SRs) in the ITS.

The staff also acknowledges that, as indicated in the Final Policy Statement, the conversion to ITS is a voluntary process. Therefore, it is acceptable that the ITS differ from the STS to reflect the current licensing basis for PINGP. The staff approves the licensee's changes to the CTS with modifications documented in the licensees supplemental submittals.

For the reasons stated infra in this SE, the staff finds that the ITS issued with these license amendments comply with Section 182a of the Atomic Energy Act, 10 CFR 50.36, and the guidance in the Final Policy Statement, and that they are in accord with the common defense and security and provide adequate protection of the health and safety of the public.

2.0 BACKGROUND

Section 182a of the Atomic Energy Act requires that applicants for nuclear power plant operating licenses will state:

[S]uch technical specifications, including information of the amount, kind, and source of special nuclear material required, the place of the use, the specific characteristics of the facility, and such other information as the Commission may, by rule or regulation, deem necessary in order to enable it to find that the utilization . . . of special nuclear material will be in accord with the common defense and security and will provide adequate protection to the health and safety of the public. Such technical specifications shall be a part of any license issued.

In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TS. In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and the mitigation of accident consequences. As recorded in the Statements of Consideration, "Technical Specifications for Facility Licenses; Safety Analysis Reports" (33 FR 18610, December 17, 1968), the Commission noted that applicants were expected to incorporate into their TS "those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity." Pursuant to 10 CFR 50.36, TS are required to include items in the following five specific categories related to station operation: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) SRs; (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plants TS.

For several years, NRC and industry representatives have sought to develop guidelines for improving the content and quality of nuclear power plant TS. On February 6, 1987, the Commission issued an interim policy statement on TS improvements, "Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors" (52 FR 3788). During the period from 1989 to 1992, utility owners groups and the staff developed improved STS, such as NUREG-1431, that would establish models of the Commission's policy for each primary reactor type. In addition, the staff, licensees, and owners groups developed generic administrative and editorial guidelines in the form of a "Writers Guide" for preparing TS, which gives greater consideration to human factors principles and was used throughout the development of licensee-specific ITS.

In September 1992, the Commission issued NUREG-1431, Revision 0, which was developed using the guidance and criteria contained in the Commission's Interim Policy Statement. The STS in NUREG-1431 was established as a model for developing the ITS for Westinghouse plants, in general. The STS reflect the results of a detailed review of the application of the interim policy statement criteria to generic system functions, which were published in a "Split Report" issued to the nuclear steam supply system (NSSS) vendor owners groups in May 1988. STS also reflect the results of extensive discussions concerning various drafts of STS, so that the application of the TS criteria and the Writer's Guide would consistently reflect

detailed system configurations and operating characteristics for all reactor designs. As such, the generic Bases presented in NUREG-1431 provide an abundance of information regarding the extent to which the STS present requirements that are necessary to protect public health and safety. The STS in NUREG-1431 apply to PINGP .

On July 22, 1993, the Commission issued its Final Policy Statement, expressing the view that satisfying the guidance in the policy statement also satisfies Section 182a of the Act and 10 CFR 50.36. The Final Policy Statement described the safety benefits of the STS and encouraged licensees to use the STS as the basis for plant-specific TS amendments and for complete conversions to ITS based on the STS. In addition, the Final Policy Statement gave guidance for evaluating the required scope of the TS and defined the guidance criteria to be used in determining which of the LCOs and associated SRs should remain in the TS. The Commission noted that, in allowing certain items to be relocated to licensee-controlled documents while requiring that other items be retained in the TS, it was adopting the qualitative standard enunciated by the Atomic Safety and Licensing Appeal Board in Portland General Electric Co. (Trojan Nuclear Plant), ASLAB-531, 9 NRC 263, 273 (1979). There, the Appeal Board observed:

[T]here is neither a statutory nor a regulatory requirement that every operational detail set forth in an applicants safety analysis report (or equivalent) be subject to a technical specification, to be included in the license as an absolute condition of operation which is legally binding upon the licensee unless and until changed with specific Commission approval. Rather, as best we can discern it, the contemplation of both the Act and the regulations is that technical specifications are to be reserved for those matters as to which the imposition of rigid conditions or limitations upon reactor operation is deemed necessary to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to the public health and safety.

By this approach, existing LCO requirements that fall within or satisfy any of the criteria in the Final Policy Statement should be retained in the TS; those LCO requirements that do not fall within or satisfy these criteria may be relocated to licensee-controlled documents. The Commission codified the four criteria in 10 CFR 50.36 (60 FR 36953, July 19, 1995). The four criteria are as follows:

Criterion 1 Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2 A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 3 A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

Criterion 4 A structure, system, or component which operating experience or probabilistic safety assessment has shown to be significant to public health and safety.

Part 3.0 of this SE explains the staffs conclusion that the conversion of the PINGP CTS to ITS based on STS, as modified by plant-specific changes, is consistent with the PINGP current licensing basis and the requirements and guidance of the Final Policy Statement and 10 CFR 50.36.

3.0 EVALUATION In its review of the PINGP ITS application, the staff evaluated five kinds of changes to the CTS as defined by the licensee. The staff s review also included an evaluation of whether existing regulatory requirements are adequate for controlling future changes to requirements that are removed from the CTS and placed in licensee-controlled documents. Following are the five types of CTS changes:

A Administrative - Changes to the CTS that do not result in new requirements or change operational restrictions and flexibility.

M More Restrictive - Changes to the CTS that result in added restrictions or reduced flexibility.

L Less Restrictive - Changes to the CTS that result in reduced restrictions or added flexibility.

LR Less Restrictive Relocated Details - Changes to the CTS that eliminate detail and relocate the detail to a licensee-controlled document. Typically, this involves details of system design, system description including design limits, description of system or plant operation, procedural detail for meeting TS requirements and relocated reporting requirements, and redundant requirement references.

R Relocated Specifications - Changes to the CTS that relocate the requirements that do not meet the selection criteria of 10 CFR 50.36(c)(2)(ii).

The ITS application included a justification for each proposed change to the CTS in a numbered discussion of change (DOC), using the above letter designations as appropriate. In addition, the ITS application included an explanation of each difference between ITS and STS requirements in a numbered justification for deviation (JFD).

In its review, the staff identified the need for clarifications and additions to the December 11, 2000, ITS application in order to establish an appropriate regulatory basis for translation of CTS requirements into ITS. The staffs comments were documented as RAIs and forwarded in letters dated June 1, June 15, August 17, December 6, December 19, December 21, 2001, and

January 8, 2002. The licensee provided responses to the RAIs in supplemental letters dated March 6, July 3, August 13, November 12, and December 12, 2001, and January 25, January 31, February 14, February 15, February 16, March 6, April 11, May 10, May 30, June xx, and June xx, 2002. The letters clarified the licensees basis for translating the CTS requirements into ITS. For items that have been reviewed by the staff as stated in this SE, the staff finds that the licensee's submittals, including the responses to the RAIs, provide sufficient detail to allow the staff to reach a conclusion regarding the adequacy of the licensee's proposed changes to the CTS.

The changes to the CTS, as presented in the ITS application, are listed and described in the following five tables attached to this SE:

Table A - Administrative Changes

Table M - More Restrictive Changes

Table L - Less Restrictive Changes

Table LR- Less Restrictive Relocated Details

Table R - Relocated Specifications These tables provide a summary description of the proposed changes to the CTS, references to the specific CTS requirements that are being changed, and the specific ITS requirements that incorporate the changes. The tables are only meant to summarize the changes being made to the CTS. The details as to what the actual changes are and how they are being made to the CTS or ITS are provided in the licensees application and supplemental letters.

The staffs evaluation and additional description of the kinds of changes to the CTS requirements listed in Tables A, M, L, LR, and R are presented in Sections A through E below, as follows:

Section A Administrative Changes

Section B More Restrictive Changes

Section C Less Restrictive Changes

Section D Less Restrictive Relocated Details

Section E Relocated Specifications The control of specifications, requirements, and information relocated from the CTS is described in Section F below, and other CTS changes (i.e., beyond-scope changes) are described in Section G below.

A. Administrative Changes to the CTS Administrative (non-technical) changes are intended to incorporate human factors principles into the form and structure of the ITS so that plant operations personnel can use them more easily. These changes are editorial in nature or involve the reorganization or reformatting of CTS requirements without affecting technical content or operational restrictions. Every section of the ITS reflects this type of change. In order to ensure consistency, the staff and the licensee have used the STS as guidance to reformat and make other administrative changes.

