ML020700361

Part 7 of 10, North Anna Power Station, Units 1 & 2, Proposed Improved Technical Specifications Comments on Draft Safety Evaluation, Certified Improved Technical Specifications (ITS) & Bases & Proposed License Conditions, Attachment

ML020700361
Person / Time
Site:	North Anna
Issue date:	02/22/2002
From:	Hartz L Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
02-053, CM/RAB R0, TAC MB0799, TAC MB0800
Download: ML020700361 (155)

Text

Attachment Proposed Improved Technical Specifications Comments on Draft Safety Evaluation Virginia Electric and Power Company (Dominion)

North Anna Power Station Units I and 2

DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. XXX TO FACILITY OPERATING LICENSE NO. NPF-4 AND AMENDMENT NO. XXX TO FACILITY OPERATING LICENSE NO. NPF-7 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-338 AND 50-339

1.0 INTRODUCTION

North Anna Power Station, Units 1 and 2 (NAPS) have been operating with Technical Specifications (TS) issued with the original Operating Licenses on

,,tobcr 5, 1970 November 26, 1977(for Unit 1), and Ma--h 8, 1973*April 11, 1980 (for Unit 2), as amended. By application dated December 11, 2000, as supplemented by letters dated May 30, June 18, July 16, July 20, August 13, August 27, September 27, October 10, October 17, November 8, November 19, November 29, December 3, December 7, December 12, and December 13, 2001, and January 2, January 25, and January 31, 2002, Virginia Electric and Power Company (the licensee) requested an amendment to the Operating Licenses and TS for the NAPS. Hereinafter, the proposed improved TS for NAPS 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 SE is provided in Attachment 1 to this SE. This proposed amendment would convert the CTS to ITS.

The proposed conversion to the ITS is based upon:

NUREG-1 431, "Standard Technical Specifications for Westinghouse Plants," Revision 1, dated April 1995; The current NAPS 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).

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 submittal letters of request for additional information (RAI) and a series of telephone conference calls were required during the course of this review. These RAIs and conference calls were necessary for the staff to clarify the proposed ITS with respect to the guidance in the Final Policy Statement and the STS and the NAPS's plant specific features and design. In addition to information from these RAIs and 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 Institute's Technical Specifications Task Force (TSTF).

These generic issues were considered for specific applications in the NAPS ITS. Consistent with the Final Policy Statement, the licensee proposed transferring some CTS requirements to licensee-controlled documents (such as the NAPS Updated Final Safety Analysis Report (UFSAR), 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 NAPS in accordance with 10 CFR 50.36. Since the licensee submitted the December 11, 2000, application, a number of amendments to the NAPS operating license have been approved. The following table provides the subjects of the amendments and the dates of issuance.

Amendment Nos.

Unit 1 Unit Description of Change Date 2

225 206 Increase Boron Concentration Limits in Reactor Coolant 3/20/01 System during Refueling and Establish Boron Limits for Spent Fuel Pool.

226 207 Pressure-Temperature Limits, Low Temperature 5/02/01 Overpressure Protection (LTOP) System Setpoints, and LTOP System effective temperature.

227 208 Increase Fuel Enrichment and Spent Fuel Pool Soluble 6/15/01 Boron and Fuel Burnup Credit 228 209 Control Room Emergency Habitability Systems Increase 12/12/01 Number of Compressed Air Bottles and revise Differential pressure Limit for Filter Assemblies 229 210 Elimination of Post Accident Sampling System 12/19/01 Requirements 230 Delete Obsolete License Conditions for North Anna Unit 1 1/31/02 The licensee has incorporated these amendments, as appropriate, into the ITS.

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 Commission's proposed action on the NAPS application for amendment dated December 11, 2000 was published in the Federal Register on XXXXX (xx FR xxxxx).

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 staff 's 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 NAPS. This SE also explains the staff 's conclusion that the ITS, which are based on the STS as modified by plant-specific changes, are consistent with the NAPS current licensing basis and the requirements of 10 CFR 50.36.

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 NAPS. The staff approves the licensee's changes to the CTS with modifications documented in the licensee's 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 plant's 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 "Writer's 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 NAPS.

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. Further, 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), ALAB-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 applicant's 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 staff 's conclusion that the conversion of the NAPS CTS to ITS based on STS, as modified by plant-specific changes, is consistent with the NAPS 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 NAPS 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.

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 staff 's comments were documented as requests for additional information (RAIs) and forwarded in letters dated April 23, May 21, June 1, June 4, June 22, July 2, July 30, July 31, September 6, September 7, September 18, October 3, October 10, October 16, November 7, and December 7, 2001. The licensee provided responses to the RAIs in supplemental letters dated By

,pplic-ti-n dat*d May 30, June 18, July 16, July 20, August 13, August 27, September 27, October 10, October 17, November 8, November 19, November 29, December 3, December 7, December 12, and December 13, 2001, and January 2, January 25, and January 31, 2002. The letters clarified the licensee's basis for translating the CTS requirements into ITS. For items that have been reviewed by the staff as stated in this Draft Safety Evaluation, 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.

Following are the five types of CTS changes:

A Administrative - changes to the CTS that result in no changes to existing restrictions and flexibility (i.e., nontechnical changes in the presentation of CTS requirements).

M More Restrictive - changes to the CTS that result in added restrictions or reduced flexibility (i.e., additional TS requirements).

L Less Restrictive "Specific" - changes to the CTS that result in reduced restrictions or added flexibility (i.e., changes, deletions, and relaxations of CTS requirements).

LA Removed Details - changes to the CTS that move details out of the CTS and into the Bases, UFSAR, or other appropriate licensee-controlled documents (i.e., design details, system descriptive details, and procedural details). This type of change is included with ReIGeatieRs-Relocated Specifications in Table R as described below.

R ReleGatieR-sRelocated Specifications - relaxations in which whole CTS specifications are removed from the CTS to licensee-controlled documents.

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 difference (JFD).

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

"* Table A - Administrative (A) Changes to the CTS

"* Table M - More Restrictive (M) Changes to the CTS

"* Table L - Less Restrictive (L) Changes to the CTS

"* Table R - Relocated Specifications (R) and Removed Details (LA) from the CTS 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 licensee's application and supplemental letters.

A. Administrative Changes Administrative (nontechnical) 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 which 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 NAPS 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 Writer's Guide and the STS, do not result in any change in operating requirements, and are consistent'with the Commission's regulations.

B. Technical Changes - More Restrictive 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 which 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 NAPS 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. Technical Changes - Less Restrictive 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 the 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 NAPS design was also reviewed to determine if the specific design basis and licensing basis for NAPS are consistent with the technical basis for the model requirements in the STS, and thus provide a basis for the ITS.

All of the less restrictive changes to the CTS have been evaluated. The majority of the changes were-and found to involve deletions and relaxations to portions of the CTS requirements that can grouped in the following seve'-eight categories:

Relaxation of LCO Requirement (Category 1)

"* Relaxation of Applicability (Category 2)

"* Relaxation of Completion Time (Category 3)

"* Relaxation of Required Action (Category 4)

Deletion of Surveillance Requirements (Category 5)

Relaxation of Surveillance Requirements Acceptance Criteria (Category 6)

"* Relaxation of Surveillance Frequency (Category 7)

"* Deletion of Reporting requirements (Category 8)

The following discussions address why portions of various specifications within each of these eight categories of information or specific requirements are not required to be included in ITS:

1.

Relaxation of LCO Requirement (Category 1)

The CTS contain LCOs that are overly restrictive because they specify limits on operational and system parameters and on system operability beyond those necessary to meet safety analysis assumptions. The CTS also contain administrative controls that do not contribute to the safe operation of the plant. The ITS, consistent with the guidance in the STS, omit such operational limits and administrative controls. This category of change includes: (1) deletion of equipment or systems addressed by the CTS LCOs that are not required or assumed to function by the applicable safety analyses; (2) addition of explicit exceptions to the CTS LCO requirements (e.g., modc ntry rr*ic*tions cquiv.-alcnt to those of ITS LCO 3.0.)' consistent with the guidance of the STS and normal plant operations to provide necessary operational flexibility but without a significant safety impact; and (3) deletion of miscellaneous administrative controls such as

.e-*,,*g FequFemeRt_-, sometimes contained in action requirements, that have no effect on safety.

Deletion of such administrative controls allows operators to more clearly focus on issues important to safety. The ITS LCOs and administrative controls resulting from these changes will continue to maintain an adequate degree of protection consistent with the safety analysis, while providing an improved focus on issues important to safety and necessary operational flexibility without adversely affecting the safe operation of the plant.

Therefore, these less restrictive changes, which are consistent with STS and fall within Category 1, are acceptable.

2.

Relaxation of Applicability (Category 2)

Reactor operating conditions are used in CTS to define when LCO features are required to be operable. CTS applicability requirements can be specifically defined terms of reactor conditions, such as hot shutdown, cold shutdown, reactor critical, or power operating conditions. CTS applicability requirements can also be more general.

Depending on the circumstances, the CTS may require that an LCO be maintained within limits in "all modes" or "any operating mode." Generalized applicability conditions are not contained in STS; therefore, ITS eliminates CTS requirements such as "all modes" or "any operating mode," replacing them with ITS-defined modes or applicable conditions that are consistent with the application of the plant safety analysis assumptions for operability of the required features.

In another application of this category of change, CTS requirements may be eliminated during conditions for which the safety function of the specified safety system is met because the feature is performing its intended safety function. Deleting applicability requirements that are indeterminate or which are inconsistent with application of accident analyses assumptions is acceptable because when LCOs cannot be met, the TS are satisfied by exiting the specified LCO's applicability, thus taking the plant out of the conditions that require the safety system to be operable. Therefore, these changes, which are consistent with STS and fall within Category 2, are acceptable.

3.

Relaxation of Completion Time (Category 3)

Upon discovery of a failure to meet an LCO, the 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. 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 operable 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.

Therefore, required action completion time extensions, which are consistent with STS and fall within Category 63, are acceptable.

4.

Relaxation of Required Actions (Category 4)

An LCO is the lowest functional capability or performance level of equipment required for safe operation of the facility. When an LCO is not met, the 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, the ITS actions result in extcnding the time pcriedless restrictive requirements for taking the plant outside the applicability into shutdown conditions. For example, changes in this category include providing an option to (1) isolate a system, (2) place equipment in the state assumed by the safety analysis, (3) satisfy alternate criteria, (4) take manual actions in place of automatic actions, (5) "restore to operable status" within a specified time frame, (6) place alternate equipment into service, or (7) 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. Therefore, these changes, which are consistent with STS and fall within Category 4, are acceptable.

5.

Deletion of Surveillance Requirements (Category 5)

CTS require maintaining the LCO equipment operable by conducting SRs in accordance with the plant specific equipment. The changes in this type relate to elimination of surveillance requirements in CTS that were no longer required er--because equipment had been replaced or the features that required surveillance actions had been replaced, or features with surveillance activities te-bewere duplicated by other new ITS requirements. These changes fall in Category 5 and are consistent with the STS, and therefore are acceptable.

6.

Relaxation of Surveillance Requirements Acceptance Criteria (Category 6)

Relaxation of CTS SR acceptance criteria provide operational flexibility, consistent with the guidance of the STS, but do not reduce the level of assurance of operability provided by the successful performance of the surveillance. Such revised acceptance criteria are acceptable because they remain consistent with the application of the plant safety analysis assumptions for operability of the LCO-required features.

Relaxation of CTS SR performance conditions include not requiring testing of de energized equipment (e.g., instrumentation channel checks) and equipment that is already performing its intended safety function (e.g., position verification of valves locked in their safety actuation position). These changes are acceptable because the existing surveillances are not necessary to ensure the capability of the affected components to perform their intended functions. Another relaxation of SR performance conditions is the allowance to verify the position of valves in high radiation areas by administrative means.

This change is acceptable because licensee controls regarding access to high radiation areas make the likelihood of mispositioning such valves negligible. Therefore, these changes, which are consistent with STS and fall within Category 6, are acceptable.

Upon di"oaery of a failurc to mcct an LOC, TS specify times for completing Required Actions of the as-ociated T-S oRnditiens. Required Ac*t*ios establish remedial measures that must be taken w.ithin specified completion times (allowed outage times). These times define li*itS during which operation in a degraded condition is per,*i.d.

Incor.porating completion time e.tensions is a.ceptable because completion ti*es take int o account the operability status of the redundant systems of TS r.eqUied features, the capacity and capability Of remnaining features, a reasonable time for repairs or replacemnent of required features, vendor developed standard repair times, and the low probability of a design basis accident (DBA) occSurring during the repair period. These c~hanges are consistent with STS, and allowed outage time extensions specified as Type 6 are acceptable.

7.

Relaxation of Surveillance Frequency (Category 7)

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 frequency thereafter, the CTS require verifying 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. These CTS SR relaxations are consistent with the STS.

Relaxations of CTS SR accGeptance criteria pro)vide operational flexibility, consistent with the guidance of the ST-S, but do not reduce the level of assurance of operability provide by the successful performance of the suryeillance. Such revised acceptance criteriaar f ability of the LCO required features.

S................. LI I*

W I *1 B I v e, aRE

-*r iS

... fG Relaxations of CTS SR performnan~e conditions include not requiring testing of de energized equipmeint (e.g., intuetto hannel checks) and equipment that is alFeady petforming its intended safety funci (e.g., pesition aerification of valves loeke in theif safety actuation position). These changes are acceptable because the existing suhseillanies are not necessay to ensure the capability of the affeted components to peafnrm their intended functions. Anotherf elaxatine of SR peformane conditions sit th allowance to verify the position of valves in high radiation areas by adm~inistrative m~eanS.

Trhis chage is acgeptable because lieng see onCtrols regaredng access te high radiati areas make the likelihoed ef mipoCatiening such valves negligible-.

Finally, 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. Therefore, because of the reasons stated, less restrictive changes to CTS SRs falling within Category 7 are acceptable.

8.

Deletion of Reporting requirements (Category 8)

CTS include requirements to submit special reports to the NRC when specified limits or conditions are not met. Typically, the time period for the report to be issued is "within 30 days." However, the ITS eliminates the TS requirements for special reports and instead relies on the reporting requirements of 10 CFR 50.73. The changes to the reporting requirements are acceptable because 10 CFR 50.73 provides adequate reporting requirements, and the special reports do not affect continued plant operation.

CTS also include requirements for reports to be made to the NRC on data gathered as part of routine plant programs. These requirements are removed from the ITS. The requirement to report test frequency changes that occur due to consecutive SR failures has been deleted since the test schedule is already covered by the TS. In addition, a historical review has shown the SR has never failed. These changes are consistent with STS, are specified as Type 8, aTheaFeand are acceptable.

For the reasons presented above, these less restrictive requirements are acceptable because they will not affect the safe operation of the plant. The ITS requirements are consistent with current licensing practices, 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 NAPS ITS conversion. Table L 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 as 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 Commission's regulations Table L includes all L changes and is organized by ITS section. The table specifies: the section designation; a summary description of the change; CT-S and ITS LCO references; a refeerence to the specific change type as discussed above; and a characterization of the DOG.

Table L is organized in STS o

,dr by each L type DOC. For each change, the table lists (1) the DOG identifier (e.g., 3.1.1 followed by Li mncans STS 3.1.1, DOG LI); (2) a summar~ny deScriptinr of the cshange; (3) the refcencRAe nIUMbcrs of the aSSociatcd ITS rcqUircments; (4) the rcfc~Rone numbers of the associated CTS requirements; and (5) the less rcStrictiy.e change category.

D. Technical Changes - Less Restrictive Femoval-Removal of Details (R-and-LA)

When requirements have been shown to give little or no safety benefit, their removal from the TS may be appropriate. These are grouped as LA changes in the R Tables. 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 NAPS 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 5 that follow:

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 UFSAR 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 (UFSAR Chapter 17). In 10 CFR 50.59, controls are specified for changing the facility as described in the UFSAR (including the Technical Requirements Manual, (TRM)), and in 10 CFR -50.54(a) criteria are specified for changing the QA plan. The TRM is a general reference in the UFSAR, and is subject to the administrative controls that include the requirement to perform 1 OCFR50.59 evaluations for changes made to the TRM. This is consistent with NEI 98 03 Revision 1,"Guidelines for Updating Final Safety Analysis Reports," which the NRC endorsed, without exception, in Reg. Guide 1.181, dated September 1999. The ITS Bases also contain descriptions of system design. ITS 5.5.11 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 UFSAR, which references the TRM,(incIwdR§ TRM-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). ITS Section 5.6, Administrative Cor,-olsReporting Requirements, includes the programmatic requirements for the COLR.

Type 2 - Removing Descriptions of System Operation The plans for the normal and emergency operation of the facility are required to be described in the UFSAR by 10 CFR 50.34. ITS 5.4.1.a requires written procedures to be established, implemented, and maintained for plant operating procedures including procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978. Controls specified in 10 CFR 50.59 apply to changes in procedures as described in the UFSAR. The ITS Bases also contain descriptions of system operation.

The NAPS CTS include instrumentation trip setpoints and Allowable Values. Trip setpoints are instrument field settings. Allowable Values are the limiting values of the instrument trip setpoints before the LCO is exceeded, and the relationship between the trip setpoints and the Allowable Values is determined through the setpoints methodology approved by the staff. Trip setpoints are system operation details that can be adequately controlled by licensee-controlled documents without adversely affecting safe operation of the plant. Allowable Values are specified in the ITS, while trip setpoints are relocated to the TRM.

It is acceptable to remove details of system operation from the TS because this type of information will be adequately controlled in the UFSAR,

'k"h'"4i"49which references the TRMJ and the TS Bases, as appropriate.

Type 3 - Removing Procedural Details for Meeting TS Requirements and Related Reporting Details for performing TS Actions and SRs are more appropriately specified in the plant procedures required by ITS 5.4.1, and described in the UFSAR and ITS Bases. For example, control of the plant conditions appropriate to perform a surveillance test is an issue for procedures and scheduling and has previously been determined to be unnecessary as a TS restriction. As indicated in GL 91-04, allowing this procedural control is consistent with the vast majority of other SRs that do not dictate plant conditions for surveillances. Prescriptive procedural information in an ITS action requirement is unlikely to contain all procedural considerations necessary for the plant operators to complete the actions required, and referral to plant procedures is therefore required in any event. Other changes to procedural details include those associated with limits retained in the ITS. For example, the ITS requirement may refer to programm..atic requirements Guha OL=R, included in ITS Section 5.6, which spccifics the Scope of the limitS contained inthe COLR and mandates NRC approval of the anal'Aica Sethedelogy-The QA Program is approved by the NRC and contained in UFSAR Chapter 17, and changes to the QA Program are controlled by 10 CFR 50.54(a). The Offsite Dose Calculation Manual (ODCM) is required by ITS 5.5.1. The TRM is WAeeF eated-lay-referenced in te-the UFSAR, and changes to the TRM are controlled by 10 CFR 50.59. The Inservice Test (IST) program is required by ITS 5.5.7.