Among the changes proposed by the licensee and found acceptable by the staff are:

Identifying plant-specific wording for system names, etc.;

Splitting up requirements currently grouped under a single current specification and moving them to more appropriate locations in two or more specifications of ITS;

Combining related requirements currently presented in separate specifications of the CTS into a single specification of ITS;

Presentation changes that involve rewording or reformatting for clarity (including moving an existing requirement to another location within the TS) but that do not involve a change in requirements;

Wording changes and additions that are consistent with CTS interpretation and practice, and that more clearly or explicitly state existing requirements;

Deletion of TS which no longer apply;

Deletion of details that are strictly informational and have no regulatory basis; and

Deletion of redundant TS requirements that exist elsewhere in the TS.

Table A lists the administrative changes being made in the PINGP ITS conversion. Table A is organized in STS order by each A-type DOC to the CTS, provides a summary description of the administrative change that was made, and provides CTS and ITS references. The staff reviewed all of the administrative and editorial changes proposed by the licensee and finds them acceptable because they are compatible with the Writers Guide and the STS, do not result in any change in operating requirements, and are consistent with the Commission's regulations.

B. More Restrictive Changes to the CTS The licensee, in electing to implement the specifications of the STS, proposed a number of requirements more restrictive than those in the CTS. The ITS requirements in this category include requirements that are either new, more conservative than corresponding requirements in the CTS, or have additional restrictions that are not in the CTS but are in the STS. Examples of more restrictive requirements are placing an LCO on plant equipment that is not required by the CTS to be operable, more restrictive requirements to restore inoperable equipment, and more restrictive SRs. Table M lists the more restrictive changes being made in the PINGP ITS conversion. Table M is organized in STS order by each M-type DOC to the CTS and provides a summary description of the more restrictive change that was adopted, and the CTS and ITS references. These changes are additional restrictions on plant operation that enhance safety and are acceptable.

C. Less Restrictive Changes to the CTS Less restrictive requirements include deletions and relaxations to portions of the CTS requirements that are being retained in the ITS. When requirements have been shown to give little or no safety benefit, their relaxation or 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 (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the Owners Groups comments on STS. The NRC staff reviewed generic relaxations contained in the STS and found them acceptable because they are consistent with current licensing practices and the Commission's regulations. The PINGP design was also reviewed to determine if the specific design basis and licensing basis are consistent with the technical basis for the model requirements in the STS and thus provide a basis for ITS.

All of the less-restrictive changes to the CTS have been evaluated and found to involve deletions and relaxations to portions of CTS requirements that can be grouped in the following nine types:

Type 1 Relaxation of LCO Requirements Type 2 Relaxation of Applicability Type 3 Relaxation of Surveillance Requirement Type 4 Relaxation of Required Action Type 5 Relaxation of Completion Time Type 6 Deletion of Requirements Redundant to Regulation or Design Information Type 7 Relaxation of Surveillance Frequency from 18 months to 24 months Type 8 Relaxation of 3.0.C Requirements Type 9 Relaxation of SR Frequency The following discussions address why these types of less-restrictive changes are acceptable.

Type 1 Relaxation of the LCO Requirements Certain CTS LCOs contain operational and system parameters beyond those necessary to meet safety analysis assumptions and therefore are considered overly restrictive. CTS also contain limits which have been shown to give little or no safety benefit to the safe operation of the plant. The ITS, consistent with the guidance in the STS, delete or revise operating limits in this type. CTS LCO changes included in this type are: (1) redefines operating modes including Mode title changes; (2) deleting or revising operational limits to establish requirements consistent with applicable safety analyses; (3) deleting equipment or systems which establish redundant system capability beyond that assumed to function by the applicable safety analyses or which are implicit to the ITS requirement for systems, components and devices to be operable; and (4) adding allowances to use administrative controls on plant devices and equipments during times when automatic control is required or to establish temporary administrative limits, as appropriate, to allow time for systems to establish equilibrium operation. TS changes represented by this type allow operators to more clearly focus on issues important to safety. The resultant ITS LCOs maintain an adequate degree of protection consistent with the safety analysis. They also improve focus on issues important to safety and provide reasonable operational flexibility without adversely affecting the safe operation of the plant. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 2 Relaxation of Applicability The CTS require compliance with the LCO during the Operational Mode(s) or other conditions specified in the LCO Applicability statement. Six Operating Modes are defined by TS according to average reactor coolant temperature, the position of the reactor mode switch located in the control room, and reactor vessel head closure bolt tensioning; Power Operation, Startup, Hot Standby, Hot Shutdown, Cold Shutdown and Refueling. When CTS Applicability requirements are inconsistent with the applicable accident analyses assumptions for a system, subsystem or component specified in the LCO, the LCO is changed in the ITS to establish a consistent set of requirements. These modifications or deletions are acceptable because, during the conditions referenced in the ITS, the operability requirements are consistent with the applicable safety analyses. These changes

are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 3 Relaxation of Surveillance Requirement CTS require maintaining the LCO equipment operable by meeting the SRs in accordance with the specified SR Frequency. This requires conducting tests to demonstrate equipment is operable, or that LCO parameters are within specified limits. When the test acceptance criteria and any specified conditions for the conduct of the test are met, the equipment is deemed operable. The changes in this type relate to relaxation of CTS SR acceptance criteria and/or the conditions for performing the SR.

Relaxing the SR acceptance criteria for these items provides operational flexibility consistent with the objective of the STS without reducing confidence that the equipment is operable. For example, the ITS also permits the use of an actual, as well as a simulated, actuation signal to satisfy SRs for automatically actuated systems. TS required features cannot distinguish between an actual signal and a test signal. The changes to TS acceptance criteria are acceptable because appropriate testing standards are retained for determining that the LCO-required features are operable.

Relaxing conditions for performing SRs include, for example, not requiring testing of de-energized equipment (e.g., instrumentation channel checks) or equipment that is already performing its intended safety function (e.g., position verification of valves locked in their safety actuation position). The changes also include the allowance to verify the position of valves in high radiation areas by administrative means. ITS administrative controls (ITS 5.7) regarding access to high radiation areas make the likelihood of mispositioning valves small. These changes are acceptable because the changes do not affect the ability to determine whether equipment is capable of performing its intended safety function.

These relaxations of CTS SRs optimize test requirements for the affected safety systems and increase operational flexibility. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 4 Relaxation of Required Action LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, CTS specify actions to be taken until the equipment is restored to its required capability or performance level, or remedial measures are established. Compared to CTS required actions, certain proposed ITS actions result in extending the time period for taking the plant outside the applicability into shutdown conditions. For example, changes in this type include providing an option to:

isolate a system, place equipment in the state assumed by the safety analysis, satisfy alternate criteria, take manual actions in place of automatic actions, restore to operable status within a specified time frame, place alternate equipment into service, or use more conservative TS setpoints. The resulting ITS actions continue to provide measures that conservatively compensate for the inoperable equipment. The ITS actions are commensurate with safety importance of the inoperable equipment, plant design and industry practice and do not compromise safe operation of the plant. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 5 Relaxation of Completion Time Upon discovery of a failure to meet an LCO, TS specify times for completing Required Actions of the associated TS conditions. Required Actions establish remedial measures that must be taken within specified completion times (allowed outage times). These times define limits during which operation in a degraded condition is permitted. Incorporating completion time extensions is acceptable because completion times take into account the operability status of the redundant systems of TS required features, the capacity and capability of remaining features, a reasonable time for repairs or replacement of required features, vendor-developed standard repair times, and the low probability of a design basis accident (DBA) occurring during the repair period. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 6 - Deletion of Requirements Redundant to Regulation or Design Information CTS contain requirements that are redundant to regulations in 10 CFR. CTS reporting requirements 10 CFR 50.73. The CTS include requirements that Reportable Event is any of those conditions specified in 10 CFR 50.73. However, the ITS, consistent with the STS, omit many of the CTS reporting requirements because the reporting requirements in the regulations cited do not need repeating in the TS to ensure timely submission to the NRC.

Therefore, this type of change has no impact on the safe operation of the plant. Deletion of these requirements is beneficial because it reduces the administrative burden on the plant which in turn allows increased attention to plant operations that are important to safety.

These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

CTS contain design information that is deleted in ITS. In conformance with the guidance of STS ITS fuel assemble design features are simplified. The ITS includes allowances for changes to core designs if certain conditions are met. The allowances require that limited substitutions of filler rods for fuel rods are permitted if fuel assemblies comply with fuel safety design bases and a limited number of untested lead test assemblies may be installed in non-limiting core regions. All fuel assemblies with substitutions will be evaluated in accordance with NRC approved codes and methods or will be tested to show they comply with all fuel safety design bases. Lead test assemblies will not challenge any reactor operating limits since, by the requirements of this specification the substitutions are restricted to non-limiting core regions. Additionally, core performance is monitored throughout the operating cycle to assure that it performs safely. Thus the reactor will remain safe when operated in accordance with the provisions of ITS 4.2.1 which implements guidance provided in Generic Letter 90-02, Alternative Requirements for Fuel Assemblies in the Design Features Section of Technical Specifications, Supplement 1. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 7 Relaxation of Surveillance Frequency from 18 months to 24 months CTS require maintaining the LCO equipment operable by conducting SRs in accordance with the specified SR interval. The changes in this type relate to extending SR test intervals. Improved reactor fuels allow the licensee to consider an increase in the duration

of the fuel cycle for their facility. TS that specify an 18-month surveillance interval or refueling interval or during shutdown are changed to specify a 24-month interval. The CTS 4.0.A (ITS SR 3.0.2) provision to extend surveillances by 25 percent of the specified interval would extend the time limit for completing these surveillances from the CTS limit of 22.5 months to a maximum of 24 months. The staff review of these items is covered in more detail in Section G of this SE. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 8 Relaxation of CTS 3.0.C Requirements CTS 3.0.C (ITS LCO 3.0.3) establishes actions that must be implemented when an LCO is not met and either an associated Required Action and Completion Time is not met and no other Condition applies, or the condition of the unit is not specifically addressed by the associated TS Actions. This Specification delineates the time limits for placing the unit in a safe Mode or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its Actions. It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable. Unless otherwise stated LCO 3.0.3 always applicable to ITS LCO Actions. However, new Required Actions are included within the Actions of ITS LCOs that provide guidance for taking the plant to a specified condition or applicable mode in which the LCO does not apply without requiring entry into LCO 3.0.3 which would require a shutdown to Mode 5. These new remedial actions place the plant in a safe condition in a controlled manner, thus reducing the likelihood that additional structures, systems or components will be unavailable to mitigate operational occurrences or plant transients and therefore these proposed changes do not impact safe operation of the plant. These changes are consistent with precedence establish by STS in consideration of PINGP current licensing basis and are acceptable.

Type 9 Relaxation of SR Frequency Prior to placing the plant in a specified operational mode or other condition stated in the applicability of an LCO, and in accordance with the specified SR time interval thereafter, the CTS require establishing the operability of each LCO-required component by meeting the SRs associated with the LCO. This usually entails performance of testing to demonstrate the operability of the LCO-required components, or the verification that specified parameters are within LCO limits. A successful demonstration of operability requires meeting the specified acceptance criteria as well as any specified conditions for the conduct of the test. Relaxations of CTS SRs include relaxing both the acceptance criteria and the conditions of performance. Also, the ITS permits the use of an actual, as well as a simulated, actuation signal to satisfy SRs for automatically actuated systems. This is acceptable because TS-required features cannot distinguish between an actual signal and a test signal. These relaxations of CTS SRs optimize test requirements for the affected safety systems and increase operational flexibility. These CTS SR relaxations are consistent with precedence establish by STS in consideration of PINGP current licensing basis.

For the reasons presented above, these less-restrictive changes to CTS are acceptable because they will not affect the safe operation of the plant. The ITS requirements are consistent with current licensing basis, operating experience, and plant accident and transient analyses, and provide reasonable assurance that public health and safety will be protected.

Table L lists the less restrictive changes being made in the PINGP ITS conversion. Table L, which is organized in STS order by each L-type DOC to the CTS, provides a summary description of the less restrictive change that was made, the CTS and ITS references, and a reference to the specific change type discussed above. The staff reviewed all of the less restrictive changes proposed by the licensee and finds them acceptable because they are compatible with the STS, do not result in any change in operating requirements, and are consistent with the Commissions regulations.

D. Less Restrictive Relocated Details When requirements have been shown to give 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 (1) generic NRC actions, (2) new staff positions that have evolved from technological advancements and operating experience, or (3) resolution of the owners groups comments on STS. The staff reviewed generic relaxations contained in the STS and found them acceptable because they are consistent with current licensing practices and the Commission's regulations. The PINGP design was also reviewed to determine if the specific design basis and licensing basis are consistent with the technical basis for the model requirements in the STS and thus provide a basis for ITS. A significant number of changes to the CTS involved the removal of specific requirements and detailed information from individual specifications evaluated to be Types 1 through 3 as described below:

Type 1 - Removing Details of System Design and System Description, Including Design Limits The design of the facility is required to be described in the USAR by 10 CFR 50.34. In addition, the quality assurance (QA) requirements of Appendix B to 10 CFR Part 50 require that plant design be documented in controlled procedures and drawings and maintained in accordance with an NRC-approved QA plan (USAR Chapter 17). 10 CFR 50.59 specifies controls for changing the facility as described in the USAR. 10 CFR 50.54(a) specifies criteria for changing the QA plan. The Technical Requirements Manual (TRM) is a general reference in the USAR and is subject to administrative controls that include the requirement to perform 10 CFR 50.59 evaluations for changes made to the TRM. The ITS Bases also contain descriptions of system design. ITS 5.5.12 specifies controls for changing the Bases. Removing details of system design from the CTS is acceptable because this information will be adequately controlled in the USAR in accordance with 10 CFR 50.59 or the ITS Bases, as appropriate. Cycle-specific design limits are contained in the Core Operating Limits Report (COLR) in accordance with Generic Letter 88-16. ITS Section 5.6, Administrative Controls, includes the programmatic requirements for the COLR. Therefore, it is acceptable to remove Type 1 details from the CTS and place them in licensee-controlled documents.

Type 2 - Removing Descriptions of System or Plant Operation The plans for the normal and emergency operation of the facility are required to be described in the USAR by 10 CFR 50.34. Specification 5.4.1.a and e. require written procedures to be established, implemented, and maintained for plant operating procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978, and in all programs specified in Specification 5.5, respectfully. The ITS Bases also contain descriptions of system operation. Controls specified in 10 CFR 50.59 apply to changes in procedures as described in the USAR and ITS Bases. ITS 5.5.12 specifies controls for changing the Bases. It is acceptable to remove details of system operation from the TS because this type of information will be adequately controlled in the USAR (which references the TRM), in the TS Bases and in Specification 5.5, Programs and Manuals, as appropriate. Therefore, it is acceptable to remove Type 2 details from the CTS and place them in licensee-controlled documents.

Type 3 - Removing Procedural Details for Meeting TS Requirements Details for performing TS SRs are more appropriately specified in the plant procedures.

Prescriptive procedural information in an ITS requirement is unlikely to contain all procedural considerations necessary for the plant operators to comply with TS, and referral to plant procedures is therefore required in any event. Changes to procedural details include those associated with limits retained in the ITS. For example, Specification 5.4.1 requires written procedures to be established, implemented, and maintained covering activities that include all programs specified in Specification 5.5.

The Inservice Testing (IST) program is required by Specification 5.5.7. ITS 5.5.7, Inservice Testing Program requires a program to provide controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program includes defining testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda. CTS also contains requirements to test specific components such as pumps valves which establish inservice testing of Quality Group A, B and C pumps and valves performed in accordance with the requirements for American Society of Mechanical Engineers (ASME) Code Class 1, 2 and 3 components specified in Section XI of the applicable ASME Boiler and Pressure Vessel Code Edition and Addenda, subject to the applicable provisions of 10 CFR 50.55a. Therefore, it is acceptable to remove Type 3 details from the CTS and place them in licensee-controlled documents.

Type 4 - Relocated Redundant Requirements Certain CTS administrative requirements are redundant to regulations and thus are relocated to the USAR or other appropriate licensee-controlled documents. The Final Policy Statement allows licensees to relocate to licensee-controlled documents CTS requirements that do not meet any of the criteria for mandatory inclusion in the TS.

Changes to the facility or to procedures as described in the USAR are made in accordance with 10 CFR 50.59. Changes made in accordance with the provisions of other licensee-controlled documents are subject to the specific requirements of those documents. For example, 10 CFR 50.54(a) governs changes to the QA plan, and ITS 5.5.13 governs changes to the ITS Bases. Therefore, it is acceptable to remove Type 4 details from CTS and place them in licensee-controlled documents.

Table LR lists the less restrictive removal of detail changes being made in the PINGP ITS conversion. Table LR is organized in STS order by each LR type DOC. It includes the following: (1) the DOC identifier (e.g., 3.4 followed by LR-74 means STS Chapter 3.4, DOC LR-74; (2) a summary description of the relocated details and requirements; (3) the name of the licensee-controlled document to contain the relocated details and requirements (location); (4) the regulation (or ITS Specification) for controlling future changes to relocated requirements (change control process); (5) the reference numbers of the associated CTS requirements; and (6) a characterization of the type of change.

The staff has concluded that these types of detailed information and specific requirements do not need to be included in the ITS to ensure the effectiveness of the ITS to adequately protect the health and safety of the public. Accordingly, these requirements may be moved to one of the following licensee-controlled documents for which changes are adequately governed by a regulatory or TS requirement:

Bases controlled in accordance with ITS 5.5.13, Technical Specifications (TS)

Bases Control Program.