Type 4 - Removing Performance Requirements for Indication-Only Instrumentation and Alarms Details for performance requirements for indication-G4y-only Instr'Umcntat~ons instrumentation and AlaF-ms-alarms are more appropriately specified in the plant procedures required by ITS 5.4.1, the UFSAR, and the Bases. For example, CTS 4.6.1.1.d states, "Each time containment integrity is established after vacuum has been broken by pressure testing the butterfly isolation valves in the containment purge lines and the containment vacuum ejector line." ITS SR 3.6.3.4 states, "Perform leakage rate testing for containment purge valves with resilient seals." This changes the CTS by moving the detaileds performance requirement, specifically naming butterfly valves and the containment vacuum air ejector line, to the Bases. Prescriptive procedural information in an action requirement is unlikely to contain all procedural considerations necessary for the plant operators to complete the actions required, and referral to plant procedures, based on TS Bases is therefore required in any event. The removal of these kinds of procedural details from the CTS is acceptable because they will be adequately controlled by NRC requirements, the UFSAR, plant procedures, and the Bases, as appropriate. This approach provides an effective level of regulatory control and provides for a more appropriate change control process. Removal of requirements for indication only instrumentation is acceptable because such instrumentation usually does not support system operability. Therefore, it is acceptable to remove Type 4 details from the CTS and place them in licensee-controlled documents.

Type 5 - Removal of Cycle-Specific Parameter Limits from the Technical Specifications to the Core Operating Limits Report Other changes to procedural details include those associated with limits retained in the ITS. For example, the ITS requirement may refer to programmatic requirements such as COLR, included in ITS Section 5.56, which specifies the scope of the limits contained in the COLR and mandates NRC approval of the analytical methodology. Removal of requirements for programmatic requirements such as COLR is acceptable because such program usually does not support system operability. Therefore, it is acceptable to remove Type 5 details from the CTS and place them in licensee-controlled documents with references to ITS Se-tieA-Chapter 5.0.

Table R lists the less restrictive removal of detail changes being made in the NAPS ITS conversion. Table R is organized in STS order by each LA-type and R-type DOC. It includes the following: (1) the DOC identifier (e.g., 3.1.1 followed by LA1 means STS 3.1.1, DOC LA1); (2) the reference numbers of the associated CTS requirements; (3) a summary description of the relocated details and requirements; (4) the name of the licensee-controlled document to contain the relocated details and requirements (location); (5) the regulation (or ITS Specification) for controlling future changes to relocated requirements (change control process); and (6) a characterization of the type of change (not applicable to R-type DOCs).

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."

UFSAR (which iRGI*edes-references the TRM as)Appendi 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.

OGDQM-ODCM controlled by ITS 5.5.1.

COLR controlled by ITS 5.6.5.

PTLR *ontroulcd by ITS 5.6.5.

QA Plan, as approved by the NRC and referenced in the UFSAR, 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 (R) from the CTS 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 of the SE 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, which is referenced in the UFSARUFSAR, which incudcc the TRM, the Environmcntal Manual (E.), and the ODCM, as appropriate. The NRC staff has reviewed the licensee's submittals and finds that relocation of these requirements to the UFSAR, TRM, EM, and ODCM is acceptable in that changes to the UFSAR, TRM, EM, and 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 all specifications that are being relocated from the CTS to licensee-controlled documents. Table R is combined with LA; however the relocated LA items are organized as described in Section 3.0.D above.

Table R lists the relocated changes being made in the NAPS ITS conversion. Table R 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 of Attachments to the SE.

1. 3.1.1.3.1 BORON DILUTION - Reactor Coolant Flow CTS 3.1.1.3.1 requires a minimum reactor coolant system flow of 3000 gpm in all MODES. Various accident analyses assume adequate reactor coolant flow for heat removal and boron mixing. However, a specific flow rate is not assumed as an initial condition of any design basis accident or transient and is not credited for mitigation of any design basis accident or transient. Other specifications in the ITS contains adequate controls to ensure that RCS flow meets the general accident analysis assumption. In MODES 1, 2, and 3, at least one Reactor Coolant Pump (RCP) is required to be in operation, which provides flow in excess of 3000 gpm. In MODE 4, either an RCP or Residual Heat Removal (RHR) train is required to be in operation, and in MODES 5 and 6, at least one RHR train is required to be in operation. The ITS Bases state that when an RHR train is required to provide RCS flow, the flow rate must be sufficient for decay heat removal and boron mixing. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Boron Dilution - Reactor Coolant Flow LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

2. 3.1.2.1 FLOW PATHS - Shutdown CTS 3.1.2.1 provides requirements on the boration systems flow paths during shutdown. The boration systems are part of the Chemical and Volume Control System (CVCS) and provides the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The boration system is not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Flow Paths - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

3. 3.1.2.2 FLOW PATHS - Operating CTS 3.1.2.2 provides requirements on the boration systems flow paths during operation. The boration systems are part of the CVCS and provides the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The boration system is not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient.

The Emergency Core Cooling System (ECCS) and Refueling Water Storage Tank are credited in the accident analyses. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Flow Paths - Operating LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

4. 3.1.2.3 CHARGING PUMP - Shutdown CTS 3.1.2.3 provides requirements on the charging pumps during shutdown when used as part of the boration system. The charging pumps in the boration system are part of the CVCS and provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The charging pumps in the boration system are not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient.

In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. OPERABILITY of the charging pumps is required as part of the Emergency Core Cooling System, which is addressed in other specifications.

The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Charging Pump - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

5. 3.1.2.4 G

g,,*,

p6,,..,CHARGING PUMPS - Operating CTS 3.1.2.4 provides requirements on the charging pumps during operation when used as part of the boration system. The charging pumps in the boration system are part of the CVCS and provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The charging pumps in the boration system are not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient.

The ECCS is and Refueling Water Storage Tank are credited in the accident analyses. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. OPERABILITY of the charging pumps is required as part of the ECCS, which is addressed in other specifications. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Charging Pumps - Operating LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

6. 3.1.2.5 Unit 1; BORIC ACID TRANSFER PUMPS - Shutdown Unit 1 CTS 3.1.2.5 provides requirements on the boric acid transfer pumps during shutdown. The boric acid transfer pumps are part of the CVCS and provides the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The boric acid transfer pumps are not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Boric Acid Transfer Pumps - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

7. 3.1.2.6 Unit 1; BORIC ACID TRANSFER PUMPS - Operating Unit 1 CTS 3.1.2.6 provides requirements on the boric acid transfer pumps during operation. The boric acid transfer pumps are part of the CVCS and provides the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The boric acid transfer pumps are not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient. The ECCS and Refueling Water Storage Tank are credited in the accident analyses. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Boric Acid Transfer Pumps - Operating LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

8. 3.1.2.7 BORATED WATER SOURCES - Shutdown CTS 3.1.2.7 provides requirements on the borated water sources during shutdown.

The borated water sources - shutdown are part of the CVCS and provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The borated water sources are not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Borated Water Sources - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

9.3.1.2.8 BORATED WATER SOURCES - Operating CTS 3.1.2.8 provides requirements on the borated water sources during operation.

-The borated water sources - operating are part of the CVCS and provide the means to control the chemical neutron absorber (boron) concentration in the RCS and to help maintain the shutdown margin. The borated water sources are not assumed to be OPERABLE to mitigate the consequences of a design basis accident or transient. The ECCS and Refueling Water Storage Tank are credited in the accident analyses and are required by other specifications. In the case of the boron dilution accident, the accident is addressed by preventing its occurrence or by terminating the event before the required shutdown margin is lost, not by boration. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Borated Water Sources - Operating LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

10.3.1.3.3 POSITION INDICATOR CHANNELS - Shutdown CTS 3.1.3.3 provides requirements on the rod position indicator channels during shutdown (MODES 3, 4, and 5 with the reactor trip system breakers in the closed position). The control rod position indicator channels provide ORdiGatei indication of rod position to the operator. This indicator is used by the operator to verify that the rods are correctly positioned, and to verify the rods are inserted into the core following a reactor trip. Rod position indicator is also used during reactor startup.

However, no DBA or Transient initiated in MODES 3, 4, or 5 with the reactor trip system breakers in the closed position assumes operator action to manually trip the reactor or to take some alternative action if an automatic reactor trip does not occur. With the reactor critical, rod position indicator is used to verify that the insertion, sequence, and overlap limits are met. These are related to SHUTDOWN MARGIN and core power distribution limits. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Position Indicator Channels - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

11.3.3.3.1 RADIATION MONITORING INSTRUMENTATION CTS 3.3.3.1 states the radiation monitoring instrumentation channels shown in Table 3.3-6 shall be OPERABLE with their alarm/trip setpoints within the specified limits. Portions of the Radiation Monitoring Instrumentation specification, as shown in the CTS markup, are addressed in ITS 3.4.15, RCS Leakage Detection Instrumentation, and ITS 3.3.3, Post Accident Monitoring (PAM) Instrumentation.

Those portions are not addressed in this change. The Radiation Monitoring Instrumentation monitors radiation levels in selected plant locations and indicates abnormal or unusually high radiation levels. The radiation monitors are not assumed in the accident analyses to provide signals to prevent initiation of a DBA or transient or to mitigate a DBA or transient. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Radiation Monitoring LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

12.3.3.3.2 MOVABLE INCORE DETECTORS CTS 3.3.3.2 provides requirements on the Movable Incore Detector Instrumentation when required to monitor the flux distribution within the core. The Movable Incore Detector System is used for periodic surveillance of the power distribution, and for calibration of the excore detectors. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Movable Incore Detectors LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

13. 3.3.3.3 SEISMIC INSTRUMENTATION CTS 3.3.3.3 for Unit 1 states the Seismic Monitoring Instrumentation shown in Table 3.3-7 shall be OPERABLE. The Seismic Monitoring Instrumentation is used to record data for use in evaluating the effect of a seismic event. The Seismic Monitoring Instrumentation is not used to mitigate a DBA or transient. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Seismic Instrumentation LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

14.3.3.3.4 METEOROLOGICAL INSTRUMENTATION CTS 3.3.3.4 for Unit 1 states the Meteorological Monitoring Instrumentation shown in Tables 3.3-8 and 4.3-5 shall be OPERABLE. The Meteorological Monitoring Instrumentation is used to record meteorological data for use in evaluating the effect of an accidental radioactive release from the plant. The Meteorological Monitoring Instrumentation is not used to mitigate a DBA or transient. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Meteorological Instrumentation LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

15.3.3.3.9 LOOSE PARTS MONITORING SYSTEM Unit 1 CTS 3.3.3.9 requires the OPERABILITY of the loose parts detection instrumentation which can detect loose metallic parts in the Reactor Coolant System in order to avoid damage to the Reactor Coolant System components. The Unit 2 Technical Specifications do not contain this Specification. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Loose Parts Monitoring System LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

EXPLOSIVE GAS MONITORING INSTRUMENTATION

16. 3.3.3.11 CTS 3.3.3.11 requires the Explosive Gas Monitoring Instrumentation be OPERABLE. The Explosive Gas Monitoring Instrumentation is used to ensure that the oxygen limits of the Waste Gas Holdup System are not exceeded. The oxygen concentration limit in the Waste Gas Holdup Tank ensures that the concentration of potentially explosive gas mixtures in the Waste Gas Holdup System is maintained below the flammability limits. This instrumentation is not credited in preventing or mitigating any DBA or transient as the safety analysis concerning the Waste Gas Holdup System assumes a storage tank rupture with no mitigation. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Explosive Gas Monitoring Instrumentation LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

17.3.4.6.3 PRIMARY TO SECONDARY LEAKAGE CTS 3.4.6.3 provides limits on primary to secondary leakage in addition to the limits in CTS 3.4.6.2 and ITS 3.4.13. These additional limits lower the amount of allowed primary to secondary leakage when the reactor is operating above 50%

power and were implemented to reduce the probability of a steam generator tube rupture following the Unit 1 steam generator tube rupture event at NAPS Unit 1 on July 15, 1987. The CTS 3.4.6.2 leakage limits weFe-continued to be used in the accident analysis, not the additional limits in CTS 3.4.6.3. The NAPS Units 1 and 2 steam generators have been replaced with models that are not susceptible to the fatigue induced cracks which resulted in the tube rupture. As a result, these additional limits are not needed to lower the probability of a steam generator tube rupture. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Primary to Secondary Leakage LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

18 3.4.6.4 PRIMARY TO SECONDARY LEAKAGE DETECTION SYSTEMS CTS 3.4.6.4 states requirements on primary to secondary leakage detection systems. These leakage detection systems are in addition to those systems required by CTS 3.4.6.1 and ITS 3.4.15 and were installed to monitor the stringent primary to secondary leakage limits in CTS 3.4.6.3. These additional primary to secondary leakage detection systems were added to the Technical Specifications following the Unit 1 steam generator tube rupture (SGTR) event at NAPS Unit 1 on July 15, 1987. Subsequently, the NAPS Units 1 and 2 steam generators have been replaced and steam generator primary to secondary leakage is insignificant.

As a result, the requirements in ITS 3.4.15 are sufficient to indicate significant abnormal RCS leakage. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Primary to Secondary Leakage Detection Systems LCO and Surveillances -may be relocated to other plant controlled documents outside the ITS.

19. 3.4.7 CHEMISTRY CTS 3.4.7 provides limits on the oxygen, chloride and fluoride content in the RCS to minimize corrosion. Minimizing corrosion of the RCS will reduce the potential for RCS leakage or failure due to stress corrosion, and ultimately ensure the structural integrity of the RCS. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Chemistry LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

20.3.4.9.2 PRESSURIZER CTS 3.4.9.2 states that the pressurizer temperature shall be limited to a maximum heatup of 100_F or cooldown of 2000-F in any one hour period and a maximum spray water temperature and pressurizer temperature differential of 320°-F. The pressurizer temperature limits are placed on the pressurizer to prevent non-ductile failure. The limits meet the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Pressurizer LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

21. 3.4.10.1 STRUCTURAL INTEGRITY-ASME Code Class 1, 2 & 3 Components CTS 3.4.10.1 provides requirements for the ASME Code Class 1, 2 and 3 components to ensure their structural integrity. These requirements are in addition to the requirements in CTS 4.0.5. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Structural Integrity ASME Code Class 1, 2 & 3 Components LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

22. 3.4.11.1 REACTOR VESSEL HEAD VENT CTS 3.4.11.1 provides requirements on the reactor vessel head vents. The reactor coolant head vents are provided to exhaust noncondensible gases or steam, which could inhibit core cooling, from the RCS. The reactor vessel head vents are not credited in any UFSAR accident analysis. The reactor vessel head vents are included in the Emergency Operating Procedures for mitigation of beyond design basis accidents. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Reactor Vessel Head Vent LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

23.3.5.4.2 HEAT TRACING CTS 3.5.4.2 states, "At least two independent channels of heat tracing shall be OPERABLE for the boron injection tank and for the heat traced portions of the associated flow paths." The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Boron Injection Tank Heat Tracing LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

STEAM TURBINE ASSEMBLY 24.3.7.1.6 CTS 3.7.1.6 states that the structural integrity of the steam turbine assembly shall be maintained in MODES 1 and 2. The steam turbine assembly is used to provide the motive force for the main electrical generator. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Steam Turbine Assembly LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

25. 3.7.1.7 TURBINE OVERSPEED CTS 3.7.1.7 states that at least one turbine overspeed protection system shall be OPERABLE in MODES 1, 2, and 3. The turbine overspeed protection system is used to prevent a turbine overspeed condition that could result in turbine damage.

The turbine overspeed protection system serves no accident mitigation function in any MODE. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Turbine Overspeed LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

26.3.7.2.1 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION CTS 3.7.2.1 states that the temperature of both the primary and secondary coolants in the steam generators shall be greater than 70O1 F when the pressure of either coolant in the steam generator is greater than 200 psig at all times. The Steam Generator Pressure/Temperature Limitation serves no accident mitigation function in any MODE. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Steam Generator Pressure /

Temperature Limitation LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

27.3.7.3.1 COMPONENT COOLING WATER SUBSYSTEM -

Operating CTS 3.7.3.1 states that three component cooling (CC) water system loops shall be OPERABLE. It is applicable when either unit is in MODES 1, 2, 3, or 4. The primary function of the CC System is to provide cooling water to the RHR heat exchangers. Unlike other Westinghouse plants, the RHR at NAPS does not share components with the Emergency Core Cooling System (ECCS), and thus does not play a role in DBA mitigation. At NAPS, this post-accident heat removal function is provided primarily by the Recirculation Spray System and the Low Head Safety Injection pumps. For this reason, CC is not required for DBA mitigation, and, like RHR, does not meet Criterion 3 of 10 CFR 50.36(c)(2)(ii) for retention in the Technical Specifications for MODES 1, 2, 3, and 4. Other plants use CC for DBA mitigation functions other than ECCS, such as containment cooling, but the CC system at NAPS does not. This makes the CC System at NAPS different from the CC System described in the IST, and retaining the CC requirement for supporting RHR or any other components not assumed in DBA analysis is inappropriate. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Component Cooling Water Subsystem - Operating LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

28.3.7.3.2 COMPONENT COOLING WATER SUBSYSTEM - Shutdown CTS 3.7.3.2 states that two CC loops shall be OPERABLE. It is applicable when both units are in MODES 5 or 6. The primary function of the CC System is to provide cooling water to the RHR heat exchangers, but does not warrant its own LCO. If insufficient CC is available for RHR, RHR is declared inoperable and the Conditions and Actions for CC in CTS are the same as those for RHR. Unlike other Westinghouse plants, RHR does not share components with the ECCS, and thus does not play a role in DBA mitigation in MODES 1, 2, 3, and 4. Other plants use CC for DBA mitigation functions other than ECCS in MODES 1, 2, 3, and 4, but the CC system at NAPS does not. This makes the CC System at NAPS different from the CC System described in the NUREG STS, and retaining the CC requirement for MODES 5 and 6 for supporting RHR or any other components not assumed in DBA analysis is inappropriate. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Component Cooling Water Subsystem - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

29.3.7.4.2 SERVICE WATER SYSTEM - Shutdown CTS 3.7.4.2 states that one service water loop shall be OPERABLE when both units are in MODES 5 or 6. The Service Water (SW) System in MODES 5 or 6 is used to provide cooling water to various safety and nonsafety related systems. Its principal safety function is to cool the Recirculation Spray (RS) heat exchangers which are not required to be OPERABLE in MODES 5 or 6. It also provides cooling water to the Component Cooling Water system (which supports no accident loads),

the main control room coolers, instrument air compressors, and charging pump gearbox coolers. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Service Water System - Shutdown LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

30. 3.7.5.1.b ULTIMATE HEAT SINK - North Anna Reservoir CTS 3.7.5.1.b states that one of the ultimate heat sinks that shall be OPERABLE is the North Anna Reservoir with a minimum water level at or above elevation 244 Mean Sea Level, USCG Datum, and average water temperature of - 950 F -as measured at the condenser inlet. The North Anna Reservoir provides makeup to the Service Water Reservoir for 30 days after a DBA as necessary to maintain cooling water inventory, ensuring a continued cooling capability. The Service Water Reservoir is credited as the ultimate heat sink for the DBA. The Service Water Reservoir contains adequate water to provide at least 30 days of cooling to support simultaneous safe shutdown and cooldown of both units and their maintenance in a safe-shutdown condition. The Service Water Reservoir also provides sufficient cooling for at least 30 days in the event of an accident in one unit, to permit control of that accident and permit simultaneous safe shutdown and cooldown of the remaining unit and maintain them in a safe-shutdown condition.