USAR (which references TRM) controlled by 10 CFR 50.59.

Programmatic documents required by ITS Section 5.5 and controlled by ITS Section 5.4.

Inservice Inspection (ISI) and IST Programs controlled by 10 CFR 50.55a.

ODCM controlled by ITS 5.5.1.

COLR controlled by ITS 5.6.4.

QA Plan, as approved by the NRC and referenced in the USAR, controlled by 10 CFR Part 50, Appendix B, and 10 CFR 50.54(a).

Site Emergency Plan controlled by 10 CFR 50.54(q).

To the extent that information has been relocated to licensee-controlled documents, such information is not required to obviate the possibility of an abnormal situation or event giving rise to an immediate threat to public health and safety. Further, where such information is contained in LCOs and associated requirements in the CTS, the staff has concluded that they do not fall within any of the four criteria set forth in 10 CFR 50.36(c)(2)(ii) and discussed in the Final Policy Statement (see Section 2.0 of this SE). Accordingly, existing detailed information, such as generally described above, may be removed from the CTS and not included in the ITS.

E. Relocated Specifications The Final Policy Statement states that LCOs and associated requirements that do not satisfy or fall within any of the four specified criteria (now contained in 10 CFR 50.36(c)(2)(ii)) may be relocated from existing TS (an NRC-controlled document) to appropriate licensee-controlled documents. This section discusses the relocation of entire specifications in the CTS to licensee-controlled documents. These specifications include the LCOs, Action Statements (i.e., Actions), and associated SRs. In its application and its supplements, the licensee proposed relocating such specifications from the CTS to the TRM, and the ODCM, as appropriate. The NRC staff has reviewed the licensee's submittals and finds that relocation of these requirements to the TRM, and the ODCM, is acceptable in that the LCOs and associated requirements were found not to fall within the scope of 10 CFR 50.36(c)(2)(ii) and changes to the TRM, and the ODCM, will be adequately controlled by 10 CFR 50.59, 10 CFR 50.54(a),

10 CFR 50.55a, and ITS 5.5.1, as applicable. These provisions will continue to be implemented

by appropriate station procedures (i.e., operating procedures, maintenance procedures, surveillance and testing procedures, and work control procedures).

Table R lists the relocated changes being made in the PINGP ITS conversion and lists all specifications that are being relocated from the CTS to licensee-controlled documents. Table R includes: (1) references to the DOCs; (2) references to the relocated CTS specifications; (3) summary descriptions of the relocated CTS specifications; (4) names of the documents that will contain the relocated specifications (i.e., the new location); and (5) the methods for controlling future changes to the relocated specifications (i.e., the regulatory control process).

The staff s evaluation of each relocated specification listed in Table R is provided below, mostly in CTS order. New locations for relocated CTS are listed in Table R.

3.11 Core Surveillance Instrumentation The core surveillance instrumentation are relocated to the TRM. CTS requires that the moveable detector and core thermocouple instrumentation systems be operable. The moveable detector system is required to be operable following each refueling so that the power distribution can be confirmed. In addition, sufficient detectors, drives, and readout equipment to map these thimbles is required to be operable during recalibration of the excore axial offset detection system per technical specifications. Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for this instrumentation is not assumed in the safety analysis nor the PI site-specific PRA since it is a non-significant risk contributor. The moveable detector and core thermocouple instrumentation systems do not meet the requirements for TS and have been relocated to the TRM.

2.3.C, Control Rod Withdrawal Stops, Limiting Safety System Settings, Protective Instrumentation The Control Rod Withdrawal Stops, P-2 interlock is relocated to the TRM. This instrumentation is provided to prevent movement of rods using automatic rod controls.

When the plant power level is less than 15% RTP the turbine impulse pressure (PT-485) input to the rod control system is blocked, preventing automatic rod withdrawal. This interlock enables automatic rod control when the power reaches 15%. Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for the P-2 interlocks is not assumed in the safety analysis. These automatic stops do not meet the requirements for TS and have been relocated to the TRM.

Table 4.1-1C, Function 13, Containment Sump A and C Level The containment sump level instrumentation are relocated to the TRM. This discussion of containment sump level does not include consideration of the containment sump B wide range indication which is included in ITS LCO 3.3.3. The containment sump level

instruments are installed instrumentation that are capable of indicating in the control room a significant abnormal degradation of the reactor coolant pressure boundary.

However, this instrumentation is less sensitive to reactor coolant pressure boundary leakage than other instrumentation which is required by ITS LCO 3.4.16 and are not included in the safety analysis discussion of reactor coolant system leakage detection systems. Credit for the containment sump level instrumentation is not assumed in the safety analysis nor the PI IPE. Since containment sump level instruments are not required in ITS, the SRs on this instrumentation can be relocated to the TRM Table 4.1-1C, Function 16, Emergency Plan Radiation Instruments The emergency plan radiation instrumentation are relocated to the TRM. The emergency plan radiation instruments are used to gather environmental information following an accident which requires entry into the emergency plan. Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for the emergency plan radiation instruments is not assumed in the safety analysis nor the PI IPE. Since emergency plan radiation instruments are not required in ITS, the SRs on this instrumentation can be relocated to the TRM.

Table 4.1-1C, Function 17, Seismic Monitors The seismic monitors (instrumentation) are relocated to the TRM. The seismic monitors are used to record data for use in evaluating the effect of a seismic event.

Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for the seismic monitors is not assumed in the safety analysis nor the PI IPE. Since seismic instruments are not required in ITS, the SRs on this instrumentation can be relocated to the TRM.

Table 4.1-1C, Function 19, CRDM Cooling Shroud Exhaust Air Temperature The CRDM cooling shroud exhaust air temperature instrumentation are relocated to the TRM. The CRDM cooling shroud exhaust air temperature instrumentation is installed instrumentation for indicating the cooling air temperature above the reactor pressure vessel. Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for the CRDM cooling shroud exhaust air temperature instrumentation is not assumed in the safety analysis nor the PI IPE. Since CRDM cooling shroud exhaust air temperature instruments are not required in ITS, the SRs on this instrumentation can be relocated to the TRM.

Table 4.1-1C, Function 20, Reactor Gap Exhaust Air Temperature The Reactor Gap Exhaust Air Temperature instrumentation are relocated to the TRM.

The Reactor Gap Exhaust Air Temperature instrumentation is installed instrumentation for indicating the cooling air temperature around the reactor pressure vessel.

Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission

product barrier. Credit for the Reactor Gap Exhaust Air Temperature instrumentation is not assumed in the safety analysis nor the PI IPE. Since reactor gap exhaust air temperature instruments are not required in ITS, the SRs on this instrumentation can be relocated to the TRM.

Table 4.1-1C, Function 31, Turbine Overspeed Protection Trip Channel The turbine overspeed protection trip instrumentation is relocated to the TRM. The turbine overspeed protection trip instrumentation provides a means to detect a turbine overspeed condition and trip the turbine. Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for the turbine overspeed protection trip instrumentation is not assumed in the safety analysis nor the PI IPE. Since turbine overspeed protection trip channels are not required in ITS, the SRs on this instrumentation can be relocated to the TRM.

3.4.C, Table 3.5-1 Function 8, 4.8.C and Table 4.1-1C Function 24, The Steam Exclusion System The Steam Exclusion System (SES) actuation instrumentation and the associated setpoint have been relocated to the TRM. The SES is an installed system which monitors Auxiliary Building and Turbine Building ventilation duct temperatures and, upon a high temperature condition due to a high energy line break, isolates the ducts and prevents steam from reaching safeguards equipment. The SES is a plant system.

Credit for the SES actuation instrumentation and the associated setpoint is not assumed in the safety analysis nor the PI IPE. The SES actuation instrumentation and the associated setpoint do not meet the requirements for TS and have been relocated to theTRM.

3.1.A.3 and 4.18 Reactor Vessel Head Vent System The reactor vessel head vent system is relocated to the TRM. CTS requires that the reactor shall not be made or maintained critical nor shall reactor coolant system average temperature exceed 200 degrees Fahrenheit unless the reactor head vent system paths from both the reactor vessel head and pressurizer steam space are OPERABLE and closed. These vents are designed to exhaust noncondensable gases and steam from the RCS which could inhibit natural circulation following an accident with an extended loss of offsite power. Credit for this vent system is not assumed in the safety analyses nor in the PI IPE. The requirements of the reactor vessel head vent system LCO do not meet the requirements for TS and have been relocated to the TRM.

3.1.B.2 Pressurizer Heatup and Cooldown Limits The pressurizer heatup and cooldown limits are relocated to the PTLR. The shutdown requirements associated with the pressurizer heatup and cooldown limitations are relocated to the TRM. CTS requires specific limits on the maximum heatup and maximum cooldown in any one hour period. The pressurizer heatup and cooldown limits are not process variables, design features, or operating restrictions that are initial conditions of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. PI does not take any credit for the pressurizer heatup and cooldown limits in any accident analysis. Therefore, pressurizer heatup and cooldown limits are not operational limits that are an initial assumption of any DBA or transient analysis.

3.1.B.3 Steam Generator Pressure/Temperature Limits The steam generator pressure/temperature limits are relocated to the PTLR. The shutdown requirements associated with the steam generator pressure/temperature limits are relocated to the TRM. CTS requires that the secondary side of the steam generator not be pressurized above 200 psig if the temperature of the steam generator is below 70 degrees Fahrenheit. This operating restriction does not present a challenge to the integrity of a fission product barrier and these limits are not required for safe operation of the facility. PI does not take any credit for the steam generator pressure/temperature limits in any accident analysis. Therefore, steam generator limits are not operational limits that are an initial assumption of any DBA or transient analysis.

3.8.C, Small Spent Fuel Pool Restrictions The Small Spent Fuel Pool Restrictions are relocated to the TRM. The requirement limits the number of recently discharged fuel assemblies stored in the small pool (Pool 1). Spent fuel storage limits retained in TS are relied upon as a process variable that is an initial condition of a design-basis accident or tranient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for limits on the number of recently discharged fuel asssemblies stored in the small pool are not assumed in the safety analysis. Limits on the number of recently discharged fuel asssemblies stored in the small pool do not meet the requirements for TS and have been relocated to the TRM.

CTS 3.12, 4.13 and Table 4.13-1, Snubber Requirements The snubber requirements are relocated to the TRM. Snubbers are passive devices that are designed to prevent unrestrained pipe motion under dynamic loads and allow normal thermal expansion of piping and nozzles to eliminate excessive thermal stresses during heatup or cooldown. The TS action statement for snubbers only requires that an inoperable snubber be replaced or repaired. The surveillance requirements for snubbers is that they be periodically examined under the inservice inspection program.

Credit for snubber operability requirements are not assumed in the safety analysis.

Limits for operable snubbers do not meet TS requirements and have been relocated to the TRM.

Table 4.1-2A, Function 11, Turbine Stop Valves, Governor Valves and Intercept Valves Turbine stop valves, governor valves and intercept valve requirements are relocated to the TRM. Periodic testing of the turbine stop valves, governor valves and intercept valves are performed to reduce the probability of a turbine missle ejection incident.

These surveillance requirements are part of the Turbine Overspeed Protection System.

The turbine overspeed protection trip instrumentation provides a means to detect a turbine overspeed condition and trip the turbine. Instrumentation systems retained in TS are relied upon to form a part of the primary success path which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Credit for the turbine overspeed protection trip instrumentation is not assumed in the safety analysis nor the PI IPE. Since turbine overspeed protection trip channels are not required in ITS, the SRs on turbine stop valves, governor valves and intercept valves can be relocated to the TRM.

4.19, Auxiliary Building Crane Lifting Devices Auxiliary Building Crane Lifting Devices are relocated to the TRM. The testing required for the equipment in this specification verifies that special lifting devices and slings used in conjunction with the auxiliary building crane are operable prior to their use in supporting heavy loads over safe shutdown equipment or spent fuel in the spent fuel pool. Credit for the Auxiliary Building Crane Lifting Devices is not assumed in the safety analysis. The Auxiliary Building Crane Lifting Devices do not meet the requirements for TS and have been relocated to the TRM.

The relocated specifications from the CTS discussed above are not required to be in the TS because they do not fall within the criteria for mandatory inclusion in the TS as stated in 10 CFR 50.36(c)(2)(ii). These specifications are not needed to obviate the possibility that an abnormal situation or event will give rise to an immediate threat to the public health and safety.

In addition, the staff has concluded that appropriate controls have been established for all of the current specifications and information that are being moved to the TRM, ODCM, ISI, or IST Programs. These relocations are the subject of a new license condition discussed in Section 5.0 of this SE. Until incorporated in licensee-controlled documents, changes to these specifications and information will be controlled in accordance with the current applicable procedures and regulations that control these documents. Following implementation, the NRC may audit the removed provisions to ensure that an appropriate level of control has been achieved. The staff has concluded that, in accordance with the Final Policy Statement, sufficient regulatory controls exist under the regulations, particularly 10 CFR 50.59 and 10 CFR 50.55a. Accordingly, the specifications and information, as described in detail in this SE, may be relocated from the CTS and placed in the licensee-controlled documents identified in the licensees submittals.

F. Control of Specifications, Requirements, and Information Relocated from the CTS In the ITS conversion, the licensee will be relocating specifications, requirements, and detailed information from the CTS to the licensee-controlled documents outside the CTS. This is discussed in Sections 3.0.D and 3.0.E above. The facility and procedures described in the USAR and TRM can only be revised in accordance with the provisions of 10 CFR 50.59, which ensure records are maintained and establish appropriate control over requirements removed from the CTS and over future changes to the requirements. Other licensee-controlled documents contain provisions for making changes consistent with applicable regulatory requirements. For example, the ODCM can be changed in accordance with ITS 5.5.1, and the administrative instructions that implement the QA Plan can be changed in accordance with 10 CFR 50.54(a) and 10 CFR Part 50, Appendix B. The documentation of these changes will be maintained by the licensee in accordance with the record retention requirements specified in the QA Plan and such applicable regulations as 10 CFR 50.59.

The license condition for the relocation of requirements from the CTS, which is discussed in Section 5.0 of this SE, will address the implementation of the ITS conversion and the schedule for the relocation of the CTS requirements into licensee-controlled documents.

G. Evaluation of Other TS Changes (Beyond-Scope Changes) Included in the Application for Conversion to ITS This section evaluates other TS changes included in the licensees conversion application.

These include items that deviate from both the CTS and the STS, do not fall clearly into a category, or are in addition to those changes that are needed to meet the overall purpose of the conversion. These changes are termed beyond-scope issues (BSIs), which have been identified by the licensee in their submittal, and by the staff during the course of the staff review.

These BSIs were included in the Notice of Consideration of Issuance of Amendments to Facility Operating Licenses and Opportunity for a Hearing published in the Federal Register on June xx, 2002 ( FR ).

G.1 BSI Changes Identified by the Licensee:

The changes discussed below are licensee-identified BSIs and are listed in the order of the applicable ITS specification or section, as appropriate. Also provided are references to the associated DOC to the CTS and JFD from the STS given in the licensees application.

G.1.1 Extension of Surveillance Intervals (From 18 month to 24 month)

As part of this conversion process, the licensee has proposed to clarify the requirements for those surveillances which are intended to be performed during plant shutdown conditions or require operating cycle interval.

The current fuel cycle at Prairie Island cannot exceed 24 months and the CTS does not include a provision for any extension of this surveillance interval. The CTS 4.0.A.2 states that The intervals between tests scheduled for refueling shutdowns shall not exceed two years. The proposed ITS SR 3.0.2 retains this CTS requirement by specifying in part: ...The specified Frequency is met for each SR with a specified Frequency of 24 months if the Surveillance is performed within 24 months, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met...

The licensee has categorized each ITS SRs with 24 month Frequencies into three groups depending on their origin and CTS Frequency requirements. The first group of SRs (Table 1) have CTS SR Frequency statements as: each refueling shutdown, R (Each Refueling shutdown), once each refueling interval, each reactor refueling shutdown, each refueling outage, or each refueling shutdown on STB [staggered test basis], indicating operating cycle intervals. Prairie Island could operate on 24 month cycles and still meet the CTS required test interval for these SRs. Thus, no change to CTS is required for this group of SRs.

The second group of SRs (Table 2) have CTS SR Frequency statements as: once per 18 months; once each 18 months; every 18 months; once per operating cycle, or once each 18 months whichever occurs first, or similar words indicating 18 month intervals. (The provision to extent surveillances by 25 percent of the specified interval would extend the time limit for completing these surveillances to a maximum of 22.5 months.)

The third group of SRs (Table 3) are those that are not included in the CTS.