The North Anna Reservoir serves as a backup to the Service Water Reservoir. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Ultimate Heat Sink - North Anna Reservoir LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

31.3.7.6.1 FLOOD PROTECTION CTS 3.7.6.1 states the maximum elevation of the North Anna Reservoir. If this limit is exceeded, flood control measures are required to protect safety related equipment. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Flood Protection LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

32.3.7.9.1 RESIDUAL HEAT REMOVAL SYSTEMS - (RHR) Operating CTS 3.7.9.1 states that two RHR subsystems shall be OPERABLE in MODES 1, 2, and 3. The RHR System is used to remove decay heat from the reactor in MODES 4, 5, and 6. The RHR does not operate in MODES 1, 2 and 3 and must be isolated from the reactor coolant system in those MODES to prevent over pressurization of the RHR components. The RHR System serves no accident mitigation function in any MODE. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the RHR - Shutdcwn Operating LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

33.3.7.10 SNUBBERS CTS 3.7.10 states that snubbers shall be OPERABLE. The OPERABILITY of snubbers ensures that the Reactor Coolant System and other safety related fluid systems are adequately restrained and supported during an earthquake and are free to expand and contract during normal operation as the system temperature changes. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Snubbers LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

34. 3.7.11.1 SEALED SOURCE CONTAMINATION CTS 3.7.11.1 states each sealed source containing radioactive material either in excess of 100 micro curies of beta and/or gamma emitting materials or 5 micro curies of alpha emitting material, shall be free of greater than or equal to 0.005 micro curies of removable contamination. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Sealed Source Contamination LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

35. 3.7.12.1 SETTLEMENT OF CLASS 1 STRUCTURES CTS 3.7.12.1 and Table 3.7-5 provide limits on the total and differential settlement of Class 1 structures. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Settlement of Class 1 Structures LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

36.3.7.13 GROUNDWATER LEVEL - Service Water Reservoir CTS 3.7.13 requires periodic measurement of the groundwater level at locations around the Service Water Reservoir. The groundwater level of the Service Water Reservoir is used to monitor long-term performance of the Service Water Reservoir dike. Failure to meet the requirements of the LCO does not result in the inoperability of the Service Water System. The ACTIONS direct that evaluations be performed to determine cause and consequences of the high groundwater level.

The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Groundwater Level - Service Water Reservoir LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

37. 3.8.2.5 (Unit 2) CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES Unit 2 CTS 3.8.2.5 states the primary and backup containment penetration conductor overcurrent protective devices associated with each containment electrical penetration circuit shall be OPERABLE. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Containment Penetration Conductor Overcurrent Protective Devices LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

38.3.8.2.6 (Unit 2) MOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION DEVICES Unit 2 CTS 3.8.2.6 states the thermal overload protection devices, integral with the motor starter, of each valve in the safety system shall be OPERABLE. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Motor-Operated Valves Thermal Overload Protection Devices LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

39.3.8.2.7 (Unit 2) NORMALLY DE-ENERGIZED POWER CIRCUITS Unit 2 CTS 3.8.2.7 states that all circuits that have containment penetrations and are not required during reactor operations shall be de-energized. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Normally De-Energized Power Circuits LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

40.3.9.3 DECAY TIME CTS 3.9.3 states that the reactor must be subcritical for at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> prior to movement of movement of irradiated fuel in the reactor pressure vessel. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Decay Time LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

41.3.9.5 COMMUNICATIONS CTS 3.9.5 states that direct communications shall be maintained between the control room and personnel at the refueling station during CORE ALTERATIONS.

This ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS. The prompt notification of the control room of a fuel handling accident is an assumption in the Fuel Handling Analysis. This prompt notification is used to ensure that the control room is isolated promptly and is necessary to meet the control room operator dose limits in General Design Criteria 19. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Communications LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

42.3.9.6 MANIPULATOR CRANE OPERABILITY CTS 3.9.6 states that the manipulator crane and auxiliary hoist shall be used for movement of control rods or fuel assemblies and shall be OPERABLE during movement of control rods or fuel assemblies within the reactor pressure vessel.

This specification ensures that the lifting device on the Manipulator Crane has adequate capacity to lift the weight of a fuel assembly and a Rod Control Cluster Assembly, and that an automatic load limiting device is available to prevent damage to the fuel assembly during fuel movement. This specification also ensures that the auxiliary hoist on the Manipulator Crane has adequate capacity for latching and unlatching control rod drive shafts. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Manipulator Crane Operability LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

43.3.9.7 CRANE TRAVEL - SPENT FUEL PIT CTS 3.9.7 places restriction on movement of loads over irradiated assemblies in the spent fuel pit in excess of 2500 pounds. This represents the working load of the fuel assembly plus gripper. The LCO ensures that in the event this load is dropped the activity release will be limited to that contained in a single fuel assembly and any possible distortion of fuel in the storage racks will not result in a critical array. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Crane Travel - Spent Fuel Pit LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

44.3.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM CTS 3.9.9 states requirements for the containment purge and exhaust isolation system, which automatically closes the containment purge and exhaust isolation valves in MODE 6. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Containment Purge and Exhaust System LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

45. 3.9.10.2 WATER LEVEL - Reactor Vessel Control Rods CTS 3.9.10.2 states that the refueling cavity water level must be at least 23 feet above the fuel during MODE 6 during movement of control rods within the reactor pressure vessel. Movement of control rods is not an initiator of any UFSAR accident analysis. The staff has determined that the screening criteria of 10 CFR 50.36 have not been satisfied, and thus the Water Level - Reactor Vessel - Control Rods LCO and Surveillances may be relocated to other plant controlled documents outside the ITS.

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 UFSAR, TRM, ODCM, Rl-,

or ISI Program. 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 licensee's 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 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 UFSAR and TRM can only be revised in accordance with the provisions of 10 CFR 50.59, which ensures records are maintained and establishes 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 OGQM-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. The rcFlc3ticns to theUFSAR hih include the TRM, shall be incl nthe net required update of the UFSAR in accordance with 10 CFIR 50.71 (e-)-.

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 licensee's conversion application.

These include items which 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 (BSI), which have been identified by the licensee in their submittal, and by the staff during the course of the staff review. These BSIs are included in the notice of consideration of amendment published in the Federal Register on xx, xx, 2002.

G.1 BSI Changes identified by the Licensee:

The changes discussed below are licensee-identified BSI 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 licensee's application.

1.

ITS SR 3.3.1.6, (DOC L.16 and JFD 15)

ITS states "Compare results of the excore channels to incore detector measurements." Note is added to require NIS channel adjustment if absolute difference is > 3%.

The licensee responded to the staff's RAI with a letter, dated November 8, 2001. In this letter, the licensee proposed to use a TSTF presently under NEI review. Subsequently, during a conference call on December 2, 2001, the licensee proposed to replace this BSI with a submittal to extend the SR from 92 days to 6 months. As of January 18, 2002 the staff has not received this submittal.

2.

ITS 3.3.1 Function 6 OTDT Allowable Value Note 1, DOC L.21, JFD 7, CTS Table 2.2-1 Function 7 OTDT Allowable Value Note 3 (DOC L.21 and JFD 7)

ITS states the % allowed for the trip setpoints may differ from the Allowable Value by 2.3%

The licensee in a letter, dated January 2, 2002, withdrew this BSI.

3.

ITS 3.3.2 ESFAS INTERLOCK P DOC M.7, JFD 1, CTS Table 3.3-3 ESF Interlock P-12 (DOC M.7 and JFD 1)

ITS states the Allowable Value for the P-12 interlock as 542 degrees.

The staff received the licensee's submittal, dated October 17, 2001. This BSI is under staff review.

4. ITS 3.3.2 ESFAS Functions 1.c, 1.d, M.f, 2.c, 4.c, and 4.d, DOC M.7, JFD 1, CTS Table 3.3-4 ESF Functions 1.c, 1.

1.f, 2.c, 4.c, and 4.d (DOC M.7and JFD 1)

NUREG brackets the Allowable Values for the fe4owipg-functions.

In a letter, dated December 13, 2001, the licensee requested staff approval of plant specific methodology for NAPS and Surry Power Station. This BSI is under staff review

5.

ITS 3.4.12, Low Temperature Overpressure Protection (LTOP) System, Condition C (DOC M.4 and JFD 6):

ITS states for Condition C that when an accumulator is not isolated or power is available to one or more accumulator isolation valve operators, the accumulator must be isolated immediately and power removed from affected accumulator isolation valve operator in one hour. Note modifies the Condition to state that it is only applicable when accumulator pressure is greater than PORV lift setpoints.

This beyond scope issue is related to NUREG-1431, STS 3.4.12, "Low Temperature Overpressure Protection (LTOP) System" regarding the accumulator isolation requirement. The licensee proposed ITS 3.4.12, will,: 1) add a note to ACTION C which indicates that the accumulator isolation is only applicable when accumulator pressure is greater than power operated relief valve (PORV) setting7-; 2) add REQUIRED ACTION C.2 to state toat-"Remove power from affected accumulator isolation valve operators;7" and 3) add a note in LCO section which states that, "Accumulator isolation with power removed from the isolation valve operators is only required when accumulator pressure is greater than the PORV lift setting." STS 3.4.12 of NUREG-1431 has:-: 1) a note in the APPLICABILITY section which states that "accumulator isolation is only required when accumulator pressure is greater than or equal to the maximum RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided in the PTLR,'-';" and 2) ACTION C contains thýe-similar words as the note in the APPLICABILITY section.

The primary purpose of the accumulator isolation during LTOP conditions is to prevent inadvertent injection of water from the accumulators into RCS which may be a challenge to reactor vessel P/T limits during low temperature operating conditions. The PORVs at NAPS Units are served as an LTOP system with their setpoints designed to protect reactor vessel P/T limits under the limiting mass addition and heat addition transients. The settings of these PORVs are lower than the P/T limits in various temperature regions.

The proposed ITS 3.4.12, in the areas of accumulator isolation, will require that when the plant is operating in the LTOP conditions, and the accumulator pressures are above the PORV settings, the accumulators are required to be isolated with power removed from the isolation valve operators. The staff evaluated the licensee's submittals and concludes that the licensee proposed ITS 3.4.12 regarding accumulator isolation is acceptable. The bases for the staff acceptance are _-:1) The proposed ITS will only allow the accumulators connected to RCS when the accumulator pressures are lower than the PORV settings. Since the PORVs are designed to mitigate the limiting mass addition from a charging pump, IW-it is unlikely that the P/T limits will be challenged by water injecting to RCS from the accumulators;7 2)The proposed ITS is more conservative than STS 3.4.12 in NUREG 1431 since the STS would allow the accumulators to be connected to RCS when the accumulator pressures are below the P/T limits but above the PORV settings;7 3) The proposed ITS will require power removal from the isolation valve operators for added assurance for accumulator isolation;7 and 4) The proposed ITS add plant operational restrictions to NAPS current licensing bases regarding the requirement of accumulation isolation.

There is no such requirement in their CTS.

Based on the above review, the staff finds that the licensee proposed ITS 3.4.12 in the areas of requiring accumulator isolation during LTOP conditions are more conservative than that in CTS and STS 3.4.12 of NUREG-1431.

Therefore, the proposed ITS 3.4.12 regarding accumulator isolation is acceptable.

6. ITS 3.7.3 - Main Feedwater Isolation Valves (MFIVs), Main Feedwater Pump Discharqge Valves (MFPDVs), Main Feedwater Regulating Valves (MFRVs), and Main Feedwater Regulating Bypass Valves (MFRBVs) (DOC M1 and JFD 3);

The licensee proposed the adoption of Section 3.7.3 of the STS. Adoption of Section 3.7.3 presents several deviations to the standard format provided in NUREG-1431 and has therefore been identified as a "beyond scope" issue. The title of this section as adopted into the NAPS Units 1 and 2 ITS iseieeITS 3.7.3, "Main Feedwater Isolation Valves (MFIVs), Main Feedwater Pump Discharge Valves (MFPDVs), Main Feedwater Regulating Valves (MFRVs), and Main Feedwater Regulating Bypass Valves (MFRBVs)."

The NAPS feedwater system consists of three main feedwater pumps with associated Main Feedwater Pump Discharge Valves that feed a common header. From this header are three lines feeding the three steam generators. On each line is a Main Feedwater Isolation Valve in series with a Main Feedwater Regulating Valve. On a line which bypasses each MFIV and MFRV is a Main Feedwater Regulating Bypass Valve. Each of these valves, the MFPDVs, MFIVs, MFRVs, and MFRBVs, close on receipt of a Safety Injection or Steam Generator Water Level High-High Signal. The MFIVs and the MFRVs provide single failure protection for each other.

The MFPDVs and the MFRBVs provide single failure protection for each other. Therefore, all four valve types are required to meet the safety analysis assumptions.

The mosGt significant deviation in content that ITS 3.7.3 presents is that ITS 3.7.3 Requi, Actions A.2, B.2, 0.2, and added Required Action ID.2, are revised to state, "Verify by adMinictrative mneans [MFIV or MF=PDV or MFRV or MFRBVI is closed Or isolated." The phFaeI specified valves are elosed aRd isolated, thee* is no indication available in the Control Room ot the valve position. Therefere, this verification must be peIoRmed by plant pIrse1Iel access*Rg the area where the valve is and Yerifying it's position visually. The licensee indicated that this administrative action will be pecformed according to NAPS in house procedure. The staff accepts this deviation from the STS. (DELETED - RAI 3.7.3 - 1 LETTER DATED 12/3/01, Serial Number 01-645)

The most significant deviation in format to the STS is that the NAPS ITS 3.7.3 will include Main Feedwater Pump Discharge Valves. The STS 3.7.3 (as written in NUREG-1431) addresses Main Feedwater Isolation Valves and Main Feedwater Regulating Valves and associated bypass valves but not MFPDVs. Because NAPS's Main Feedwater System includes Main Feedwater Pump Discharge Valves, and because they provide single failure protection for the MFRBVs (and therefore are required to meet safety analysis assumptions), it is appropriate that the MFPDVs be included in ITS 3.7.3.

Other changes being made are the inclusion of NAPS plant specific values and information, where appropriate, in place of those presented in Section 3.7.3 of the STS. An example of this is the isolation time for the MFIVs, MFRVs, and MFPDVs. The time presented in ITS surveillance requirement (SR) 3.7.3.1 was changed to represent the NAPS requirement and differs slightly from the isolation time presented in the STS SR 3.7.3.1. The ITS SR 3.7.3.1 also adds the requirement to test the closure time of each MFPDV.

Based on our review, the staff finds the proposed change to adopt STS 3.7.3 for NAPS to be acceptable with the deviations from the STS cited above.

7. ITS 3.7.7, DOC R.1, JFD 1, CTS 3.7.3.1 and 3.7.3.42 (DOC R.1 and JFD 1)

ITS does not include an LCO for the Component Cooling System.

In response to the staffs RAI, the licensee provided a submittal dated November 19, 2001.

During a conference call conducted on January 3, 2002, the staff requested additional information, which the licensee has agreed to submit by the end of January 2002.

8.

ITS 3.7.9, DC R.--, CTS 3.7.6.1 (DOC R.1)

NUREG includes requirements for the Ultimate Heat Sink. ITS does not include requirements for the NAPS Reservoir.

In response to the staffs RAI, the licensee provided a submittal dated November 19, 2001.

During a conference call conducted on January 3, 2002, the staff requested additional information, which the licensee has agreed to submit by the end of January 2002.

9.

ITS SR 3.7.11.1 - Main Control Room/Emerqency Switchgear Room Air Conditioninq System(DOC M.2 and JFD 4)

The licensee proposed changing the surveillance requirement frequency of SR 3.7.11.1, from "18 months" to "18 months on a Staggered Test Basis."

An air conditioning system with two independent 100% capacity trains for each unit which supplies the relay rooms and common control room are designed for 75 OF dry bulb at approximately 50% relative humidity during normal operation. For emergency conditions, there is sufficient cooling capacity to maintain the control room, computer room, and relay room space temperature well below the design maximum of 120 OF. A third chiller is provided for each reactor unit as an alternative for either train. One 100% capacity cooling system which supplies the relay rooms and common control room in order to meet the signal failure criterion is installed for each reactor unit. The cooling systems cannot be cross connected between the two reactor units. Only one train for each unit is used at a time.

The emergency ACS for the MCR/ESGR envelope consists of two independent 100% redundant subsystems, one chiller in one subsystem and two chillers in the other. Each subsystem consists of two air handling units, one for the MCR and one for the ESGR to provide the heat removal function during post accident conditions as well as during normal operation. The licensee added Staggered Test Basis with the 18 months surveillance test frequency of chillers. The staff finds the proposed change acceptable, because there are three chillers with 100% cooling operation capability, either of which can be used by the subsystem and in staffs judgement, changing the surveillance frequency to every 18 month on a Staggered Test Basis provides an acceptable level of confidence that the system will function as assumed in the accident analysis.

10. ITS 3.7.12 LCO Note, DOC M.2, JFD 14, CTS 3.7.8.1(DOC M.2 and JFD 4)

The licensee proposed to add the phrase "not open by design" to ITS 3.7.12 LCO to convey that the ECCS pump room boundary openings not open by design may be opened. This additionaI-is deviated deviates from the NUREG NOTE, which states that the ECCS pump room boundary openings may be opened intermittently under administrative control.

This item is under review by the staff.

11. ITS Surveillance Requirements (SR) 3.7.12.2 and 3.7.12.4, CTS 4.7.8.1.a.1 - Emer~qency Core Coolinq System (ECCS) Pump Room Exhaust Air Cleanup System (PREACS)(DOC M.1 and JFD 7)

The licensee proposed adding the following surveillance requirement as ITS SR 3.7.12.2, with a surveillance frequency of 31 days: "Actuate each ECCS PREACS train by aligning Safeguards Area exhaust flow and Auxiliary Building Central exhaust flow through the Auxiliary Building HEPA filter and charcoal adsorber assembly".

The Emergency Core Cooling System (ECCS) Pump Room Exhaust Cleanup System (PREACS) filters air from the area of the active ECCS components during the recirculation phase of a loss of coolant accident (LOCA). The ECCS PREACS, in conjunction with other normally operating systems, also provides environmental control of temperature in the ECCS pump room areas.

The licensee stated that this surveillance requirement, ITS SR 3.7.12.2, is added to divert Safeguards Area exhaust flow and Auxiliary Building Central exhaust system flow through the Auxiliary Building HEPA filter and charcoal adsorber assembly for the operating Safeguards Area fan, from the control room, every 31 days. This ITS SR 3.7.12.2 requires certain dampers associated with the Auxiliary Building Central exhaust system to be manually actuated, and tested. This provides additional assurance that the exhaust flow can be diverted through the filters in case of a Design Basis Accident (DBA) that requiresiR§ their actuation. The licensee also stated that the 31 days test frequency is based on the known reliability of the equipment and the availability of redundant trains.

This new SR is added to ensure that in the event of a postulated DBA, the ECCS PREACS train is operable to reduce the potential dose risk from a radiological event. The staff concludes that the proposed SR is a conservative addition and therefore finds it acceptable. With this proposed change, the STS surveillance requirement 3.7.12.2 is then renumbered to become ITS SR 3.7.12.3. This an administrative change and the staff finds it acceptable. Similarly, STS SR 3.7.12.3 is renumbered to become ITS SR 3.7.12.4. This is also an administrative change which the staff finds acceptable. In addition, the.I....ee proposed changing this SR from "Verify each EGGS PREACS train actuates On an actual or simulated actuation signal" to "Verify Safeguards Area exhaust flow is diverted and eacsh Auxiliary Building filter bank is actuated On an actual Or "simulated actuation signal". (REDUNDANT)

In addition, STS SR 3.7.12.3 requires verifying each ECCS PREACS train to actuate on an actual or simulated actuation signal. The licensee proposed a change to this SR by replacing "Verify each ECCS PREACS train actuates on an actual or simulated actuation signal" with "Verify Safeguards Area exhaust flow is diverted and each Auxiliary Building filter bank is actuated on an actual or simulated actuation signal" on a surveillance frequency of every 18 months. The staff finds this change acceptable because this SR ve4fyf-verifies proper operation of actuation signal and assures that the each Auxiliary Building filter bank signal will actuate in case of an accident.