GROUP 1 SRs ITS SR TYPE OF SR ITS SR TYPE OF SR 3.3.1.10 Channel Calibration 3.5.2.5 Verify ECCS valve position 3.3.1.11 Channel Calibration 3.5.2.6 Verify ECCS pump starts 3.3.1.12 Channel Calibration 3.6.3.7 Verify Containment valve position 3.3.1.13 Channel Operational Test 3.6.5.3 Verify Containment cooling flow 3.3.1.14 TADOT 3.6.5.5 Verify CS valve position 3.3.1.16 RTS Response Time Test 3.6.5.6 Verify CS pump starts

3.3.2.4 TADOT 3.6.5.7 Verify Containment cooling starts 3.3.2.5 TADOT 3.6.6.4 Verify CS additive valve position 3.3.2.6 Channel Calibration 3.6.7.1 SFT for Hydrogen Recombiner 3.3.3.2 Channel Calibration 3.6.7.2 VT for Hydrogen Recombiner 3.3.3.3 TADOT 3.6.7.3 ET for Hydrogen Recombiner 3.3.4.3 Channel Calibration 3.6.8.2 Channel Calibration 3.3.5.4 TADOT 3.6.9.4 Verify SBVS damper actuations 3.3.5.5 Channel Calibration 3.7.7.2 Verify CC valve actuations 3.4.1.3 Verify RCS flow rate (COLR limit) 3.7.7.3 Verify CC pump starts 3.4.12.5 Channel Calibration 3.7.8.5 Verify CL valve actuations 3.4.13.6 Channel Calibration 3.7.8.6 Verify DDCLP and MDCLP starts 3.4.15.1 Verify PIV leakage GROUP2 SRs ITS SR TYPE OF SR ITS SR TYPE OF SR 3.4.9.2 Verify Pressurizer Heater power source 3.7.12.4 Verify ABSVS train actuates 3.4.11.2 Cycling of PORVs 3.7.13.3 Verify SFPSVS train actuates 3.5.2.7 Verify ECCS throttle valve position 3.7.13.4 Verify SFPSVS fan flow rate 3.6.9.3 Verify SBVS train actuation 3.8.1.7 EDG load reject test 3.7.5.3 Verify AFW valve actuation/position 3.8.1.8 Verify EDG trip bypass on SI signal 3.7.5.4 Verify AFW pump starts 3.8.1.9 EDG 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months run test 3.7.10.3 Verify CRSVS train actuation 3.8.1.10 LOOP/SI actuation test 3.7.10.4 Verify CRSVS fan flow rate GROUP3 SRs ITS SR TYPE OF SR ITS SR TYPE OF SR 3.3.1.13 Channel Operational Test 3.7.2.2 Verify MSIV actuation 3.3.2.7 Master-slave relay test 3.7.3.2 Verify MFRV and bypass valve actuation 3.3.5.6 Master-slave relay test 3.7.4.2 Master-slave relay test 3.3.6.3 Master-slave relay test 3.7.11.1 Verify SCWS loop actuation 3.3.6.4 Master-slave relay test 3.8.1.11 Master-slave relay test

3.3.7.3 Master-slave relay test 3.8.4.2 Battery charger load test 3.4.16.3 Channel Calibration 3.8.4.3 Verify Battery charger capacity 3.4.16.4 Channel Calibration 3.9.3.2 Channel Calibration 3.5.2.8 Verify ECCS train sump suction 3.9.4.2 Verify actuation (Containment purge valve) 3.6.2.2 Verify air lock doors The staff has reviewed and approved a number of requests to extend surveillance requirements to accommodate a 24-month fuel cycle. (The provision to extend surveillances by 25 percent of the specified interval would extend the time limit for completing these surveillances to a maximum of 30 months).

As described in the preceding section, the Prairie Islands request is somewhat different than the previous cases because (1) Prairie Islands CTS allows surveillance intervals up to 24 months; and (2) the proposed ITS retains this specification without the provision to extend surveillance intervals by 25 percent (i.e., up to 30 months). Therefore, no change to CTS is required for a majority of the 24-month interval SRs in the proposed ITS. Specifically, the Group 1SRs (Table 1) which have CTS SR Frequency statements as: each refueling shutdown, R (Each Refueling shutdown), once each refueling interval, each reactor refueling shutdown, each refueling outage, or each refueling shutdown on STB [staggered test basis], requires no change to CTS. Prairie Island could operate on 24 month cycles and still meet the CTS required test interval for these SRs.

There are a number of similar SRs, however, due to inconsistencies in the CTS, have CTS SR Frequency statements as: once per 18 months; once each 18 months; every 18 months; once per operating cycle, or once each 18 months whichever occurs first, or similar words indicating 18 month intervals (Table 2 - Group 2 SRs). The provision to extent surveillances by 25 percent of the specified interval would extend the time limit for completing these surveillances to a maximum of 22.5 months, versus 24 month maximum interval for the ITS.

For this reason, the staff has requested additional justification from the licensee based on Generic Letter 91-04, Changes in Technical Specification Surveillance Intervals to Accommodate a 24-Month Fuel Cycle, issued on April 2, 1991.

Generic Letter 91-04 provides staff guidance that identifies the types of information that must be addressed when proposing extensions of the fuel cycle to 24 months. The GL identified the following information to support conversion to a 24-month operating cycle:

1) Licensees should evaluate the effect on safety of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the effect on safety is small.

2) Licensees should confirm that historical plant maintenance and surveillance data support this conclusion.

3) Licensees should confirm that assumptions in the plant licensing basis would not be invalidated on the basis of performing any surveillance at the bounding surveillance interval limit provided to accommodate a 24-month fuel cycle.

(4) In consideration of these confirmations, the licensees need not quantify the effect of the change in surveillance intervals on the availability of individual systems or components.

In accordance with the above guidance, the licensee evaluated the effect on safety of an increase in 18-month surveillance intervals to 24-month surveillance intervals, and concluded that the effect on plant safety is small. The licensee has also concluded that the assumptions in the plant licensing basis would not be invalidated based on the performance of any surveillance at the bounding surveillance interval limits (not to exceed 24 months). In addition, the licensee has reviewed historical plant maintenance and surveillance data. As a result of this review, the licensee identified a total of three failures in three different, unrelated systems. With the exception of the three identified failures, all other surveillance tests were passed. Those systems for which the surveillance tests passed maintained a 100% reliability. The three systems which has a failed surveillance test maintained a 84% reliability. These systems passed their surveillance test when minor repairs were completed and the failed surveillance was re-run.

Based on this information, the staff concludes that the impact of the proposed change for the Group 2 SRs on plant safety is small and, therefore, the proposed change is acceptable.

Group 3 SRs (Table 3) represent new requirements in the ITS which the licensee has adopted as part of the TS conversion process. Since these are new SRs that are not required in the CTS, these are considered more restrictive changes to the CTS, and therefore, acceptable.

G.1.2 EXTENSION OF AOT FOR ECCS ACCUMULATORS As part of this conversion process, the licensee has proposed to revise TS 3.3, Engineered Safety Features, to extend the allowed outage time of an inoperable accumulator from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This proposal is based on the methodology described in Topical Report WCAP-15049-A, Risk-Informed Evaluation of an Extension to Accumulator Completion Times.

The purpose of the emergency core cooling system (ECCS) accumulators are to supply water to the reactor vessel during the blow-down phase of a loss-of-coolant accident (LOCA). The accumulators are large volume tanks, filled with borated water and pressurized with nitrogen.

The cover-gas pressure is less than that of the reactor coolant system (RCS) so that when the RCS pressure decreases below the tank pressure, the accumulators inject borated water into the RCS cold legs.

The CTS has an allowed outage time (AOT) for one accumulator of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . However, the Westinghouse Owners Group (WOG) determined that a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AOT is insufficient for responding to accumulator inoperability. Therefore, the WOG submitted Topical Report WCAP-15049-A, which generically evaluated the risk associated with extending accumulator allowed outage times from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for reasons other than boron concentration out of specification at Westinghouse plants. In the report, the WOG did not request an extension to the accumulator boron out of specification TS, which for most plants had been set to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

They determined that the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months were an adequate amount of time to correct the boron related problems. The NRC reviewed and approved WCAP-15409-A for referencing in licensing applications in its Safety Evaluation, Acceptance for Referencing of Westinghouse Owners Group Topical Report WCAP-15049-A, Risk-Informed Evaluation of an Extension to Accumulator Completion Times, dated February 19, 1999 (Ref. 3).

CTS 3.3.A.2.e allows for one of the two accumulators to be inoperable for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during startup operation or power operation when the pressurizer pressure is greater than 1000 psig. With one accumulator inoperable, the remaining accumulator will be available to mitigate the consequences of a LOCA. The licensee is requesting to increase the AOT for one accumulator from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Prairie Island Updated Safety Analysis Report (USAR) safety analyses include the assumption that both accumulators are operable in case of design basis accidents. These accidents do not consider the case when the accumulators are in their AOT.

The licensee is not changing the design or the operating characteristics of the accumulators.

Therefore, the current safety analysis that evaluates the operation of the accumulators remains unchanged by the extension of the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AOT to a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months AOT for one accumulator.

However, the change in the AOT will affect the overall risk at the plant.

One portion of this TS AOT extension request is not covered by WCAP-15049-A, because this extension also applies to the condition where an accumulator is inoperable due to the accumulator boron concentration being out of specification. As mentioned in the Introduction Section above, the WOG did not propose extending the boron out of specification AOT for an inoperable accumulator. The WOG did not request this extension because many Westinghouse plants have this AOT already set to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , which is adequate to correct the accumulator boron out of specification issues. PINGP, on the other hand, had only one TS AOT for all accumulator inoperability conditions. An increase to the AOT of the boron out of specification condition of the TS is acceptable because the boron concentration in the accumulators is considered only during the recirculation phase of the LOCA, and the impact of a single accumulators borated water volume is not significant when compared to the total borated water volume present during the recirculation phase. Therefore, applying the AOT extension that the NRC found acceptable in WCAP-15049-A to all cases of an inoperable accumulator (including boron out of specification) for Prairie Island would not invalidate the current safety analysis, and we find the proposed change acceptable.

With its current PRA model, the licensee estimates the average internal events core damage frequency (CDF) for the PINGP to be 2.2E-05/yr. According to Westinghouse Owners Group (WOG) Topical Report WCAP-15049-A, which was prepared to support requests for increases in accumulator completion time (CT) to 24 hr, the change in the average CDF for a two loop plant with an increase in accumulator CT to 24 hr is approximately 3E-07/yr. Based on these levels of risk, RG 1.174 indicates the proposed change can be considered on a risk-informed basis. The NRC staff has not reviewed the licensees PRA nor the details of its numerical analysis, and has not performed an independent analysis of the proposed change. However, the licensees CDF appears reasonable, it is not too different from the CDF reported in the PINGP Individual Plant Examination (IPE) (5E-05/yr) considering improvements to the PRA model that the licensee listed in the amendment request, and the staff has no reason to think the current PRA model is not adequate to support this request. With regard to WCAP-15049-A, as mentioned earlier, it was reviewed and the risk analysis was found to be acceptable by the NRC staff for use, by reference, in licensing applications extending accumulator CTs to 24 hr provided the plant design and operating characteristics are consistent with the limitations stated in the WCAP and associated NRC safety evaluation. This review does not repeat review of matters described in the WOG WCAP except to ensure that PINGP is within the stated limitations of the report and associated safety evaluation.

The licensee discussed the limitations of WCAP-15049-A, expressed in terms of assumptions and parameters used in the WCAP, with respect to its applicability to PINGP design and operating characteristics. PINGP is a two loop plant and the WCAP includes analysis of a two loop plant. PINGP does not perform any accumulator test activities nor preventative maintenance activities at power and its accumulator corrective maintenance frequency is 0.1/yr (from 1995 - 2001), which are assumptions incorporated in the WCAP. As in the Prairie Island PRA, the initiating events for which accumulators are modeled in the WCAP are large, medium, and small LOCAs; for all LOCA categories, the frequencies used in the WCAP bound those used in Prairie Island PRA (which are probably more realistic, being based on data from NUREG/CR-5750). The success criteria assumed in the WCAP analysis are the same as those used in the current plant specific PRA model. For these reasons, WCAP-15049-A is considered applicable to PINGP.

Therefore, the staff has concluded that the proposed changes to TS 3.3.A.2.e for extending the AOT for the Prairie Island ECCS Accumulators from 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is acceptable.

Additionally, the applicability of this TS was changed from whenever the pressurizer pressure is greater than 1000 psig to whenever the reactor coolant system pressure is greater than 1000 psig. Since the RCS system pressure and the pressurizer pressure are identical, this change is editorial in nature and has no effect on plant safety. Therefore, the staff finds this change acceptable.

G.1.3 Missed surveillance Consolidated Line Item Improvement As part of the ITS package, the licensee proposed to adopt TSTF-358, "Missed Surevillance Consolidated Line Item Improvement," to extend the delay period for a missed SR from the current limit of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to "... up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater." In addition, the following requirement would be added to ITS SR 3.0.3:

"A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed." TSTF-358 is one of the industrys initiatives under the Risk-informed Technical Specifications Program that has been approved by the staff as documented in the staffs safety evaluation published on Federal Register dated June 14, 2001 (66 FR 32400).

Since the licensees request conforms to TSTF-358, the staff finds this change acceptable.

G.1.4 Generic Letter 99-02 Charcoal Filter Issue (Open)

G.1.5. A new Shutdown margin methodology (Open)

G.1.6 A new FQA methodology (open)

G.2 Additional BSI Changes identified by the Staff:

G.2.1 SR 3.4.5.2 Verification of SGs capability to remove decay heat The STS SR 3.4.5.2 requires verification of the steam generator secondary side water levels to be greater than or equal to 17 percent (narrow range instrumentation) for both loops. The purpose of this SR is to verify steam generator capability to remove decay heat via natural circulation. For the ITS, the licensee proposal is to verify steam generator secondary side water level by using greater than or equal to 60 percent (wide range instrumentation) for both loops. Since there is no corresponding SR in the CTS, this would be a new SR in the ITS.

The primary function of the RCS is to remove decay heat and transfer this heat via the SG to the secondary plant. Generally, the primary source of heat removal is the residual heat removal system and the secondary source of heat removal would be SGs via natural circulation. The SR 3.4.5.2 requirement ensures the ability of the SG to provide adequate heat sink for decay heat removal and provides assurance that the SG tubes remain covered.

NRC Information Notice 95 - 35, Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation dated August 28, 1995 (IN 95-35) discussed two incidents when SGs were being relied upon for decay heat removal in MODE 5 with loops filled per the TS, however, the SGs were not capable of performing their function because the RCS could not be pressurized. The information notice stated that if the RCS pressure at the top of the SG tubes is allowed to fall below the primary fluid saturation temperature, flashing and steam voiding may occur which interrupts or degrades the natural circulation flow path. The information notice stated that when relying on SGs to remove decay heat the following factors deserve careful

consideration. The plant configuration should have : (1) the ability to pressurize and control pressure in the RCS, (2) the necessary secondary side water level in the SG relied upon for decay heat removal, (3) the availability of a supply of feedwater, and (4) the availability of an auxiliary feedwater pump capable of injecting into the relied-upon SG.

The current ISTS SR that requires that the secondary side water level of one SG is to be greater than or equal to 17 percent is insufficient to ensure that the SG can remove decay heat from the RCS in the applicable conditions. This concern is also applicable to a similar requirement for SG water level in MODES 3 and 4. Industry/TS Task Force is considering revising PWRs ISTS (WOG - 155, Rev.0) to make the required SG capable of removing decay heat requirement via natural circulation consistent during MODES 3,4 and 5.

During its review, the staff requested information regarding how the licensee satisfied the four factors described in IN 95 - 35. In a letter dated June xx, 2002, the licensee described plant procedures that require the operators to keep at least one RCP, the condensate storage tanks and the auxiliary feedwater system OPERABLE during MODE 3 operation. These plant procedures also require the RCS pressure and temperature to be maintained within a prescribed band and the minimum water level in the SGs is to be 28 % on the narrow range.

The SG low - low water level trip and AFW start value is 11.3 % on the narrow range instrumentation. Plant analyses demonstrate that the SGs can support decay heat removal at down to the allowable value since 28 % SG water level is well above the ITS low-low SG allowable value of 11.3 % during MODE 3. These requirements ensure that SGs are capable of removing decay heat in accordance with the guidance of IN 95 - 35. Similarly, the plant operating procedures discuss the plant systems and equipment availability during MODE 4 and MODE 5 with loops filled to remove decay heat via the SGs.

The staff has reviewed the licensee approach to maintain SG tubes covered in order to remove decay heat via natural circulation during the MODES 3,4 and 5. The plant operating procedures ensure that SGs are capable of removing decay heat via natural circulation.

Because this SR enhancement will assure the RCS pressure boundary and fuel integrity, it meets the current regulatory requirements and the guidance of IN 95 - 35 . Therefore, we find the licensee approach reasonable and acceptable.

G.2.2 SR 3.4.1.3 Note - RCS flow rate measurement to be performed after the power level is equal to or greater than 90% of RTP within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during MODE 1 operation. (Open)

G.2.3 Instrument Setpoint methodology and new allowable values In support of the conversion to ITS which involves many changes to instrument allowable values, the licensee submitted the instrument setpoint methodology along with two sample calculations by letter dated March 6, 2002. Additional information was provided by the licensee by letter dated January 31, 2002, in response to the staff request for additional information.

Paragraph (c)(1)(ii)(A) of 10 CFR 50.36, Technical Specifications, requires, in part, that, where a limiting safety system setting is specified for a variable on which a safety limit has been placed, the setting be so chosen that automatic protective action will correct the abnormal situation before a safety limit is exceeded. NRC Regulatory Guide 1.105, Setpoints for Safety-Related Instrumentation, Revision 3 and ANSI Standard ANSI/ISA-S67.04-1994, Setpoints for Nuclear Safety-Related Instrumentation, provide guidance and methodology for determining appropriate settings. The staff has based its acceptance of the licensees setpoint methodology on these regulatory requirements.