12. ITS 3.7.13, DOG M.2. JFD) 7.

T--T 3.7.7.1 Action b. (DOC M.2 and JFD 7)

CTS allows the bottled air system to be inoperable for seven days. ITS allows two or more required trains of the MCR/ESGR bottled air system to be inoperable for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This item is under review by the staff.

13. ITS 3.7.15, DOC L.2, JFD 5. CTS 3.9.12 (DOC L.2 and JFD 5)

Fuel Building Ventilation System (FBVS) - CTS SR 4.9.12.

The Fuel Building Ventilation System (FBVS) consists of dual exhaust fans and two-speed supply fans. One supply fan serves the spent fuel pit area and one feF-serves the remote equipment space at Evaluation Elevation 249 ft. 4 in. Both take suction from a common plenum fitted with a combination roll and high efficiency filter (95% atmospheric dust spot efficiency) and steam coils for air tempering and space heating. The exhaust fans discharge through vent stack B and are arranged for selective alignment through the auxiliary building HEPA/charcoal filter bank. The area of the remote equipment room subject to radioactive contamination is exhausted by a branch from the decontamination building exhaust system.

The licensee proposed to eliminate the testing requirement for the fuel building filtration system from the ITS by deleting CTS SR 4.9.12 (a) and CTS SR 4.9.12 (c). The purpose of these SRs is to verify that the fuel building filters can perform as required. In the submittal, the licensee states that the deleted SRs are not necessary to verify that the equipment used to meet the LCO are consistent with the safety analysis and can perform its required functions. Thus, appropriate equipment continues to be tested in a manner and at a frequency necessary to give confidence that the equipment can perform its assumed safety function. Furthermore, the licensee stated that the deletion of the requirement for the FBVS filters is acceptable because the NAPS radiological analysis of the fuel handling accident (FHA) in the fuel building assumes that all of the radionuclides released from the fuel pool are released without credit for filtration of the released material.

In order to determine the acceptability of the deletion of requirements for the FBVS filters, the staff examined the licensee's design basis radiological analysis of the FHA as documented in the licensee's UFSAR, Chapter 15.4.5. The previous licensee analysis along with the resulting dose consequences were found to be acceptable by the staff. The staff verified that the current fuel building FHA radiological analysis does not take credit for filtration of the released material and that the analysis assumptions as listed in the UFSAR are consistent with Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors."

The dose consequences of the FHA were previously found by the staff to be well within the dose guidelines given in 10 CFR Part 100 for offsite doses and also meet dose criteria in 10 CFR 50, Appendix A, General Design Criterion (GDC) 19 for the control room. The staff finds the proposed changes to the NAPS TS that remove requirements for testing the FBVS filtration capability are consistent with assumptions used in the current design basis analysis found in the NAPS UFSAR.

The licensee proposes, in accordance with TSTF-51, to add the term recently irradiated fuel as fuel that has been part of a critical reactor core within a licensee-specified number of days. The proposed TS bases state that until analyses are performed to determine a specific value, recently irradiated fuel is defined as any irradiated fuel. CTS 3.9.3 "Decay Time" is being relocated to the TRM. The required decay time is 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> before allowing movement of irradiated fuel, which is longer than the assumed decay time of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> in the UFSAR FHA radiological analysis. The staff finds that the licensee's proposed definition of recently irradiated fuel is consistent with the NAPS design basis analysis. Based on the above evaluation, the staff concludes that the proposed changes to SR 4.9.12 incorporated into the ITS are acceptable.

14. ITS 3.9.4, LAj. A.5., JF 2, CTS 3.9.4 (DOC LA.1, A.5 and JFD 2)

The licensee in their submittal dated November 8, 2001, withdrew their request to use this BSI.

15. ITS Table 5.5.8 Steam Generator Tube Inspection (DOC L.22 3nd JFD 1). CTS Table 4.4-2. Unit 1 CTS steam qenerator tube inspection requirements(DOC L.22 and JFD 1)

The licensee proposed to delete the requirements in the current technical specifications Section 4.4.5, Table 4.4-2, "Report to NRC and Obtain Approval Prior to Operation," in the event an additional steam generator is found to be in category C-3. The licensee stated that the requirement is not specified in the STS. The proposed deletion makes this table consistent with the corresponding table in the STS.

The proposed administrative TS retains the requirement to notify the NRC if inspection results fall into category C-3. This notification is to be made pursuant to 10 CFR 50.72, and the "approval" requirement was imposed on the licensee prior to replacement of steam generators when tube leaks at times during operation were frequent. However, the licensee has since Fepla.Ged installed wih-new steam generators and the steam generator performance is significantly improved, and thus the staff concurs that an "approval" requirement is no longer necessary. This deletion will make the NAPS technical specifications consistent with the STS in NUREG 1431.

The proposed change is not expected to have any affect on safety, and therefore, the staff finds that the proposed change acceptable.

G.2 Additional BSI Changes identified by the Staff:

1.

ITS 3.3.1, (JFD 14, DOC A.24)

In April of 2001, Westinghouse published NUREG-1431, Rev 2, "NUREG 1431, Rev 2, "Standard Technical Specifications, Westinghouse Plants." Many of the Westinghouse designed plants including North Anna Units 1 and 2 are converting to the ITS to provide consistency in their technical specifications and reduce regulatory burden. The staff is responsible for reviewing the conversion of each plant to the ITS format from their CTS. The staff must ensure all safety and regulatory requirements are met.

North Anna Units 1 and 2 used the Westinghouse ITS and WCAP-14483-A "Generic Methodology for Expanded Core Operating Limits Report" to develop their ITS and new Core Operating Limits Report (COLR). The COLR allows licensees to change cycle-specific technical values without NRC approval, provided that NRC approved methodologies are used to determine the values. The staff reviews the implementation of a COLR to ensure that the proper approved methodologies are being used.

Due to the differences between the format and content of the CTS and the proposed ITS for North Anna Units 1 and 2, the staff must review any changes involved in the conversion. This safety evaluation discusses the review of the following two BSIs for North Anna Units 1 and 2:

1) The overtemperature AT and overpower AT formulas contained in Notes 1 and 2 of ITS Table 3.3.1-1 have been modified in the proposed ITS to reflect those used as the licensing basis in the North Anna CTS.

2) The licensee stated that these changes reflect the plant specific CTS formulas in the proposed ITS requirements. The licensee proposed to exclude the statement "with gains to be selected based on measured instrument response during plant startup tests such that:" in Table 2.2-1, Note 1 of the CTS, from the proposed ITS. This statement describes the methodology used to determine the gains used in the calculation of the overtemperature AT trip setpoints. The licensee's justification for deletion contends that this statement is for information only and since the gains have not been adjusted without engineering evaluation and NRC approval since their initial calculation, the removal is administrative.

With regard to the first BSI, the staff reviewed the formulas for the overtemperature and overpower AT functions in Notes 1 and 2 of the ITS Table 3.3.1-1, and found that they are identical to those in Notes 1 and 2, respectively, of the CTS Table 2.2-1. Since these formulas were previously approved by the NRC as the licensing basis in the North Anna CTS and have not been changed in the conversion to the ITS, the staff finds their use in the ITS acceptable.

In evaluating the second BSI, the staff reviewed the methodologies used by the licensee to calculate the allowable overtemperature AT gains and trip setpoints. The staff conducted this review to determine if it was acceptable for the licensee to exclude the statement "with gains to be selected based upon measured instrument response during plant startup tests such that:"

from Note 1 of ITS Table 3.3.1-1. This statement appears in CTS Note 1 of Table 2.2-1 and describes how gains for the axial flux difference are determined and used in the calculation of the overtemperature AT trip setpoints. In two separate RAIs dated September 7 and November 7, 2001, the staff requested the license provide detailed information on the procedures and methodologies used to determine the allowable values for the gains and setpoints. The licensee provided responses dated October 10 and December 12, 2001, which indicate the procedures and NRC approved methodologies used in determining the appropriate gains and trip setpoints. The licensee stated that they used the NRC approved methodology contained in WCAP-8-7488745-P-A, "Design Bases for the Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Functions." The staff has approved this topical report for calculation of the constants used in the overtemperature and overpower AT formulas. Since the licensee is using NRC approved methodologies used for the calculation of the allowable overtemperature AT gains and trip setpoints, the staff finds it acceptable to exclude the identified statement from ITS Table 3.3.1-1, Note 1.

In reviewing the December 12, 2001, RAI response, the staff noted licensee statements to conditionally adopt WCAP-14483-A, "Generic Methodology for Expanded Core Operating Limits Report" (COLR), to allow relocation of overtemperature and overpower AT allowable values to the COLR. The staff reviewed the response to determine if an NRC approved methodology was used in calculating the allowable values and gains for the purpose of acceptability to remove the statement on how gains are determined from the ITS. This safety evaluation has not reviewed the response to determine acceptability of relocating values to the COLR for North Anna, Units 1 and 2 because no clear position on licensee use of WCAP 14483-A was established.

The staff reviewed two BSIs related to the licensee's conversion from the CTS to the Westinghouse ITS. First, the staff approves the use of the plant specific ITS equations for the overtemperature and overpower AT equations shown in Table 3.3.1-1, Notes 1 and 2. The staff has concluded that these equations are identical to those previously approved in CTS Table 2.2-1, Notes 1 and 2. Secondly, the staff approves the exclusion of the statement "with gains to be selected based upon measure instrument response during plant startup tests such that:" from ITS Table 3.3.1-1, Note 1. The staff concluded that the licensee used NRC approved methodologies to calculate the allowable overtemperature AT gains and trip setpoints.

2.

ITS 3.3.1, DCG-L L.-Unit2(DOC L.7)

Category 6 - Relaxation of Surveillance Requirement Acceptance Criteria.

The licensee will adopt TSTF-371, which is under staff review.

3.

ITS 3.3.1 - Reactor Trip System (RTS) Instrumentation: fDOCG-t--.%-,,Relaxation of LCO Requirements, Allowable Values for the P-7 function come from the requirements of P-10 and P-13(DOC L.8)

The licensee proposed a change to the allowable values of the setpoints for the P-7 interlock (Low Power Reactor Trips Block) to a value not currently allowed by their current TS. The original allowable value for P-7 was <11 percent. The staff reviewed the proposed change and finds a change to the CTS which lists the allowable value as NA (Not Applicable). However, the P-7 interlock uses the P-1 0 and P-1 3 interlocks for inputs. The licensee proposed new allowable value for P-1 0 and P-1 3 of < 11 percent. This change effectively modifies the P-7 actuation from

<11 percent to *11 percent, thus including 11 percent as an allowable value. The staff considers this change to be less restrictive, however it is considered to have a negligible effect. Based on this review, the staff finds the proposed change acceptable.

4. ITS 3.3.1 - Reactor Trip System (RTS) Instrumentation: DOC-L.A"4)i Relaxation of LCO Requirements, Allowable Value Chanqes for P-6, P-8, and P-13 interlocks(DOC L.14)

The licensee proposed changes to the allowable values for the P-6, P-8, and P-13 interlocks.

The P-6 interlock function for increasing power (intermediate range above setpoints) is to allow the operators to manually block the Source Range channels trip capability. Securing the Source Range channels trip is not a safety function, but is an equipment protection function. The licensee proposed removing this P-6 setting from the improved TS. The staff reviewed the change and finds this removal acceptable. However, the P-6 interlock function while decreasing power (intermediate range below setpoints) is safety related. This interlock activates the Source Range channels trip capability. The allowable value for the decreasing power P-6 interlock is listed as >3x1 0-' in the current Technical Specifications. The proposed allowable value is listed as >3xi01 0. This change is less restrictive, but is considered to have a negligible effect. Based on this review, the staff finds the proposed change acceptable.

When below the defined setpoints, the P-8 interlock prevents a reactor trip for the following conditions: low flow in a single loop, a single reactor coolant pump breaker open, or a turbine trip.

This function (power range below setpoints) is not a safety function and the associated setpoints have been removed from the proposed TS. The staff finds this removal acceptable. However, when above ITS setpoints, the P-8 interlock allows a reactor trip for the above conditions. The current TS list the allowable value for the setpoints as <31 percent on the power range channels.

The licensee proposed changing the allowable value to *31 percent. This change is less restrictive, but is considered to have a negligible effect. Based on this review, the staff finds the proposed change acceptable.

The P-1 3 interlock (Turbine Impulse Pressure) is an input to the P-7 interlock. When above the setpoints, P-13 (in conjunction with P-10) allows a reactor trip under the following conditions in more than one loop: low flow, reactor coolant pump breaker open, under voltage on the reactor coolant pump busses, and under frequency on the reactor coolant pump buses. P-1 3 also allows a reactor trip on pressurizer low pressure or pressurizer high level when above the setpoints.

The current TS list the allowable value as < 11 percent. The licensee proposed changing the allowable value to *11 percent. The inclusion of 11 percent is less restrictive, but it is considered negligible. Based on this review, the staff finds the proposed change acceptable.

When below the setpoints, P-1 3 (in conjunction with P-1 0) prevents a reactor trip when any of the following conditions occur: reactor coolant system low flow, reactor coolant pump breakers open, reactor coolant pump busses under voltage, reactor coolant pump busses under frequency, pressurizer low pressure, and pressurizer high level. This function of P-13 is not assumed in the safety analyses. Therefore, the licensee proposed removing the setpoints and allowable values for this function of P-13 from their TS. Based on this review, the staff finds this removal acceptable.

85. ITS Table 3.3.2-1, ESFAS Instrumentation Function 7. Automatic Switch over to Containment Sump( DOC M.3)

The following proposed ITS changes have been determined to be beyond the scope of the conversion to the STS format for North Anna Power Station Units 1 and 2:

Functional Unit 7, "Automatic Switchover to Containment Sump" is being included in Technical Specifications 3/4.3.2, "Engineered Safety Feature Actuation System (ESFAS)

Instrumentation," with allowed outage time and additional channel bypass.

In addition, as revised by the licensee in their May 30, 2001 submittal, the proposed ITS 3.3.2 limiting condition for operation section in the Bases section is as follows:

3.3.2 - Action I, RWST Level-Low Low Coincident with Safety Injection (Bases pages 3.3.2-38 and 3.3.2-39)

The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is justified in a plant-specific risk assessment, consistent with Reference 8.

The total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to reach Mode 3 and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for a second channel to be bypassed is acceptable based on the results of a plant-specific risk assessment, consistent with Reference 8.

A plant-specific risk assessment was completed to include an assessment of Functional Unit 7, "Automatic Switchover to Containment Sump." Functional Unit 7 had been included as a new unit in the Improved Technical Specifications for consistency with NUREG-1431. The plant specific risk evaluation assessed the change in core damage frequency (CDF) and the incremental conditional core damage probability (ICCDP) as a result of the WCAP changes for the additional functions.

The licensee developed the CDF sensitivity for this function in the same manner as the WCAP-10271 and WCAP-14333 analyses. The automatic containment switchover function is similar to that of some of the WCAP channels and was estimated by comparison to similar functions. Once the channel failure impacts were quantified, these numbers were converted to a CDF impact by looking at the associated CDF sensitivity from the North Anna probable risk assessment model for the same function or a higher level function.

The automatic switchover to containment sump occurs when the refueling water storage tank level drops to the established setpoint. Automatic switchover failure probability is estimated to increase by approximately 1.3E-4 as result of the proposed changes. This increase is based upon the assumption that the full allowed outage time will be used on a regular basis every year.

The result of a plant-specific risk assessment for this function related to CDF impact is negligible (less than 0.01% of the CDF) based on the baseline CDF (3.3E-5/yr) at North Anna.

This risk assessment demonstrates that the effect on CDF and ICCDP is negligible for the potential unavailability changes associated with this function. The staff concludes that the licensee's proposed Functional Unit 7 Technical Specification allowed outage and bypass times are acceptable.

96. ITS 3.7.11 Actions D and E, DOC M.1 3nd M.3, JFD 3, CTS 3/4.7.7.1 Action d (DOC M.1 and M.3)

The ITS proposes to only require entry into Action D, for one AC subsystem inoperable, as long as 100% air conditioning system cooling equivalent to a single operable AC subsystem is available.

The emergency ACS for the MCR/ESGR envelope consists of two independent 100% redundant subsystems, one chiller in one subsystem and two chillers in the other. Each subsystem consists of two air handling units, one for the MCR and one for the ESGR to provide the heat removal function during post accident conditions as well as during normal operation. An air conditioning system with two independent 100% capacity subsystems for each unit which supplies the relay rooms and common control room are designed for 750 F dry bulb at approximately 50% relative humidity during normal operation. For emergency conditions, there is sufficient cooling capacity to maintain the control room, computer room, and relay room space temperature well below the design maximum of 1200 F. A third chiller is provided for each reactor unit as an alternative for either train. The cooling systems cannot be cross connected between the two reactor units. Only one train for each unit is used at a time.

The licensee stated that because the MCR/ESGR ACS includes a total of three chillers and flexibility in the use of system components, the description of system requirements, "Less than 100% of the MCR/ESGR ACS cooling equivalent to a single OPERABLE MCR/ESGR ACS subsystem available...." is proposed in the above ITS instead of a reference to two inoperable trains. The proposed ITS Conditions allow a variety of system configurations to be established that would provide sufficient cooling capacity to meet the design function and allows appropriate flexibility in operation of the system similar to ITS 3.5.2, ECCS. The licensee further stated that the Conditions D and E still require that when the design function can not be met, that the appropriate Applicability (MODES 1,2, 3, and 4 and During movement of recently irradiated fuel assemblies) be exited.

The staff has reviewed the proposed change and agrees with the licensee that the proposed ITS change is consistent with the intent of STS 3.7.11. Since there are three chillers with 100%

cooling capability, eitheF-one of which can be used by either subsystem. The staff finds the proposed ITS change acceptable because it provides the system flexibility in operation of the system, enables the various configurations to maintain the required cooling function, and provides an acceptable level of confidence that the system will function as assumed in the accident analysis.

Based on the above evaluation, the staff concludes that the proposed changes to TS 3.7.11, Actions D and E, incorporated in the ITS are acceptable.

4.0 COMMITMENTS RELIED UPON In reviewing the proposed ITS conversion for NAPS, the staff has relied upon the licensee's commitment to relocate certain requirements from the CTS to licensee-controlled documents as described in Table R, "Relocated Specifications and Removal of Details" (Attachment 5 to this SE). This table reflects the relocations described in the licensee's 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 Appendix*- efthe Operating License. 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 performancesubject 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 s,Api-2 2, 2will be within 5 months after receipt of an approved Safety Evaluation.. 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 R (Attachment 5 to this SE), and Section 3.0.D, "Removed Details," and Section 3.0.E, "Relocated Specifications,"

above. The license condition states that the relocations would be completed no later than December 31, 2001 the implementation date of ITS. This schedule is acceptable.

As a pat ef the ITS convversion, the oicvnsee also pFrposed to delote v

v evisting licen*v condeitios related to compliancoe with CTS8 repo~ting and record retention requirements. These Me conditions, 3.G and &D, are no longer necessary because they duplicate regulationVs regarding repovting and record keoping. They also duplicatv Licence Condition 3.1, "Technical Specifications," Which requires that NAPS opcrate the facility in; accor~dance with the TS. Many 9f the CTS requirements that these tio conditione refcr to arc being relocated

• ut of the ITS to iaensee controlled documents, as specified in the converSion submital and supplements theretol.

Therefore, deletion of these two license conditionS Will have no imnpact 9R the repe~ting and record keepi requirents for NAPS, and is acmeptable.

6.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Virginia State official was notified oe-of the proposed issuance of the ITS conversion amendment for NAPS. 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 xxxxx (xx FR xxxxx), for the proposed conversion of the CTS to ITS for NAPS. Accordingly, based upon the environmental assessment, the Commission has determined that issuance of these amendments will not have a significant effect on the quality of the human environment.

With respect to other changes included in the application for conversion to ITS, the items change requirements with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The staff has determined that the amendments required by these other changes involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission issued proposed findings that the amendments required by these other changes involve no significant hazards consideration, and there has been no public comment on these findings published in the Federal Register on XXXXX (XX FR XXXXX); XXXXX (XX FR XXXXX), and XXXXXX (XX FR XXXXX). Accordingly, these changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the implementation of these changes.

8.0 CONCLUSION

The NAPS ITS provides clearer, more readily understandable requirements to ensure safe operation of the plant. The staff concludes that the ITS for NAPS satisfy the guidance in the Final Policy Statement on TS improvements for nuclear power reactors with regard to the content of TS, and conform to the STS provided in NUREG-1431, Revision 1, with appropriate modifications for plant-specific considerations. The staff further concludes that the ITS satisfy Section 182a of the Atomic Energy Act, 10 CFR 50.36, and other applicable standards. On this basis, the staff concludes that the proposed ITS for NAPS are acceptable.

The staff has also reviewed the plant-specific changes to the CTS as described in this SE. On the basis of the evaluations described herein for each of the changes, the staff also concludes that these changes are acceptable.

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 R - Relocated Specifications and Removed Details Principal Contributors:N. Le S. Monarque C. Schulten B. Fu S. Peters J. Golla CY Liang A. Chu C. Harbuck R. Giardina R. Tjader S. Rhow T. Attard M. Hart R. Taylor J. Raval K. Kavanagh N. lqbal Date:

LIST OF ACRONYMS AC ADD JF-DAFD AA-NWAFW AOT Air Conditioning or Alternating Current Atmospheric Dump Valve Axial Flux Difference Auxiliary Feedwater System Allowed Outage Time AR,.DT A ii;,-,. Dr

• fe m,-e T -,,,ti, ire IE American Society of Mechanical Engineers

-M American Society for Testing and Materials VS Anticipated Transient Without Scram Component Cooling Water Circulating Water Code of Federal Regulations

• CFT Channel Functional Test

-CIV Containment Isolation Valve

_R Core Operating Limits Report CtRi.*mc*nt Prccu-c GoRdc.*-

"te Is,1-n;",

D-.

Control Rod Drive GHEFSCREFSControl Room Emergency Filtration System IR(*S Gntrnl Reem \\VnfilatI*bOR System C;RAW

'--4 r_ ST rndtcnqatP Ntnmnoc I;nk CTS CVCS DBA DG DNB DOC ECCS ECST EDGE EFPD ESPYEFPY Current Technical Specification Chemical and Volume Control System Design-Basis Accident Diesel Generator Departure from Nucleate Boiling Discussion of Change (from the CTS)

Emergency Core Cooling System Emergency Condensate Storage Tank Emergency Diesel Generator Effective Full Power Year Effective Full Power Year Elcctrical Protcction Assembly Engineered Safety Features Actuation System Federal Register Inservice Inspection Inservice Testing Improved Technical Specification Justification for Deviation Kilovolt Kilowatt Limiting Condition for Operation Loss-of-Coolant Accident Loss of Offsite Power Loss of Power ATTACHMENT 1 ASK AST ATV CC CW CFR COL CR EP-A ESFAS FR ISI IST ITS JFD kV kW LCO LOCA LOOP LOP 4

1 Q A %AI;4.L,,4 I A

.4 E-=

r=RVIFeR efflal MaR61al A

I J

LociI Powor h3nao Monitor LPM LTOP MCR EVS MFRV MAWMFW MG MFIV MSSV MSTV M-G=MTC M.D./T NAPS NMC ODCM PAM PIV P/T PORV PTLR PWR QA QPTR RAI RCS RG RHR RPV RSCS RTB RTP RTS RWST SAT SDM SEV SE SER SI SG SGTR SR SRM SRV SSER Logic System Functional Test Low Temperature Overpressure Protection Main Control Room Emergency Ventilation System Manual Feedwater Regulating Valve Main Feedwater Motor Generator Main Steam-Feedwater Isolation Valve Main Steam Safety Valve Main Steam Trip Valve Moderator Temperature Coefficient I4 1

,L r MARIAWA44~]!

IJ:I9'0-.::R1rAF I Hil North Anna Power Station II

Offsite Dose Calculation Manual Post-Accident Monitoring Pressure Isolation Valve Pressure/Temperature Power Operated Relief Valve Pressure Temperature Limits Report Pressurized Water Reactor Quality Assurance Quadrant Power Tilt Ratio Request for Additional Information r*

r l

I-al A MOU 13ucetK UR.'UrIi Rcac;r.t9r-Coolant PrcssUre BOUndari Reactor Coolant System Regulatory Guide Residual Heat Removal Reactor Pro:tection System Reactor Pressure Vessel Rod Sequence Control System Reactor Trip Breaker Rated Thermal Power Reactor Trip System Reactor Water Storage Tank Station

,uliary Transformer Shutdown oln Shutdown Margin NAM1ty Evaluatvio Safety Evaluation Safety Evaluation Report Safety Injection Steam Generator Steam Generator Tube Rupture Surveillance Requirement Source Range Monitor Safety/Relief Valve Supplemental Safety Eyaluation Report

STS Improved Standard Technical Specification, NUREG-1431, Rev. 1 SW Service Water TADOT Trip Actuating Device Operational Test TRM Technical Requirements Manual TS Technical Specification TSTF Technical Specifications Task Force (re: generic changes to the STS)

UHS Ultimate Heat Sink UFSAR Updated Final Safety Analysis Report V

Volt VEPCO Virginia Electric & Power Company

APPEN.4X-D ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NOs. NPF-4 and NPF-7 Dominion Generation shall comply with the following conditions on the schedules noted below:

Additional Conditions Date This amendment authorizes the The amendment shall relocation of certain Technical be implemented by Specification requirements to

[date].September 2,2002 licensee-controlled documents.

Implementation of this amendment shall include the relocation of these technical specification requirements to the appropriate documents, as described in Table R, which is that-a-e attached to the staff's [draft] Safety Evaluation enclosed with this amendment.

The schedule for the performance of new and revised Surveillance Requirements (SRs) shall be as follows:

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.

This amendment shall be implemented within XX-days-of-t date-ef-thisfor Unit 1, by the end of the sixteenth refueling outage and for Unit 2, by the end of the fifteenth refueling outage.

Amendment Number

Amendment Implementation Number Additional Conditions Date 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.

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.

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 implementation of this amendment.

Note for Tables A - Administrative Changes:

All "A" Discussion of Change (DOC) - Administrative Tables have been revised to include each entire A DOC, rather than just a summary of each change. For legibility reasons, the inclusion of the entire A DOC is not highlighted in red. If an A DOC has been revised for another reason (e.g., to incorporate a comment from the NRC), that change is highlighted in red, and the justification for that change is included in the right margin.

Attachment Improved Technical Specifications And Bases Revision 17 Virginia Electric and Power Company (Dominion)

North Anna Power Station Units I and 2

List of Effective Pages Improved Technical Specifications And Bases North Anna Power Station Units 1 and 2 LIST OF EFFECTIVE PAGES Technical Specifications Last Page f Conter age age North Anna Units 1 and 2 First P Title P Table o 1.1-1 1.2-1 1.3-1 1.4-1 2.0-1 3.0-1 3.1.1-1 3.1.2-1 3.1.3-1 3.1.4-1 3.1.5-1 3.1.6-1 3.1.7-1 3.1.8-1 3.1.9-1 3.2.1-1 3.2.2-1 3.2.3-1 3.2.4-1 3.3.1-1 3.3.2-1 3.3.3-1 3.3.4-1 3.3.5-1 3.4.1-1 3.4.2-1 3.4.3-1 3.4.4-1 3.4.5-1 3.4.6-1 3.4.7-1 3.4.8-1 its-i iv 1.1-7 1.2-3 1.3-12 1.4-7 2.0-1 3.0-5 3.1.1-1 3.1.2-2 3.1.3-2 3.1.4-3 3.1.5-2 3.1.6-3 3.1.7-3 3.1.8-1 3.1.9-2 3.2.1-3 3.2.2-2 3.2.3-1 3.2.4-4 3.3.1-17 3.3.42-11 3.3.3-3 3.3.4-2 3.3.5-2 3.4.1-2 3.4.2-1 3.4.3-6 3.4.4-1 3.4.5-2 3.4.6-2 3.4.7-3 3.4.8-2 Date Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments

-Amendments Amendments 231/212, 231/212 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212, 23 1/212, 23 1/212, 23 1/212, 231/212, 23 1/212, 23 1/212, 23 1/212, 23 1/212, 23 1/212, 231/212, 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/92/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/-02-/0-2 04/02/02 04/02/02 Page 1 of 6 04/02/02

List of Effective Pages Technical Specifications (continued)

First Page 3.4.9-1 3.4.10-1 3.4.11-1 3.4.12-1 3.4.13-1 3.4.14-1 3.4.15-1 3.4.16-1 3.4.17-1 3.4.18-1 3.4.19-1 3.5.1-1 3.5.2-1 3.5.3-1 3.5.4-1 3.5.5-1 3.5.6-1 3.6.1-1 3.6.2-1 3.6.3-1 3.6.4-1 3.6.5-1 3.6.6-1 3.6.7-1 3.6.8-1 3.6.9-1 3.7.1-1 3.7.2-1 3.7.3-1 3.7.4-1 3.7.5-1 3.7.6-1 3.7.7-1 3.7.8-1 3.7.9-1 3.7.10-1 3.7.11-1 Last Page 3.4.9-2 3.4.10-2 3.4.11-4 3.4.12-4 3.4.13-2 3.4.14-2 3.4.15-3 3.4.16-3 3.4.17-2 3.4.18-6 3.4.19-1 3.5.1-3 3.5.2-3 3.5.3-1 3.5.4-2 3.5.5-2 3.5.6-2 3.6.1-1 3.6.2-5 3.6.3-6 3.6.4-2 3.6.5-1 3.6.6-2 3.6.7-3 3.6.8-2 3.6.9-2 3.7.1-4 3.7.2-2 3.7.3-2 3.7.4-1 3.7.5-3 3.7.6-1 3.7.7-1 3.7.8-3 3.7.9-1 3.7.10-2 3.7.11-2 Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Date 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 231/212, 04/02/02 North Anna Units 1 and 2 Page 2 of 6 04/02/02

List of Effective Pages Technical Specifications (continued)

First Page 3.7.12-1 3.7.13-1 3.7.14-1 3.7.15-1 3.7.16-1 3.7.17-1 3.7.18-1 3.8.1-1 3.8.2-1 3.8.3-1 3.8.4-1 3.8.5-1 3.8.6-1 3.8.7-1 3.8.8-1 3.8.9-1 3.8.10-1 3.9.1-1 3.9.2-1 3.9.3-1 3.9.4-1 3.9.5-1 3.9.6-1 3.9.7-1 4.0-1 5.1-1 5.2-1 5.3-1 5.4-1 5.5-1 5.6-1 5.7-1 Last Page 3.7.12-2 3.7.13-3 3.7.14-2 3.7.15-1 3.7.16-2 3.7.17-1 3.7.18-4 3-.8.-1-18 3.8.2-3 3.8.3-3 3.8.4-4 3.8.5-2 3.8.6-4 3.8.7-2 3.8.8-2 3.8.9-3 3.8.10-2 3.9.1-1 3.9.2-1 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-3 3.9.7-1 4.0-2 5.1-1 5.2-3 5.3-1 5.4-1 5.5-19 5.6-5 5.7-5 North Anna Units 1 and 2 Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Amendments Date 231/212, 23 1/2 12, 231/2-12, 231/2-12, 231/2 12, 231/2 12, 231/212, 2-3.1/212, 231/212, 231/212, 231/212, 231/212, 231/212, 231/212.,

231/212-,

231/212, 231/212,

-231/212, 231/212, 231/212,"

231/212, 231/212, 231/2 12, 231/212, 231/212, 231/212,

-23 1/-2 112-,

231/2 12, 231/4212, 23 1/212, 231/212, 231/212, 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/02/02 04/)02/022 Page 3 of 6 04/02/02

List of Effective Pages First Page Title Page Table of Contents-i B 2.1.1-1 B 2.1.2-1 B 3.0-1 B 3.1.1-1 B 3.1.2-1 B 3.1.3-1 B 3.1.4-1 B 3.1.5-1 B 3.1.6-1 B 3.1.7-1 B 3.1.8-1 B 3.1.9-1 B 3.2.1-1 B 3.2.2-1 B 3.2.3-1 B 3.2.4-1 B 3.3.1-1 B 3.3.2-1 B 3.3.3-1 B 3.3.4-1 B 3.3.5-1 B 3.4.1-1 B 3.4.2-1 B 3.4.3-1 B 3.4.4-1 B 3.4.5-1 B 3.4.6-1 B 3.4.7-1 B 3.4.8-1 B 3.4.9-1 B 3.4.10-1 B 3.4.11-1 B 3.4.12-1 B 3.4.13-1 B 3.4.14-1 Bases Last Page iii B 2.1.1-4 B 2.1.2-4 B 3.0-22 B 3.1.1-6 B 3.1.2-6 B 3.1.3-6 B 3.1.4-10 B 3.1.5-5 B 3.1.6-7 B 3.1.7-7 B 3.1.8-3 B 3.1.9-8 B 3.2.1-9 B 3.2.2-6 B 3.2.3-5 B 3.2.4-7 B 3.3.1-59 B 3.3.2-45 B 3.3.3-14 B 3.3.4-6 B 3.3.5-8 B 3.4.1-5 B 3.4.2-3 B 3.4.3-7 B 3.4.4-4 B 3.4.5-5 B 3.4.6-5 B 3.4.7-6 B 3.4.8-4 B 3.4.9-5 B 3.4.10-5 B 3.4.11-8 B 3.4.12-12 B 3.4.13-6 B 3.4.14-8 North Anna Units 1 and 2 Date Revision Revi si on Revision Revi sion Revision Revision Revision Revi si on Revision Revi si on Revi si on Revi si on Revi si on Revision Revision Revi si on Revi sion Revision Revi sion Revision Revision Revision Revision Revision Revi si on Revi si on Revision Revi si on Revi si on Revi si on Revi si on Revi si on Revision Revision Revi si on Revi sion 0,

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List of Effective Pages North Anna Units 1 and 2 First Page B 3.4.15-1 B 3.4.16-1 B 3.4.17-1 B 3.4.18-1 B 3.4.19-1 B 3.5.1-1 B 3.5.2-1 B 3.5.3-1 B 3.5.4-1 B 3.5.5-1 B 3.5.6-1 B 3.6.1-1 B 3.6.2-1 B 3.6.3-1 B 3.6.4-1 B 3.6.5-1 B 3.6.6-1 B 3.6.7-1 B 3.6.8-1 B 3.6.9-1 B 3.7.1-1 B 3.7.2-1 B 3.7.3-1 B 3.7.4-1 B 3.7.5-1 B 3.7.6-1 B 3.7.7-1 B 3.7.8-1 B 3.7.9-1 B 3.7.10-1 B 3.7.11-1 B 3.7.12-1 B 3.7.13-1 B 3.7.14-1 B 3.7.15-1 B 3.7.16-1 B 3.7.17-1 Bases (continued)

Last Page B 3.4.15-5 B 3.4.16-6 B 3.4.17-3 B 3.4.18-8 B 3.4.19-4 B 3.5.1-8 B 3.5.2-11 B 3.5.3-3 B 3.5.4-6 B 3.5.5-4 B 3.5.6-5 B 3.6.1-4 B 3.6.2-8 B 3.6.3-11 B 3.6.4-4 B 3.6.5-4 B 3.6.6-6 B 3.6.7-9 B 3.6.8-5 B 3.6.9-5 B 3.7.1-7 B 3.7.2-6 B 3.7.3-6 B 3.7.4-4 B 3.7.5-9 B 3.7.6-4 B 3.7.7-3 B 3.7.8-7 B 3.7.9-4 B 3.7.10-6 B 3.7.11-4 B 3.7.12-7 B 3.7.13-7 B 3.7.14-5 B 3.7.15-3 B 3.7.16-3 B 3.7.17-3 Date Revi si on Revi si on Revi si on Revi si on Revi si ofn Revi si on Revi si on Revi si on Revisi-on Revi si onl Revi si on Revi si on Revi si on Revi si on Rev i sion Revi si on Revi si on Revi si on Revi si on Revi sion Revision Revi si on Rev i s i o n Revi sion Revi sion Rev i s ion Revi si on Revi si on Revi si on Revision Revi si on Revi si on Revi si on Revi si on Revi si on Revi si on Revi si on 04/02/02 Page 5 of 6 0,

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Last Page B 3.7.18-3 B 3.8.1-37 B 3.8.2-6 B 3.8.3-9 B 3.8.4-11 B 3.8.5-4 B 3.8.6-7 B 3.8.7-4 B 3.8.8-4 B 3.8.9-11 B 3.8.10-4 B 3.9.1-4 B 3.9.2-3 B 3.9.3-3 B 3.9.4-5 B 3.9.5-4 B 3.9.6-4 B 3.9.7-2 Page 6 of 6 04/02/02 Date Revision Revision Revision Revi si on Revi si on Revi si on Revi si on Revi si on Revi s ion Revi si on Revi si on Revi si on Revi si on Revi sion Revi sion Revi sion Revi sion Revision 0,

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TECHNICAL SPECIFICATIONS FOR NORTH ANNA UNITS 1 & 2

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS USE AND APPLICATION......

Definitions........

Logical Connectors Completion Times Frequency...........

SAFETY LIMITS (SLs)........

SLs...............

SL Violations........

LIMITING CONDITION FOR OPERATION (LCO)

APPLICABILITY SURVEILLANCE REQUIREMENT (SR)

APPLICABILITY.......

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN (SDM).............

Core Reactivity...........

Moderator Temperature Coefficient (MTC)

Rod Group Alignment Limits Shutdown Bank Insertion Limits Control Bank Insertion Limits.......

Rod Position Indication....

Primary Grade Water Flow Path Isolation PHYSICS TESTS Exceptions-MODE 2....