The licensee submitted Rev. 0 of Section 3.3.4.1 of Engineering Manual which is the Engineering Design Standard for Instrument Setpoint/Uncertainty Calculations and is used to

calculate instrument allowable values and trip setpoints in the TS. These instrument setpoint calculations resulted in some allowable values and trip setpoints to be less restrictive and some to be more restrictive than the values provided in the current TS. The methodology used by the licensee is based on the general guidelines provided by ISA-S67.04-1987, "Setpoints for Nuclear Safety-Related Instrumentation", and the Two Loop Group Setpoint Methodology developed by Tenera. L. P. However, the licensee did not discuss any deviations from the ISA-S67.04 and the staff has not endorsed the ISA-S67.04-1987 version by a Regulatory Guide (RG). ISA-S67.04-1982 and ISA-S67.04-1994 versions have been endorsed by RG 1.105 Rev. 2 and 3 respectively. Therefore, the staff in its RAI requested the licensee to provide the discussion on how they meet the RG 1.105 Rev. 2 or 3. In its response, the licensee stated that their setpoint methodology meets the requirements of the ISA-S67.04-1994 version. The licensee further stated that with the tacit assumption that vendor reference accuracy is provided with 95% probability and 95% confidence, they meet the RG 1.105 Rev. 3. In its RAI, the staff requested the licensee to provide the basis for the acceptability of the vendor data and what steps are taken if the vendor data were not available. The licensee in its response stated that they have compared vendor data to Westinghouse-provided instrument channel component accuracies and have determined the vendor supplied accuracy values to be equal or more conservative. In addition to this, the licensee has performed drift analyses for various components and generally found that the vendor drift data bound plant-specific drift date for these components. Also the licensee has stated that when vendor data are not available, a plant-specific drift analysis is performed to establish uncertainties for the affected components.

The staff finds the licensees response acceptable.

In its submittal of March 6, 2001, the licensee also provided two sample calculations for the Prairie Island plant which the staff reviewed and determined that the licensee followed the setpoint methodology discussed in the engineering design standard. The proposed trip setpoints and allowable values are intended to maintain acceptable margins between operating conditions and trip setpoints and do not significantly increase the likelihood of a false trip nor failure to trip upon demand. Therefore, the existing licensing basis is not affected. Based on this review, the staff find the setpoint and allowable value determination based on this methodology acceptable.

Based on the above the staff concludes that the licensee's setpoint methodology and the resulting allowable values incorporated in the ITS conversion package are consistent with the Prairie Island Units 1 & 2 licensing basis and are therefore, acceptable.

4.0 COMMITMENTS_RELIED_UPON In reviewing the proposed ITS conversion for PINGP, the staff has relied upon the licensees commitment to relocate certain requirements from the CTS to licensee-controlled documents as described in Table LR, Less Restrictive Chages Relocated Details (Attachment 5 to this SE) and Table R, Relocated Specifications (Attachment 6 to this SE). These tables reflect the relocations described in the licensees submittals on the conversion. The staff requested and the licensee submitted a license condition to make this commitment enforceable (see Section 5.0 of this SE). Such a commitment from the licensee is important to the ITS conversion because the acceptability of removing certain requirements from the TS is based on those requirements being relocated to licensee-controlled documents where further changes to the requirements will be controlled by regulations or other requirements (e.g., in accordance with 10 CFR 50.59).

5.0 LICENSE_CONDITIONS License conditions to define the schedule to begin performing the new and revised SRs after implementation of the ITS are included in the Facility Operating Licenses. These conditions are:

(1) For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of this amendment.

(2) For SRs that existed prior to this amendment, whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after implementation of this amendment.

(3) For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance subject to the modified acceptance criteria is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the implementation of this amendment.

(4) For SRs that existed prior to this amendment, whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the implementation of this amendment.

The staff has reviewed the above schedule for the licensee to begin performing the new and revised SRs and concludes that it is an acceptable schedule. The licensee stated that their implementation date for the new ITS is no later than October 31, 2002. This implementation schedule is acceptable.

Also, a license condition is to be included that will enforce the relocation of requirements from the CTS to licensee-controlled documents. The relocations are described in Table LR, Less Restrictive Chages Relocated Details (Attachment 5 to this SE), and Table R Relocated Specifications (Attachment 6 to this SE). The license condition states that the relocations would be completed no later than October 31, 2002. This schedule is acceptable.

6.0 STATE_CONSULTATION In accordance with the Commission's regulations, the MInnesota State official was notified on June xx, 2002, of the proposed issuance of the ITS conversion amendment for PINGP. The State official had no comments.

7.0 ENVIRONMENTAL_CONSIDERATION Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on July xx, 2002 ( FR ), for the proposed conversion of the CTS to ITS for PINGP. Accordingly, the Commission has determined that issuance of these amendments will not result in any environmental impacts other than those evaluated in the Final Environmental Statement.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner; (2) such activities will be conducted in compliance with the Commission's regulations; and (3) the issuance of the amendment will not be inimical to the

common defense and security, or to the health and safety of the public.

Attachments:

1. List of Acronyms

2. Table A - Administrative Changes

3. Table M - More Restrictive Changes

4. Table L - Less Restrictive Changes

5. Table LR - Less Restrictive Chages Relocated Details

6. Table R - Relocated Specifications Principal Contributors: C. Schulten T. Kim C. Harbuck R. Giardina K. Kavanagh B. Marcus E. Tomlinson S. Saba R. Tjader K. Desai J. Lazevnick R. Goel S. Peters A. Chu P. Loeser H. Garg Date:

LIST_OF_ACRONYMS ACS Air Conditioning System ASME American Society of Mechanical Engineers BSI Beyond-Scope Issue CCW Component Cooling Water CDF Core Damage Frequency CFR Code of Federal Regulations COLR Core Operating Limits Report CTS Current Technical Specification CVCS Chemical and Volume Control System DBA Design-Basis Accident DOC Discussion of Change (from the CTS)

ECCS Emergency Core Cooling System ESFAS Engineered Safety Features Actuation System FBVS Fuel Building Ventilation System FHA Fuel Handling Accident FR Federal Register GDC General Design Criteria ISI Inservice Inspection IST Inservice Testing ITS Improved Technical Specification JFD Justification for Deviation LCO Limiting Condition for Operation LOCA Loss-of-Coolant Accident LSSS Limiting Safety System Setting LTOP Low Temperature Overpressure Protection MCR/ESGR Main Control Room/Emergency Switchgear Room MFIV Main Feedwater Isolation Valve MFPDV Main Feedwater Pump Discharge Valve MFRBV Main Feedwater Regulating Bypass Valve MFRV Main Feedwater Regulating Valve NSSS Nuclear Steam Supply System ODCM Offsite Dose Calculation Manual PAM Post-Accident Monitoring PINGP Prairie Island Nuclear Generating Plant P/T Pressure/Temperature PORV Power Operated Relief Valve PREACS Pump Room Exhaust Air Cleanup System PTLR Pressure Temperature Limits Report QA Quality Assurance RAI Request for Additional Information RCP Reactor Coolant Pump RCS Reactor Coolant System RG Regulatory Guide RHR Residual Heat Removal RPV Reactor Pressure Vessel RTS Reactor Trip System RWST Refueling Water Storage Tank SE Safety Evaluation ATTACHMENT 1

LIST_OF_ACRONYMS (Continued)

SI Safety Injection SG Steam Generator SGTR Steam Generator Tube Rupture SR Surveillance Requirement STS Improved Standard Technical Specification, NUREG-1431, Rev. 1 SW Service Water TRM Technical Requirements Manual TS Technical Specification TSTF Technical Specifications Task Force (re: generic changes to the STS)

UHS Ultimate Heat Sink USAR Updated Safety Analysis Report

ATTACHMENT 2 Table A - Administrative Changes

ATTACHMENT 3 Table M - More Restrictive Changes

ATTACHMENT 4 Table L - Less Restrictive Changes

ATTACHMENT 5 Table LR - Less Restrictive Relocated Details

ATTACHMENT 6 Table R - Relocated Specifications

v t e Letter Sequence Draft Approval
TAC:MB0695, Ventilation System Envelope Allowed Outage Time (Approved, Closed) TAC:MB0696, Ventilation System Envelope Allowed Outage Time (Approved, Closed)
Initiation Acceptance, Acceptance, Acceptance, Acceptance, Acceptance Administration Questions 8 January 2002 Answers Results Approval, Approval, Approval, Approval, Approval, Approval, Approval Other: ML021830089, ML022070613, ML022070658, ML022140394
MONTHYEAR2002-06-03 [Table View]

ML021540301

Text

1.0 INTRODUCTION

2.0 BACKGROUND

8.0 CONCLUSION

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