1.1-1 1.1-1

.1.2-1

.1.3-1 1.4-1 2.0-1

. 2.0-1 2.0-1 1.0 1.1 1.2 1.3 1.4 2.0 2.1 2.2 3.0 3.0 Valves 3.1 3.1.1 3.1.2 3.1.3 3.1.4 3.1.5 3.1.6 3.1.7 3.1.8 3.1.9 3.2 3.2.1 3.2.2 3.2.3 3.2.4 3.3 3.3.1 3.3.2 3.3.3 3.3.4 3.3.5 3.4 3.4.1 3.4.2 3.4.3 3.4.4 3.4.5 3.4.6 3.4.7 3.4.8 3.4.9 3.1.1-1 3.1.1-1 3.1.2-1 3.1.3-1 3.1.4-1 3.1.5-1 3.1.6-1 3.1.7-1 3.1.8-1 3.1.9-1 3.2.1-1 3.2.1-1 3.2.2-1 3.2.3-1 3.2.4-1 3.3.1-1 3.3.1-1 3.3.2-1 3.3.3-1

... 3.3.4-1 3.3.5-1 3.4.1-1 3.4.1-1 3.4.2-1 3.4.3-1 3.4.4-1 3.4.5-1 3.4.6-1 3.4.7-1 3.4.8-1

.3.4.9-1 Amendments 231/212, 04/02/02 3.0-1

. 3.0-4 POWER DISTRIBUTION LIMITS.

Heat Flux Hot Channel Factor (F(Z)).

Nuclear Enthalpy Rise Hot Channel Factor (FýH)

AXIAL FLUX DIFFERENCE (AFD)...............

QUADRANT POWER TILT RATIO (QPTR)

INSTRUMENTATION.

Reactor Trip System (RTS)

Instrumentation.

Engineered Safety Feature Actuation System (ESFAS) Instrumentation..............

Post Accident Monitoring (PAM)

Instrumentation Remote Shutdown System Loss of Power (LOP)

Emergency Diesel Generator (EDG)

Start Instrumentation............

REACTOR COOLANT SYSTEM (RCS)

RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Limits RCS Minimum Temperature for Criticality....

RCS Pressure and Temperature (P/T) Limits.

RCS Loops-MODES 1 and 2.................

RCS Loops-MODE 3 RCS Loops-MODE 4 RCS Loops-MODE 5, Loops Filled RCS Loops-MODE 5, Loops Not Filled Pressurizer........

North Anna Units 1 and 2 i

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS REACTOR COOLANT SYSTEM (RCS)

(continued)

Pressurizer Safety Valves...............

Pressurizer Power Operated Relief Valves (PORVs)....................

Low Temperature Overpressure Protection (LTOP)

System RCS Operational LEAKAGE................

RCS Pressure Isolation Valve (PIV)

Leakage RCS Leakage Detection Instrumentation.......

RCS Specific Activity.................

RCS Loop Isolation Valves...............

RCS Isolated Loop Startup...............

RCS Loops-Test Exceptions...............

3.4 3.4.10 3.4.11 3.4.12 3.4.13 3.4.14 3.4.15 3.4.16 3.4.17 3.4.18 3.4.19 3.5 3.5.1 3.5.2 3.5.3 3.5.4 3.5.5 3.5.6 3.6 3.6.1 3.6.2 3.6.3 3.6.4 3.6.5 3.6.6 3.6.7 3.6.8 3.6.9 3.7 3.7.1 3.7.2 3.7.3 3.7.4 3.7.5 3.7.6 3.7.7 3.7.8 3.7.9 CONTAINMENT SYSTEMS......................

Containment....................

Containment Air Locks.

Containment Isolation Valves Containment Pressure Containment Air Temperature..............

Quench Spray (QS)

System Recirculation Spray (RS)

System...........

Chemical Addition System Hydrogen Recombiners PLANT SYSTEMS..................

Main Steam Safety Valves (MSSVs)..........

Main Steam Trip Valves (MSTVs)

Main Feedwater Isolation Valves (MFIVs),

Main Feedwater Pump Discharge Valves (MFPDVs),

Main Feedwater Regulating Valves (MFRVs),

and Main Feedwater Regulating Bypass Valves (MFRBVs)

Steam Generator Power Operated Relief Vaives (SG PORVs)

Auxiliary Feedwater (AFW)

System Emergency Condensate Storage Tank (ECST)

Secondary Specific Activity..............

Service Water (SW)

System...............

Ultimate Heat Sink (UHS)

Amendments 231/212, 04/02/02 EMERGENCY CORE COOLING SYSTEMS (ECCS)

Accumulators ECCS-Operating ECCS-Shutdown...........

Refueling Water Storage Tank (RWST)

Seal Injection Flow..........

Boron Injection Tank (BIT)

...3.4.10-1

. 3.4.11-1

... 3.4.12-1

... 3.4.13-1

... 3.4.14-1

... 3.4.15-1

.3.4.16-1

... 3.4.17-1

... 3.4.18-1

... 3.4.19-1 3.5.1-1 3.5.1-1

...3.5.2-1

..3.5.3-1

..3.5.4-1

...3.5.5-1

..3.5.6-1 3.6.1-1 3.6.1-1 3.6.2-1 3.6.3-1 3.6.4-1 3.6.5-1 3.6.6-1

...3.6.7-1 3.6.8-1 3.6.9-1 3.7.1-1 3.7.1-1 3.7.2-1

...3.7.3-1 3.7.4-1

...3.7.5-1

..3.7.6-1 3.7.7-1 3.7.8-1 3.7.9-1 O

g 4

0 0

Q North Anna Units 1 and 2 i i

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 3.7 3.7.10 3.7.11 3.7.12 3.7.13 3.7.14 3.7.15 3.7.16 3.7.17 3.7.18 3.8 3.8.1 3.8.2 3.8.3 3.8.4 3.8.5 3.8.6 3.8.7 3.8.8 3.8.9 3.8.10 3.9 3.9.1 3.9.2 3.9.3 3.9.4 3.9.5 3.9.6 3.9.7 ELECTRICAL POWER SYSTEMS AC Sources-Operating AC Sources-Shutdown..........

Diesel Fuel Oil and Starting Air DC Sources-Operating DC Sources-Shutdown..........

Battery Cell Parameters.......

Inverters-Operating..........

Inverters-Shutdown Distribution Systems-Operating.

Distribution Systems-Shutdown.

REFUELING OPERATIONS Boron Concentration.

Primary Grade Water Flow Path Isolation Valves-MODE 6...............

Nuclear Instrumentation...........

Containment Penetrations Residual Heat Removal (RHR) and Coolant Circulation-High Water Level Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level.

Refueling Cavity Water Level DESIGN FEATURES......................

Site Location.....................

Reactor Core Fuel Storage ADMINISTRATIVE CONTROLS Responsibility PLANT SYSTEMS (continued)

Main Control Room/Emergency Switchgear Room (MCR/ ESGR) Emergency Ventilation System (EVS)-MODES 1, 2, 3, and 4 Main Control Room/Emergency Switchgear Room (MCR/ESGR) Air Conditioning System (ACS)

Emergency Core Cooling System (ECCS)

Pump Room Exhaust Air Cleanup System (PREACS)

Main Control Room/Emergency Switchgear Room (MCR/ESGR)

Bottled Air System.........

Main Control Room/Emergency Switchgear Room (MCR/ ESGR) Emergency Ventilation System (EVS)-During Movement of Recently Irradiated Fuel Assemblies Fuel Building Ventilation System (FBVS)

Fuel Storage Pool Water Level............

Fuel Storage Pool Boron Concentration.......

Spent Fuel Pool Storage................

3.9.1-1 3.9.1-1 S....

. 3.9.2-1 S....

. 3.9.3-1 S....

. 3.9.4-1 S....

. 3.9.5-1 S....

. 3.9.6-1 S....

. 3.9.7-1 4.0-1 4.0-1 4.0-1 4.0-1 5.1-1 Amendments 231/212, 04/02/02

...3.7.10-1

.3.7.11-1

...3.7.12-1

...3.7.13-1

...3.7.14-1

.3.7.15-1

...3.7.16-1

...3.7.17-1

...3.7.18-1 3.8.1-1 3.8.1-1

... 3.8.2-1

... 3.8.3-1

... 3.8.4-1

... 3.8.5-1

... 3.8.6-1

... 3.8.7-1

... 3.8.8-1

... 3.8.9-1

...3.8.10-1 4.0 4.1 4.2 4.3 5.0 5.1 North Anna Units 1 and 2 iii

TECHNICAL SPECIFICATIONS TABLE OF CONTENTS 5.0 ADMINISTRATIVE CONTROLS (continued) 5.2 Organization.......

5.2-1 5.3 Unit Staff Qualifications.......................

.. 5.3-1 5.4 Procedures.......

.. 5.4-1 5.5 Programs and Manuals.......

5.5-1 5.6 Reporting Requirements.........

5.6-1 5.7 High Radiation Area.....

. 5.7-1 Amendments 231/212, 04/02/02 North Anna Units 1 and 2 i v

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions NOTE ----------------

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTUATION LOGIC TEST AXIAL FLUX DIFFERENCE (AFD)

CHANNEL CALIBRATION ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output.

The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.

A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.1-1

Definitions 1.1 1.1 Definitions CHANNEL CHECK CHANNEL OPERATIONAL TEST (COT)

CORE ALTERATION CORE OPERATING LIMITS REPORT (COLR)

DOSE EQUIVALENT 1-131 A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,

AEC, 1962, "Calculation of Distance Factors for Power and Test Reactor Sites."

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.1-2

Definitions 1.1 1.1 Definitions E-AVERAGE DISINTEGRATION ENERGY ENGINEERED SAFETY FEATURE (ESF)

RESPONSE

TIME LEAKAGE E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives > 15 minutes, making up at least 95% of the total noniodine activity in the coolant.

The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank;

2.

LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS)

LEAKAGE through a steam generator (SG) to the Secondary System; (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.1-3

Definitions 1.1 1.1 Definitions LEAKAGE (continued)

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE; through a component body,

c. Pressure Boundary LEAKAGE LEAKAGE (except SG LEAKAGE) nonisolable fault in an RCS pipe wall, or vessel wall.

MASTER RELAY TEST MODE OPERABLE-OPERABILITY PHYSICS TESTS A MASTER RELAY TEST shall consist of energizing all master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST shall include a continuity check of each associated required slave relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 14, Initial Tests and Operation, of the UFSAR; (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.1-4

Definitions 1.1 1.1 Definitions PHYSICS TESTS (continued)

QUADRANT POWER TILT RATIO (QPTR)

RATED THERMAL POWER (RTP)

REACTOR TRIP SYSTEM (RTS)

RESPONSE TIME SHUTDOWN MARGIN (SDM)

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2893 MWt.

The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC.

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

Amendments 231/212, 04/02/02 North Anna Units I and 2 1.1-5

Definitions 1.1 1.1 Definitions SLAVE RELAY TEST STAGGERED TEST BASIS THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)

A SLAVE RELAY TEST shall consist of energizing all slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY.

The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy.

The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.1-6

Definitions 1.1 Table 1.1-1 (page 1 of 1)

MODES REACTIVITY

% RATED AVERAGE REACTOR CONDITION THERMAL COOLANT TEMPERATURE MODE TITLE (keff)

POWER(a)

(OF) 1 Power Operation

> 0.99

> 5 NA 2

Startup

> 0.99

• 5 NA 3

Hot Standby

< 0.99 NA

> 350 4

Hot Shutdown(b)

< 0.99 NA 350 > Tavg > 200 5

Cold Shutdown(b)

< 0.99 NA

< 200 6

Refueling(c)

NA NA NA Excluding decay heat.

All reactor vessel head closure bolts fully tensioned.

One or more reactor vessel head closure bolts less than fully tensioned.

Amendments 231/212, 04/02/02 (a)

(b)

(c)

North Anna Units I and 2 1.1-7

Intentionally Blank

Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors The purpose of this section is to explain the meaning of logical connectors.

Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.

BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors.

When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.

Amendments 231/212, 04/02/02 PURPOSE North Anna Units 1 and 2 1.2-1

Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES Amendments 231/212, 04/02/02 The following examples illustrate the use of logical connectors.

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

LCO not met.

A.1 Verify...

AND A.2 Restore In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.

North Anna Units 1 and 2 1.2-2

Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES (continued)

EXAMPLE 1.2-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

LCO not met.

A.1 Trip...

OR A.2.1 Verify...

AND A.2.2.1 Reduce...

OR A.2.2.2 Perform...

OR A.3 Align...

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND.

Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.2-3

Intentionally Blank

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

BACKGROUND DESCRIPTION Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).

The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.

If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time. When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition.

Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.

(continued)

Amendments 231/212, 04/02/02 PURPOSE North Anna Units 1 and 2 1.3-1

Completion Times 1.3 1.3 Completion Times DESCRIPTION However, when a subsequent train, subsystem, component, or (continued) variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

a. Must exist concurrent with the first inoperability; and

b. Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; or

b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery.

." Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.3-2

Completion Times 1.3 1.3 Completion Times The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Condition B has two Required Actions. Each has its own separate Completion Time. Each is referenced to the time that Condition B Required Action Completion Time is entered.

The Required Actions of Condition B are to be in MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> AND in MODE 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> is allowed for reaching MODE 3 and a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br />) is allowed for reaching MODE 5 from the time that Condition B was entered. If MODE 3 is reached within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the time allowed for reaching MODE 5 is the next 33 hours3.819444e-4 days <br />0.00917 hours <br />5.456349e-5 weeks <br />1.25565e-5 months <br /> because the total time allowed for reaching MODE 5 is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

If Condition B is entered while in MODE 3, the for reaching MODE 5 is the next 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

time allowed Amendments 231/212, 04/02/02 EXAMPLES North Anna Units 1 and 2 1.3-3

Completion 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days inoperable.

OPERABLE status.

B. Required B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> When a pump is declared inoperable, Condition A is entered.

If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Condition A and B are exited, and therefore, the Required Actions of Condition B may be terminated.

When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump.

LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump.

The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A.

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.

(continued)

Amendments 231/212, 04/02/02 Times 1.3 North Anna Units 1 and 2 1.3-4

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-2 (continued)

On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for

> 7 days.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.3-5

Completion 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore Function X 7 days Function X train to OPERABLE train status.

AND inoperable.

10 days from discovery of failure to meet the LCO B. One B.1 Restore Function Y 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Function Y train to OPERABLE train status.

AND inoperable.

10 days from discovery of failure to meet the LCO C. One C.1 Restore Function X 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Function X train to OPERABLE train status.

inoperable.

OR AND C.2 Restore Function Y 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> One train to OPERABLE Function Y status.

train inoperable.

(continued)

Amendments 231/212, 04/02/02 Times 1.3 North Anna Units 1 and 2 1.3-6

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-3 (continued)

When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e.,

initial entry into Condition A).

The Completion Times of Conditions A and B are modified by a logical connector with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met.

In this example, without the separate Completion Time, it would be possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO.

The separate Completion Time modified by the phrase "from discovery of failure to meet the LCO" is designed to prevent indefinite continued operation while not meeting the LCO.

This Completion Time allows for an exception to the normal "time zero" for beginning the Completion Time "clock".

In this instance, the Completion Time "time zero" is specified as commencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.3-7

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve(s) 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves to OPERABLE inoperable, status.

B. Required B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met.

B.2 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis.

Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.

Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provided this does not result in any subsequent valve being inoperable for > 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

If the Completion Time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (including the extension) expires while one or more valves are still inoperable, Condition B is entered.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.3-8

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-5 ACTIONS NOTE--

Separate Condition entry is allowed for each inoperable valve.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> valves OPERABLE status.

inoperable.

B. Required B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met.

B.2 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> The Note above the ACTIONS Table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.

The Note allows Condition A to be entered separately for each inoperable valve, and Completion Times tracked on a per valve basis. When a valve is declared inoperable, Condition A is entered and its Completion Time starts. If subsequent valves are declared inoperable, Condition A is entered for each valve and separate Completion Times start and are tracked for each valve.

If the Completion Time associated with a valve in Condition A expires, Condition B is entered for that valve.

If the Completion Times associated with subsequent valves in Condition A expire, Condition B is entered separately for each valve and separate Completion Times start and are (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.3-9

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-5 (continued) tracked for each valve. If a valve that caused entry into Condition B is restored to OPERABLE status, Condition B is exited for that valve.

Since the Note in this example allows multiple Condition entry and tracking of separate Completion Times, Completion Time extensions do not apply.

EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable.

SR 3.x.x.x.

OR A.2 Reduce THERMAL 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> POWER to

< 50% RTP.

B. Required B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated Completion Time not met.

Entry into Condition A offers a choice between Required Action A.1 or A.2. Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance.

The initial 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval.

If Required Action A.1 is followed, and the Required Action is not met within the Completion Time (plus (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.3-10

Completion Times 1.3 1.3 Completion Times EXAMPLES Required Action A.1 has two Completion Completion Time begins at the time the and each "Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter" performance of Required Action A.1.

Times. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Condition is entered interval begins upon (continued)

Amendments 231/212, 04/02/02 EXAMPLE 1.3-6 (continued) the extension allowed by SR 3.0.2), Condition B is entered.

If Required Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is not met, Condition B is entered.

If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A.

EXAMPLE 1.3-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem A.1 Verify affected 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, subsystem isolated.

AND Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter AND A.2 Restore subsystem 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to OPERABLE status.

B. Required B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Action and associated AND Completion Time not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> North Anna Units 1 and 2 i1.3-11

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued)

If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or any subsequent 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.

IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.

Amendments 231/212, 04/02/02 North Anna Units I and 2 1.3-12

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO.

An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certain Notes in the Surveillance column that modify performance requirements.

Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillance, or both.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" or "performed" in these instances conveys specific meanings.

A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria.

(continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-1

Frequency 1.4 1.4 Frequency DESCRIPTION (continued)

Some Surveillances contain notes that modify the Frequency of performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied:

a. The Surveillance is not required to be met in the MODE or other specified condition to be entered; or

b. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or

c. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.

Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discusses these special situations.

EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS).

The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time.

Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-2

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-1 (continued) extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO).

If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so would result in a violation of SR 3.0.4.

EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits.

Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after

Ž 25% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to

Ž 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

(continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-3

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-2 (continued)

The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND").

This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2.

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example).

If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

EXAMPLE 1.4-3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOTE-------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after Ž 25% RTP.

Perform channel adjustment.

7 days The interval continues, whether or not the unit operation is

< 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after power reaches Ž 25% RTP to perform the Surveillance. The Surveillance is still considered to be performed within the "specified Frequency." Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2) interval, but operation was < 25% RTP, it would not constitute a failure of the SR or failure to meet the LCO. Also, no violation of SR 3.0.4 occurs when changing

MODES, even with the 7 day Frequency not met, provided operation does not exceed 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with power Ž 25% RTP.

(continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-4

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-3 (continued)

Once the unit reaches 25% RTP, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> would be allowed for completing the Surveillance. If the Surveillance were not performed within this 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> interval, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

EXAMPLE 1.4-4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOTE-------------

Only required to be met in MODE 1.

Verify leakage rates are within limits.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-4 specifies that the requirements of this Surveillance do not have be met until the unit is in MODE 1.

The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of the Surveillance.

Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2),

but the unit was not in MODE 1, there would be no failure of the SR nor failure to meet the LCO. Therefore, no violation of SR 3.0.2 occurs when changing MODES, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change was not made into MODE 1. Prior to entering MODE 1 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met),

SR 3.0.4 would require satisfying the SR.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-5

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-5 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOTE--------------

Only required to be performed in MODE 1.

Perform complete cycle of the valve.

7 days The interval continues, whether or not the unit operation is in MODE 1, 2, or 3 (the assumed Applicability of the associated LCO) between performances.

As the Note modifies the required performance of the Surveillance, the Note is construed to be part of the "specified Frequency." Should the 7 day interval be exceeded while operation is not in MODE 1, this Note allows entry into and operation in MODES 2 and 3 to perform the Surveillance.

The Surveillance is still considered to be performed within the "specified Frequency" if completed prior to entering MODE 1. Therefore, if the Surveillance were not performed within the 7 day (plus the extension allowed by SR 3.0.2)

interval, but operation was not in MODE 1, it would not constitute a failure of the SR or failure to meet the LCO.

Also, no violation of SR 3.0.4 occurs when changing MODES, even with the 7 day Frequency not met, provided operation does not result in entry into MODE 1.

Once the unit reaches MODE 1, the requirement for the Surveillance to be performed within its specified Frequency applies and would require that the Surveillance had been performed.

If the Surveillance were not performed prior to entering MODE 1, there would then be a failure to perform a Surveillance within the specified Frequency, and the provisions of SR 3.0.3 would apply.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-6

Frequency 1.4 1.4 Frequency EXAMPLES EXAMPLE 1.4-6 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

---

NOTE-------------

Not required to be met in MODE 3.

Verify parameter is within limits.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Example 1.4-6 specifies that the requirements of this Surveillance do not have to be met while the unit is in MODE 3 (the assumed Applicability of the associated LCO is MODES 1, 2, and 3).

The interval measurement for the Frequency of this Surveillance continues at all times, as described in Example 1.4-1. However, the Note constitutes an "otherwise stated" exception to the Applicability of this Surveillance. Therefore, if the Surveillance were not performed within the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> interval (plus the extension allowed by SR 3.0.2), and the unit was in MODE 3, there would be no failure of the SR nor failure to meet the LCO.

Therefore, no violation of SR 3.0.4 occurs when changing MODES to enter MODE 3, even with the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency exceeded, provided the MODE change does not result in entry into MODE 2. Prior to entering MODE 2 (assuming again that the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency were not met),

SR 3.0.4 would require satisfying the SR.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 1.4-7

Intentionally Blank

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR; and the following SLs shall not be exceeded.

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall maintained greater than or equal to the 95/95 DNBR criterion for the DNB correlations and methodologies specified in Section 5.6.5.

be 2.1.1.2 The peak fuel centerline temperature shall be maintained

< 47000F.

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained

• 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 2.0-1

Intentionally Blank

LCO APPLICABILITY 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO)

APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2 and LCO 3.0.7.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s),

completion of the Required Action(s) is not required unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to place the unit, as applicable, in:

a. MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />;

b.

MODE 4 within 13 hours1.50463e-4 days <br />0.00361 hours <br />2.149471e-5 weeks <br />4.9465e-6 months <br />; and

c.

MODE 5 within 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br />.

Exceptions to this Specification are stated in the individual Specifications.

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1, 2, 3,

and 4.

LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specific condition in the Applicability for an unlimited period of

time, (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.0-1

LCO APPLICABILITY 3.0 3.0 LCO APPLICABILITY LCO 3.0.4 (continued)

b. After performance of a risk evaluation, consideration of the results, determination of the acceptability of the MODE change, and establishment of risk management actions, if appropriate, or

c. When a specific value or parameter allowance has been approved by the NRC.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

LCO 3.0.4 is only applicable for entry into a MODE or other specified condition in the Applicability in MODES 1, 2, 3, and 4.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Specification 5.5.14, "Safety Function Determination Program (SFDP)."

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.0-2

LCO APPLICABILITY 3.0 3.0 LCO APPLICABILITY LCO 3.0.7 Test Exception LCOs 3.1.9 and 3.4.19 allow specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional.

When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.0-3

SR APPLICABILITY 3.0 3.0 SURVEILLANCE REQUIREMENT (SR)

APPLICABILITY SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

Surveillances may be performed by any series of sequential, overlapping, or total steps.

The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per.

.." basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

(continued)

Amendments 231/212, 04/02/02 SR 3.0.1 SR 3.0.2 SR 3.0.3 North Anna Units 1 and 2 3.0-4

SR APPLICABILITY 3.0 3.0 SR APPLICABILITY SR 3.0.3 (continued)

SR 3.0.4 When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency. When an LCO is not met, entry into a MODE or other specific condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specific condition in the Applicability for an unlimited period of

time,

b. After performance of a risk evaluation, consideration of the results, determination of the acceptability of the MODE change, and establishment of risk management actions, if appropriate, or

c.

When a specific value or parameter allowance has been approved by the NRC.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

SR 3.0.4 is only applicable for entry into a MODE specified condition in the Applicability in MODES and 4.

or other 1, 2, 3 Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.0-5

Intentionally Blank

SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO

3.1.1 APPLICABILITY

SDM shall be within the limits provided in the COLR.

MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

SDM not within limit.

A.1 Initiate boration to 15 minutes restore SDM to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within limits.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Amendments 231/212, 04/02/02 North Anna Units I and 2 3.1.1-1

Intentionally Blank

Core Reactivity 3.1.2 3.1 REACTIVITY CONTROL SYSTEMS 3.1.2 Core Reactivity LCO

3.1.2 APPLICABILITY

The measured core reactivity shall be within +/- 1% Ak/k of predicted values.

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Measured core A.1 Re-evaluate core 7 days reactivity not within design and safety limit.

analysis, and determine that the reactor core is acceptable for continued operation.

AND A.2 Establish appropriate 7 days operating restrictions and SRs.

B.

Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.2-1

Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS NOTE---------------

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within

+/- 1% Ak/k of predicted values.

FREQUENCY Once prior to entering MODE 1 after each refueling AND NOTE -----

Only required after 60 EFPD 31 EFPD thereafter Amendments 231/212, 04/02/02 SR 3.1.2.1 SURVEILLANCE North Anna Units I and 2 3.1.2-2

MTC 3.1.3 3.1 REACTIVITY CONTROL SYSTEMS 3.1.3 Moderator Temperature Coefficient (MTC)

LCO

3.1.3 APPLICABILITY

The MTC shall be maintained within the limits specified in the COLR. The upper limit specified in the COLR shall be

<0.6 x 10-4 Ak/k/°F when < 70% RTP, and ! 0.0 Ak/k/°F when

>70% RTP.

MODE 1 and MODE 2 with keff Ž 1.0 for the upper MTC limit, MODES 1, 2, and 3 for the lower MTC limit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

MTC not within upper A.1 Establish 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />

limit, administrative withdrawal limits for control banks to maintain MTC within limit.

B.

Required Action and B.1 Be in MODE 2 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion keff < 1.0.

Time of Condition A not met.

C.

MTC not within lower C.1 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.3.1 Verify MTC is within upper limit.

Once prior to entering MODE 1 after each refueling Amendments 231/212, 04/02/02 North Anna Units I and 2 3.1.3-1

MTC 3.1.3 SURVEILLANCE REQUIREMENTS NOTES--------------

1. Not required to be performed until 7 effective full power days (EFPD) after reaching the equivalent of an equilibrium RTP all rods out (ARO) boron concentration of 300 ppm

2. If the MTC is more negative than the 300 ppm Surveillance limit (not LCO limit) specified in the COLR, SR 3.1.3.2 shall be repeated once per 14 EFPD during the remainder of the fuel cycle.

3. SR 3.1.3.2 need not be repeated if the MTC measured at the equivalent of equilibrium RTP-ARO boron concentration of

• 60 ppm is less negative than the 60 ppm Surveillance limit specified in the COLR.

Verify MTC is within lower limit.

FREQUENCY Once each cycle Amendments 231/212, 04/02/02 SR 3.1.3.2 SURVEILLANCE I

I North Anna Units 1 and 2 3.1.3-2

Rod Group Alignment 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits LCO

3.1.4 APPLICABILITY

All shutdown and control rods shall be OPERABLE.

AND Individual indicated rod positions shall be within 12 steps of their group step counter demand position.

NOTE -------------

When THERMAL POWER is

• 50% RTP, the indicated position of each rod as determined by its individual rod position indicator may be within 24 steps from its group step counter demand position for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

This NOTE is not applicable for control rods known to be greater than 12 steps from the rod group step counter demand position.

MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more rod(s)

A.1.1 Verify SDM to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, within the limits provided in the COLR.

OR A.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM to within limit.

AND A.2 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Amendments 231/212, 04/02/02 Limits 3.1.4 North Anna Units 1 and 2 3.1.4-1

Rod Group Alignment Limits 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

One rod not within alignment limits.

B.1.1 Verify SDM to be within the limits provided in the COLR.

OR B.1.2 Initiate boration to restore SDM to within limit.

AND B.2.1 Reduce THERMAL POWER to

• 75% RTP.

OR B.2.2.1 Perform SR 3.2.1.1.

AND B.2.2.2 Perform SR 3.2.2.1.

AND B.3 Re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 2 hours 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 72 hours 5 days C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition B not met.

Amendments 231/212, 04/02/02 North Anna Units I and 2 3.1.4-2

Rod Group Alignment Limits 3.1.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D.

More than one rod not D.1.1 Verify SDM to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> within alignment within the limit limit.

provided in the COLR.

OR D.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore required SDM to within limit.

AND D.2 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> alignment limit.

SR 3.1.4.2 Verify rod freedom of movement 92 days (trippability) by moving each rod not fully inserted in the core Ž 10 steps in either direction.

SR 3.1.4.3 Verify rod drop time of each rod, from the Prior to reactor fully withdrawn position, is

• 2.7 seconds criticality from the beginning of decay of stationary after each gripper coil voltage to dashpot entry, removal of the with:

reactor head

a.

Tavg Ž 500'F; and

b. All reactor coolant pumps operating.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.4-3

Intentionally Blank

Shutdown Bank Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be within insertion limits specified in the COLR.

APPLICABILITY:

MODES 1 and 2.

NOTE -------------

This LCO is not applicable while performing SR 3.1.4.2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more shutdown A.1.1 Verify SDM to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> banks not within within the limits limits for reasons provided in the COLR.

other than Condition B.

OR A.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM to within limit.

AND A.2 Restore shutdown banks 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to within limits.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.5-1

Shutdown Bank Insertion Limits 3.1.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

One shutdown bank B.1 Verify SDM to be Once per inserted

• 18 steps within the limits 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> below the insertion provided in the COLR.

limit and immovable.

AND AND B.2 Restore the shutdown 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Each control and bank to within shutdown rod within insertion limit.

limits of LCO 3.1.4.

AND Each control bank within the insertion limits of LCO 3.1.6.

C.

Required Action and C.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> insertion limits specified in the COLR.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.5-2

Control Bank Insertion Limits 3.1.6 3.1 REACTIVITY CONTROL SYSTEMS 3.1.6 Control Bank Insertion Limits LCO

3.1.6 APPLICABILITY

Control banks shall be within the insertion, sequence, and overlap limits specified in the COLR.

MODE 1, MODE 2 with keff Ž 1.0.

NOTE -------------

This LCO is not applicable while performing SR 3.1.4.2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Control bank sequence A.1.1 Verify SDM to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or overlap limits not within the limits met.

provided in the COLR.

OR A.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM to within limit.

AND A.2 Restore control bank 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> sequence and overlap to within limits.

B.

Control bank insertion B.1.1 Verify SDM to be 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> limits not met for within the limits reasons other than provided in the COLR.

Condition C.

OR B.1.2 Initiate boration to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> restore SDM to within limit.

AND (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.6-1

Control Bank Insertion Limits 3.1.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

(continued)

B.2 Restore control 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> bank(s) to within limits.

C.

Control bank A, B, C.1 Verify SDM to be Once per or C inserted within the limits 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

< 18 steps below the provided in the COLR.

insertion limit and immovable.

AND AND C.2 Restore the control 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> bank to within Each control and insertion limit.

shutdown rod within limits of LCO 3.1.4.

AND Each shutdown bank within the insertion limits of LCO 3.1.5.

D. Required Action and D.1 Be in MODE 2 with 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Keff < 1.0.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> position is within the insertion limits prior to specified in the COLR.

achieving criticality Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.6-2

Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank is within the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> insertion limits specified in the COLR.

SR 3.1.6.3 Verify each control bank not fully 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> withdrawn from the core is within the sequence and overlap limits specified in the COLR.

Amendments 231/212, 04/02/02 3.1.6-3 North Anna Units 1 and 2

Intentionally Blank

Rod Position Indication 3.1.7 3.1 REACTIVITY CONTROL SYSTEMS 3.1.7 Rod Position Indication LCO

3.1.7 APPLICABILITY

The Rod Position Indication (RPI)

System and the Demand Position Indication System shall be OPERABLE.

MODES 1 and 2.

NOTE Separate Condition entry is allowed for each inoperable rod position indicator and each demand position indicator.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One RPI per group A.1 Verify the position Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> inoperable for one or indirectly of the rods more groups.

with inoperable position indicators by using movable incore detectors.

OR A.2 Reduce THERMAL POWER 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to

• 50% RTP.

Amendments 231/212, 04/02/02 ACTIONS North Anna Units 1 and 2 3.1.7-1

Rod Position Indication 3.1.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

More than one RPI per B.1 Place the control rods Immediately group inoperable, under manual control.

AND B.2 Monitor and record RCS Once per 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Tavg.

AND B.3 Verify the position of Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> the rods with inoperable position indicators indirectly by using the movable incore detectors.

AND B.4 Restore inoperable 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> position indicator to OPERABLE status such that a maximum of one RPI per group is inoperable.

C.

One or more rods with C.1 Verify the position 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> inoperable position indirectly of the rods indicators have been with inoperable moved in excess of position indicators by 24 steps in one using movable incore direction since the detectors.

last determination of the rod's position.

OR C.2 Reduce THERMAL POWER 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to

• 50% RTP.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.7-2

Rod Position Indication 3.1.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. One demand position D.1.1 Verify by Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> indicator per bank administrative means inoperable for one or all RPIs for the more banks, affected banks are OPERABLE.

AND D.1.2 Verify the most Once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> withdrawn rod and the least withdrawn rod of the affected banks are

• 12 steps apart.

OR D.2 Reduce THERMAL POWER 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to

• 50% RTP.

E.

Required Action and E.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.1 Perform CHANNEL CALIBRATION of each RPI.

18 months Amendments 231/212, 04/02/02 North Anna Units I and 2 3.1.7-3

Intentionally Blank

Primary Grade Water Flow Path Isolation Valves 3.1.8 3.1 REACTIVITY CONTROL SYSTEMS 3.1.8 Primary Grade Water Flow Path Isolation Valves LCO

3.1.8 APPLICABILITY

Each valve used to isolate primary grade water flow paths shall be secured in the closed position.

NOTE -------------

Primary grade water flow path isolation valves may be opened under administrative control for planned boron dilution or makeup activities.

MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

---

NOTE ---------

A.1 Suspend positive Immediately Required Action A.3 reactivity additions.

must be completed whenever Condition A AND is entered.

A.2 Secure valves in 15 minutes closed position.

One or more valves not secured in closed AND position.

A.3 Perform SR 3.1.1.1.

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Verify each valve in the affected flow path Within that isolates primary grade water flow 15 minutes paths is locked, sealed, or otherwise following a secured in the closed position.

boron dilution or makeup activity Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.8-1

Intentionally Blank

PHYSICS TESTS Exceptions-MODE 2 3.1.9 3.1 REACTIVITY CONTROL SYSTEMS 3.1.9 PHYSICS TESTS Exceptions-MODE 2 LCO 3.1.9 During the performance of PHYSICS TESTS, the requirements of LCO LCO LCO LCO LCO 3.1.3, 3.1.4, 3.1.5, 3.1.6, 3.4.2, "Moderator Temperature Coefficient (MTC)";

"Rod Group Alignment Limits";

"Shutdown Bank Insertion Limits";

"Control Bank Insertion Limits"; and "RCS Minimum Temperature for Criticality" may be suspended and the number of required channels for LCO 3.3.1, "RTS Instrumentation," Functions 2, 3, and 18.d, may be reduced to "3" required channels, provided:

a.

RCS lowest loop average temperature is Ž 531°F;

b. SDM is within the limits provided in the COLR; and

c. THERMAL POWER is

• 5% RTP.

APPLICABILITY:

During PHYSICS TESTS initiated in MODE 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

SDM not within limit.

A.1 Initiate boration to 15 minutes restore SDM to within limit.

AND A.2 Suspend PHYSICS TESTS 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> exceptions.

B.

THERMAL POWER not B.1 Open reactor trip Immediately within limit, breakers.

C.

RCS lowest loop C.1 Restore RCS lowest 15 minutes average temperature loop average not within limit, temperature to within limit.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.9-1

PHYSICS TESTS Exceptions-MODE 2 3.1.9 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3.

15 minutes associated Completion Time of Condition C not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.9.1 Perform a CHANNEL OPERATIONAL TEST on power Prior to range and intermediate range channels per initiation of SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1.

PHYSICS TESTS SR 3.1.9.2 Verify the RCS lowest loop average 30 minutes temperature is Ž 5310F.

SR 3.1.9.3 Verify THERMAL POWER is

• 5% RTP.

30 minutes SR 3.1.9.4 Verify SDM to be within the limits provided 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in the COLR.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.1.9-2

FQ (Z) 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

LCO

3.2.1 APPLICABILITY

ACTIONS FQ(Z), as approximated by Fm (Z), shallI be within the Ilimits specified in the COLR.

MODE 1.

CONDITION REQUIRED ACTION COMPLETION TIME A.

F*(Z) not within limit.

A.1 Reduce AFD limits Ž 1%

for each 1% FQ(Z) exceeds limit.

OR A.2.1 Reduce THERMAL POWER

Ž 1% RTP for each 1%

F*(Z) exceeds limit.

AND A.2.2 Reduce Power Range Neutron Flux-High trip setpoints Ž 1% for each 1% Fm(Z) exceeds limit.

AND A.2.3 Reduce Overpower AT trip setpoints Ž 1%

for each 1% FQ(Z) exceeds limit.

AND A.2.4 Perform SR 3.2.1.1.

15 minutes after each FM(Z) determination 15 minutes after each FQ(Z) determination 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each FQ(Z) determination 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after each F (Z) determination Prior to increasing THERMAL POWER above the limit of Required Action A.2.1 Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.1-1

FQ(Z) 3.2.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

Required Action and B.1 Be in MODE 2.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS NOTE ----------------

During power escalation, THERMAL POWER may be increased until a power level for extended operation has been achieved, at which a power distribution map is obtained.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.1-2

FQ (Z) 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1

---

NOTE-----------------

If Fm(Z) measurements indicate maximum over z K(Z) has increased since the previous evaluation of F*(Z):

a. Increase Fm(Z) by the appropriate factor and verify FM(Z) is still within limits; or

b. Repeat SR 3.2.1.1 once per 7 EFPD until two successive flux maps indicate maximum over z LK(Z) has not increased.

Verify Fm(Z) is within limit.

Once after each refueling prior to THERMAL POWER exceeding 75% RTP AND Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after achieving equilibrium conditions after exceeding, by

Ž 10% RTP, the THERMAL POWER at which F*(Z) was last verified AND 31 EFPD thereafter Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.1-3

Intentionally Blank

FNH 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FAH)

LCO

3.2.2 APPLICABILITY

FAH shall be within the limits specified in the COLR.

MODE 1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

---

NOTE ---------

A.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Required Actions A.3 to < 50% RTP.

and A.4 must be completed whenever AND Condition A is entered.

A.2 Reduce Power Range 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Neutron Flux-High trip setpoints to FaNH not within limit.

_ 55% RTP.

AND A.3 Perform SR 3.2.2.1.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> AND (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.2-1

FAH 3.2.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.4

---

NOTE------

THERMAL POWER does not have to be reduced to comply with this Required Action.

Perform SR 3.2.2.1.

Prior to THERMAL POWER exceeding 50% RTP AND Prior to THERMAL POWER exceeding 75% RTP AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after THERMAL POWER reaching

> 95% RTP B. Required Action and B.1 Be in MODE 2.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FaH is within limits specified in Once after each the COLR.

refueling prior to THERMAL POWER exceeding 75% RTP AND 31 EFPD thereafter Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.2-2

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO

3.2.3 APPLICABILITY

The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

NOTE -------------

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

MODE 1 with THERMAL POWER Ž 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

AFD not within limits.

A.1 Reduce THERMAL POWER 30 minutes to < 50% RTP.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AFD within limits for each OPERABLE 7 days excore channel.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.3-1

Intentionally Blank

QPTR 3.2.4 3.2 POWER DISTRIBUTION LIMITS 3.2.4 QUADRANT POWER TILT RATIO (QPTR)

LCO

3.2.4 APPLICABILITY

The QPTR shall be

• 1.02.

MODE 1 with THERMAL POWER > 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

QPTR not within limit.

A.1 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after

Ž3% from RTP for each each QPTR 1% of QPTR > 1.00.

determination AND A.2 Determine QPTR.

Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND A.3 Perform SR 3.2.1.1 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after SR 3.2.2.1.

achieving equilibrium Conditions from a THERMAL POWER reduction per Required Action A.1 AND Once per 7 days thereafter AND (continued)

Amendments 231/212, 04/02/02 North Anna Units I and 2 3.2.4-1

QPTR 3.2.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.4 Reevaluate safety Prior to analyses and confirm increasing results remain valid THERMAL POWER for duration of above the limit operation under this of Required condition.

Action A.1 AND A.5

---

NOTES------

1. Perform Required Action A.5 only after Required Action A.4 is completed.

2. Required Action A.6 shall be completed whenever Required Action A.5 is performed.

Normalize excore Prior to detectors to restore increasing QPTR to within limits.

THERMAL POWER above the limit of Required Action A.1 AND (continued)

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.4-2

QPTR 3.2.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.6

---

NOTE------

Perform Required Action A.6 only after Required Action A.5 is completed.

Perform SR 3.2.1.1 and Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.2.2.1.

after achieving equilibrium Conditions at RTP not to exceed 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> after increasing THERMAL POWER above the limit of Required Action A.1 B.

Required Action and B.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> associated Completion to

• 50% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1 NOTES---------------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER

• 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2.

SR 3.2.4.2 may be performed in lieu of this Surveillance.

Verify QPTR is within limit by calculation.

7 days Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.2.4-3

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.2

---

NOTE---------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER > 75% RTP.

Verify QPTR is within limit using the 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> movable incore detectors.

Amendments 231/212, 04/02/02 North Anna Units I and 2 3.2.4-4

RTS Instrumentation 3.3.1 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS)

Instrumentation LCO

3.3.1 APPLICABILITY

The RTS instrumentation for each Function in Table 3.3.1-1 shall be OPERABLE.

According to Table 3.3.1-1.

NOTE -------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more Functions A.1 Enter the Condition Immediately with one or more referenced in required channels or Table 3.3.1-1 for the trains inoperable, channel(s) or train(s).

B.

One Manual Reactor B.1 Restore channel to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Trip channel OPERABLE status.

inoperable.

OR B.2 Be in MODE 3.

54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> C.

One channel or train C.1 Restore channel or 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable, train to OPERABLE status.

OR C.2.1 Initiate action to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> fully insert all rods.

AND (continued)

Amendments 231/212, 04/02/02 ACTIONS North Anna Units 1 and 2 3.3.1-1

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME C.

(continued)

C.2.2 Place the Rod Control 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> System in a condition incapable of rod withdrawal.

D.

One Power Range Neutron Flux-High channel inoperable.

NOTE ---------

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing and setpoint adjustment of other channels.

D.1.1 Place channel in trip.

AND D.1.2 Reduce THERMAL POWER to

• 75% RTP.

OR D.2.1 Place channel in trip.

AND

---

NOTE-------

Only required to be performed when the Power Range Neutron Flux input to QPTR is inoperable.

D.2.2 Perform SR 3.2.4.2.

OR D.3 Be in MODE 3.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 78 hours 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 78 hours Amendments 231/212, 04/02/02 ACTIONS North Anna Units 1 and 2 3.3.1-2

RTS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E.

One channel NOTE---------

inoperable.

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

E.1 Place channel in trip.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR E.2 Be in MODE 3.

78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> F.

One Intermediate Range F.1 Reduce THERMAL POWER 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Neutron Flux channel to < P-6.

inoperable.

OR F.2 Increase THERMAL POWER 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to > P-10.

G.

Two Intermediate Range ------------ NOTE---------

Neutron Flux channels Limited plant cooldown or inoperable, boron dilution is allowed provided the change is accounted for in the calculated SDM.

G.1 Suspend operations Immediately involving positive reactivity additions.

AND G.2 Reduce THERMAL POWER 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to < P-6.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-3

RTS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME H.

One Source Range

---

NOTE----------

Neutron Flux channel Limited plant cooldown or inoperable, boron dilution is allowed provided the change is accounted for in the calculated SDM.

H.1 Suspend operations Immediately involving positive reactivity additions.

I.

Two Source Range I.1 Open Reactor Trip Immediately Neutron Flux channels Breakers (RTBs).

inoperable.

J.

One Source Range J.1 Restore channel to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> Neutron Flux channel OPERABLE status.

inoperable.

OR J.2.1 Initiate action to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> fully insert all rods.

AND J.2.2 Place the Rod Control 49 hours5.671296e-4 days <br />0.0136 hours <br />8.101852e-5 weeks <br />1.86445e-5 months <br /> System in a condition incapable of rod withdrawal.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-4

RTS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME K.

Required Source Range NOTE---------

Neutron Flux channel Plant temperature changes are inoperable, allowed provided the temperature change is accounted for in the calculated SDM.

K.1 Suspend operations Immediately involving positive reactivity additions.

AND K.2 Perform SR 3.1.1.1.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> AND Once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter L.

One channel NOTE---------

inoperable.

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

L.1 Place channel in trip.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR L.2 Reduce THERMAL POWER 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> to < P-7.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-5

RTS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME M.

One Reactor Coolant NOTE---------

Pump Breaker Position The inoperable channel may be channel inoperable, bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

M.1 Restore channel to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OPERABLE status.

OR M.2 Reduce THERMAL POWER 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> to < P-7.

N.

One Turbine Trip NOTE---------

channel inoperable.

The inoperable channel may be bypassed for up to 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for surveillance testing of other channels.

N.1 Place channel in trip.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> OR N.2 Reduce THERMAL POWER 76 hours8.796296e-4 days <br />0.0211 hours <br />1.256614e-4 weeks <br />2.8918e-5 months <br /> to < P-8.

0.

One train inoperable.

NOTE---------

One train may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing provided the other train is OPERABLE.

0.1 Restore train to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> OPERABLE status.

OR 0.2 Be in MODE 3.

30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-6

RTS Instrumentation 3.3.1 CONDITION REQUIRED ACTION COMPLETION TIME P.

One RTB train inoperable.

---

NOTES--------

1. One train may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing, provided the other train is OPERABLE.

2. One RTB may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for maintenance on undervoltage or shunt trip mechanisms, provided the other train is OPERABLE.

3.

One RTB train may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for concurrent surveillance testing of the RTB and automatic trip logic, provided the other train is OPERABLE.

P.1 Restore train to OPERABLE status.

OR P.2 Be in MODE 3.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 7 hours Q. One or more channels Q.1 Verify interlock is in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, required state for existing unit conditions.

OR Q.2 Be in MODE 3.

7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> Amendments 231/212, 04/02/02 ACTIONS North Anna Units 1 and 2 3.3.1-7

RTS Instrumentation 3.3.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME R.

One or more channels R.1 Verify interlock is in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> inoperable, required state for existing unit conditions.

OR R.2 Be in MODE 2.

7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> S.

One trip mechanism S.1 Restore inoperable 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> inoperable for one trip mechanism to RTB.

OPERABLE status.

OR S.2 Be in MODE 3.

54 hours6.25e-4 days <br />0.015 hours <br />8.928571e-5 weeks <br />2.0547e-5 months <br /> SURVEILLANCE REQUIREMENTS NOTE ----------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.1.2

---

NOTE----------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER is Ž 15% RTP.

Compare results of calorimetric heat 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> balance calculation to power range channel output. Adjust power range output if calorimetric heat balance calculation result exceeds power range channel output by more than +2% RTP.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-8

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.3

---

NOTE---------------

Not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after THERMAL POWER is Ž 15% RTP.

Compare results of the incore detector 31 effective measurements to Nuclear Instrumentation full power days System (NIS)

AFD. Adjust NIS channel if (EFPD) absolute difference is Ž 3%.

SR 3.3.1.4

---

NOTE---------------

This Surveillance must be performed on the reactor trip bypass breaker immediately after placing the bypass breaker in service.

Perform TADOT.

31 days on a STAGGERED TEST BASIS SR 3.3.1.5 Perform ACTUATION LOGIC TEST.

31 days on a STAGGERED TEST BASIS SR 3.3.1.6

---

NOTE ----------------

Verification of setpoint is not required.

Perform TADOT.

92 days SR 3.3.1.7

---

NOTE --------------

Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after entry into MODE 3.

Perform COT.

92 days Amendments 231/212, 04/02/02 North Anna Units I and 2 3.3.1-9

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

---

NOTE----------------

This Surveillance shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.

Perform COT.

SR 3.3.1.9

---

NOTES---------------

1. Adjust NIS channel if absolute difference Ž 3%.

2.

Not required to be performed until 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after THERMAL POWER is

Ž 50% RTP.

Compare results of the excore channels to incore detector measurements.

FREQUENCY SR 3.3.1.8

-I- _____________

92 EFPD Amendments 231/212, 04/02/02 NOTE ----

Only required when not performed within previous 92 days Prior to reactor startup AND Four hours after reducing power below P-6 for source range instrumentation AND Twelve hours after reducing power below P-10 for power and intermediate range instrumentation AND Once per 92 days thereafter SURVEILLANCE North Anna Units 1 and 2 3.3.1-10

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.10 ------------------- NOTE----------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

18 months SR 3.3.1.11 ------------------

NOTE ----------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform--CHANNEL-CALIBRATION 18-months Perform CHANNEL CALIBRATION.

18 months SR 3.3.1.12 Perform CHANNEL CALIBRATION.

18 months SR 3.3.1.13 Perform COT.

18 months SR 3.3.1.14 ------------------- NOTE----------------

Verification of setpoint is not required.

Perform TADOT.

18 months SR 3.3.1.15

---

NOTE----------------

Verification of setpoint is not required.

Perform TADOT.

Prior to exceeding the P-8 interlock whenever the unit has been in MODE 3, if not performed within the previous 31 days Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-11

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.1.16 NOTE---------------

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits.

18 months on a STAGGERED TEST BASIS Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-12

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 5)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

1. Manual Reactor Trip 1, 2 2

B SR 3.3.1.14 NA 3 (a),

4(a),

5(a) 2 C

SR 3.3.1.14 NA

2. Power Range Neutron Flux

a. High 1, 2 4

D SR 3.3.1.1

<110%, RTP SR 3.3.1.2 SR 3.3.1.3 SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16

b.

Low 1(b),

2 4

E SR 3.3.1.1

_ 26% RTP SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux Rate

a. High Positive Rate 1, 2 4

E SR 3.3.1.7

_< 5.5% RTP SR 3.3.1.11 with time constant

>_ 2 sec

b.

High Negative Rate 1, 2 4

E SR 3.3.1.7

_< 5.5% RTP SR 3.3.1.11 with time SR 3.3.1.16 constant

_ 2 sec

4. Intermediate Range Neutron Flux I(b),

2 (c) 2 F, G SR 3.3.1.1

_ 40% RTP SR 3.3.1.8 SR 3.3.1.11

5. Source Range Neutron Flux 2 (d) 2 H, I SR 3.3.1.1

_ 1.3 E5 cps SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16 3 (a) 4(a),

5(a) 2 I, J SR 3.3.1.1

_ 1.3 E5 cps SR 3.3.1.7 SR 3.3.1.11 SR 3.3.1.16 3(e) 4 (e),

5(e)

I K

SR 3.3.1.1 NA SR 3.3.1.11 With Rod Control System capable of rod withdrawal or one or more Below the P-10 (Power Range Neutron Flux) interlocks.

Above the P-6 (Intermediate Range Neutron Flux) interlocks.

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

With the Rod Control System incapable of rod withdrawal.

In this not provide reactor trip but does provide indication.

rods not fully inserted.

condition, source range Function does Amendments 231/212, 04/02/02 (a)

(b)

(c)

(d)

(e)

North Anna Units 1 and 2 3.3.1-13

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 5)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

6. Overtemperature AT 1, 2 3

E SR 3.3.1.1 Refer to SR 3.3.1.3 Note 1 (Page SR 3.3.1.7 3.3.1-16)

SR 3.3.1.9 SR 3.3.1.12 SR 3.3.1.16

7. Overpower AT 1, 2 3

E SR 3.3.1.1 Refer to SR 3.3.1.7 Note 2 (Page SR 3.3.1.12 3.3.1-17)

8. Pressurizer Pressure

a.

Low 1 (f) 3 L

SR 3.3.1.1 2! 1860 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

b. High 1, 2 3

E SR 3.3.1.1

< 2370 psig SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

9. Pressurizer Water Level-High

)3 L

SR 3.3.1.1

- 93%

SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

10. Reactor Coolant Flow-Low 1 (f) 3 per L

SR 3.3.1.1 2! 89%

loop SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

11. Reactor Coolant Pump (RCP) 1(f) 1 per M

SR 3.3.1.14 NA Breaker Position RCP

12.

Undervoltage RCPs I per L

SR 3.3.1.6

> 2870 V bus SR 3.3.1.10 SR 3.3.1.16

13. Underfrequency RCPs 1(f) 1 per L

SR 3.3.1.6(g)

>_ 56 Hz bus SR 3.3.1.10 SR 3.3.1.16

14. Steam Generator (SG)

Water 1, 2 3 per SG E

SR 3.3.1.1

_ 17%

Level-Low Low SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 (f)

Above the P-7 (Low Power Reactor Trips Block) interlock.

(g) Required to be performed for Unit 2 only.

Amendments 231/212, 04/02/02 North Anna Units I and 2 3.3.1-14

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 5)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS ALLOWABLE VALUE

15.

SG Water Level-Low Coincident with Steam Flow/Feedwater Flow Mismatch

16. Turbine Trip

a. Low Auto Stop Oil Pressure

b. Turbine Stop Valve Closure

17. Safety Injection (SI)

Input from Engineered Safety Feature Actuation System (ESFAS)

18. Reactor Trip System Interlocks

a. Intermediate Range Neutron Flux, P-6

b.

Low Power Reactor Trips Block, P-7

c.

Power Range Neutron Flux, P-8

d. Power Range Neutron Flux, P-IO

e. Turbine Impulse Pressure, P-13

19. Reactor Trip Breakers(i)

20. Reactor Trip Breaker Undervoltage and Shunt Trip Mechanisms

21. Automatic Trip Logic 1, 2

1. 2 1 (h) 1 (h) 1, 2 2 (d) 1 1

1, 2 1

1, 2 3 (a),

4 (a),

5 (a) 1, 2 3(a),

4(a),

5 (a) 2 per SG 2 per SG 3

4 2 trains 2

1 per train 4

4 2

2 trains 2 trains 1 each per RTB 1 each per RTB 2 trains 2 trains E

SR SR SR E

SR SR SR N

SR SR N

SR SR 0

SR Q

SR SR R

SR R

SR SR Q

SR SR R

SR SR P

SR C

SR S

SR 3.3.1.1 3.3.1.7 3.3.1.10 3.3.1.1 3.3.1.7 3.3.1.10 3.3.1.10 3.3.1.15 3.3.1.10 3.3.1.15 3.3.1.14 3.3.1.11 3.3.1.13 3.3.1.5 3.3.1.11 3.3.1.13 3.3.1.11 3.3.1.13 3.3.1.10 3.3.1.13 3.3.1.4 3.3.1.4 3.3.1.4

Ž 24%

! 42.5% full steam flow at RTP

Ž 40 psig

Ž 0% open NA

Ž 3E-11 amp NA

• 31% RTP

Ž7% RTP

• 11% RTP

! 11% turbine power NA NA NA C

SR 3.3.1.4 NA 0

C SR SR 3.3.1.5 3.3.1.5 NA NA With Rod Control System capable of rod withdrawal or one or more Below the P-6 (Intermediate Range Neutron Flux) interlocks.

Above the P-8 (Power Range Neutron Flux) interlock.

rods not fully inserted.

(i)

Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.

Amendments 231/212, 04/02/02 (a)

(d)

(h)

North Anna Units 1 and 2 3.3.1-15

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 5)

Reactor Trip System Instrumentation Note 1: Overtemperature AT The Overtemperature AT Function Allowable Value shall not exceed the following nominal trip setpoint by more than 2.0% of AT span.

AT-<ATO K -K 2(1 +-s))[T-T']+K3P -P')-fl(AI)

Where:

AT is measured RCS AT, °F.

ATo is the indicated AT at RTP, OF.

s is the Laplace transform operator, sec-1.

T is the measured RCS average temperature, OF.

T' is the nominal Tavg at RTP,

_ [*]OF.

P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure,

_> [*] psig K, -< [*]

K2 - [*]/-F K3 _ [*]/psig t 1 Ž

[N] sec T2

- [*] sec f 1 (AI)

= [*]{[*]% -

(qt

- qb)}

when qt -qb

< [*]% RTP 0% of RTP when [*]% RTP _ qt - qbg [*]% RTP

[*]{(qt -

qb)

[*]}

when qt -

qb > [*]%Y RTP Where qt and qb are percent RTP in the upper and lower halves of the core, respectively, and qt + qb is the total THERMAL POWER in percent RTP.

The values denoted with [*] are specified in the COLR.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-16

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 5)

Reactor Trip System Instrumentation Note 2: OverpowerAT The Overpower AT Function Allowable Value shall nominal trip setpoint by more than 2% of AT span.

not exceed the following AT_< ATO K4 - K51 I3sjT-K6 [T-T']- f 2(AI)

Where:

AT is measured RCS AT,

°F.

AT0 is the indicated AT at RTP, OF.

s is the Laplace transform operator, sec 1.

T is the measured RCS average temperature, OF.

T' is the nominal Tavg at RTP,

_ [*]°F.

K4 < [*]

K5 _> [*]/OF for increasing Tavg

[*]/OF for decreasing Tavg K6 _ [*]/°F when T > T'

[*]/°F when T _< T' 3 <- [*] sec f 2 (AI)

= [*]

The values denoted with [*] are specified in the COLR.

Amendments 231/212, 04/02/02 North Anna Units 1 and 2 3.3.1-17

Intentionally Blank