TSTF-10-12, Transmittal of TSTF-525, Revision 0, Post Accident Monitoring Instrumentation Requirements (WCAP-15981-NP-A)

Transmittal of TSTF-525, Revision 0, Post Accident Monitoring Instrumentation Requirements (WCAP-15981-NP-A)

ML102510337
Person / Time
Site:	Technical Specifications Task Force
Issue date:	08/30/2010
From:	Gambrell R, Gregoire D, Raidy T, Schrader K B & W Owners Group, BWR Owners Group, Combustion Engineering Owners Group, PWR Owners Group, Technical Specifications Task Force, Westinghouse
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TSTF-10-12, Rev 0, WCAP-15981-NP-A
Download: ML102510337 (103)

Text

TECHNICAL SPECIFICATIONS TASK FORCE TSTF A JO-I-ZVT " J/VJERS GR O9L"P A CTI VITY August 30, 2010 TSTF-10-12 PROJ0753 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Transmittal of TSTF-525, Revision 0, "Post Accident Monitoring Instrumentation Requirements (WCAP- 15981 -NP-A)"

Enclosed for NRC review is TSTF-525, "Post Accident Monitoring Instrumentation Requirements (WCAP- 15981 -NP-A)."

Any NRC review fees associated with the review of TSTF-525 should be billed to the Pressurized Water Reactor Owners Group.

Should you have any questions, please do not hesitate to contact us.

Kenneth J. Schrader (PWROG/W) Donald W. Gregoire (YWROG)

Thomas W. Raidy (PWROG/CE) Re'en 'Gambrell (PWROG/B&W)

Enclosure cc: Robert Elliott, Technical Specifications Branch, NRC Barry Miller, Licensing Processes Branch, NRC 11921 Rockville Pike, Suite 100, Rockville, MD 20852 PWROG Phone: 301-984-4400, Fax: 301-984-7600 "* OWNERS' GROUP Administration by EXCEL Services Corporation 0`4o-s

WOG-203, Rev. 0 TSTF-525, Rev. 0 Technical Specification Task Force Improved Standard Technical Specifications Change Traveler Post Accident Monitoring instrumentation Requirements (WCAP- 15981-NP-A)

NUREGs Affected: E] 1430 W 1431 E] 1432 FD 1433 D1 1434 Classification 1) Technical Change Recommended for CLIIP?: No Correction or Improvement: Improvement NRC Fee Status: Not Exempt Benefit: Increases Equipment Operability See attached justification.

Revision History OG Revision 0 Revision Status: Active Revision Proposed by: PWROG Revision

Description:

Original Issue Owners Group Review Information Date Originated'by OG: 23-Jun-10 Owners Group Comments (No Comments)

Owners Group Resolution: Approved Date: 20-Jul-10 TSTF Review Information TSTF Received Date: 20-Jul-10 Date Distributed for Review 20-Jul-10 OG Review Completed: W BWOG F] WOG [] CEOG [] BWROG TSTF Comments:

Ken Schrader provided comments. Incorporated into final revision.

TSTF Resolution: Approved Date: 30-Aug-10 NRC Review Information NRC Received Date: 30-Aug-10 Affected Technical Specifications S/A 3.3.3 Bases PAM Instrumentation NUREG(s)- 1430 Only 30-Aug-JO Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

WOG-203, Rev. 0 TSTF-525, Rev. 0 Bkgnd 3.3.3 Bases PAM Instrumentation LCO 3.3.3 PAM Instrumentation Change

Description:

Table 3.3.3-1 LCO 3.3.3 Bases PAM Instrumentation Appl. 3.3.3 PAM Instrumentation Appl. 3.3.3 Bases PAM Instrumentation SR 3.3.3 PAM Instrumentation Change

Description:

SR Note Ref. 3.3.3 Bases PAM Instrumentation Action 3.3.3.A Bases PAM Instrumentation Action 3.3.3.C Bases PAM Instrumentation Action 3.3.3.D PAM Instrumentation Action 3.3.3.D Bases PAM Instrumentation Action 3.3.3.E PAM Instrumentation Action 3.3.3.E Bases PAM Instrumentation Action 3.3.3.F PAM Instrumentation Change

Description:

Deleted Action 3.3.3.F Bases PAM Instrumentation Change

Description:

Deleted SIR 3.3.3.3 PAM Instrumentation Change

Description:

New SIR 30-Aug-10 Traveler Rev. 3. Copyright(C) 2010, EXCEL Services Corporation. Use by EXCEL Services associates, utility clients, and the U.S. Nuclear Regulatory Commission is granted. All other use without written permission is prohibited.

TSTF-525, Rev. 0 1.0 Description The proposed change revises Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," to provide an alternate method to determine the appropriate PAM instrumentation for inclusion in Table 3.3.3-1 of Technical Specification 3.3.3 and to add the use of alternate monitoring instruments for certain PAM instrument Functions.

In addition, a new Surveillance Requirement is added to Technical Specification 3.3.3.

Table 3.3.3-1 in Technical Specification 3.3.3 is revised, as well as the associated Actions and Surveillance Requirements.

The current PAM Technical Specification requires that all the plant instrumentation designated as Regulatory Guide 1.97, Revision 2 (Reference 1).Type A and all instrumentation designated as Regulatory Guide 1.97 Category 1 be included in Technical Specification 3.3.3. The current PAM Technical Specification only considers the Regulatory Guide 1.97 designation (Type A or Category 1) when determining the applicability of the Technical Specification inclusion criteria in 10 CFR 50.36(c)(2)(ii).

The alternate method for selecting PAM instrument Functions discussed in this traveler is the application of the guidance provided in WCAP-15981-NP-A, Revision 0, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants,"

(Reference 2). WCAP-15981-NP-A provides a methodology that considers the use of accident monitoring instrumentation in the Safety Analyses, PRA, Emergency Operating Procedures (EOPs), Severe Accident Management Guidance (SAMG) procedures, and Emergency Plan (E-Plan). The WCAP methodology is used to identify the PAM instrumentation important to safety (i.e., that monitors plant parameters that are the basis for important operator actions to bring the unit to a safe stable state in the event of an accident). The WCAP methodology also includes the evaluation of individual instrument to determine which criteria of 10 CFR 50.36(c)(2)(ii) each identified instrument Function satisfies to determine those which should be included in the Technical Specifications.

Therefore, the application of the WCAP-15981-NP-A methodology results in an enhanced technical basis for selecting individual PAM instrumentation for inclusion in the Technical Specifications that considers a broader scope of criteria, including PRA information and SAMG procedures which were not available when Regulatory Guide 1.97 was developed to evaluate accident monitoring instrumentation.

The current PAM Technical Specification only allows the use of alternate means of monitoring for two PAM Functions, Reactor Vessel Water Level and Containment Area Radiation (High Range). WCAP-15981-NP-A identifies several additional PAM Functions that have. a valid alternate means of monitoring. Although in Section 4.0 of the SE for WCAP-15981-NP-A (Ref. 2) the NRC did not approve the use of alternate -

instrumentation on a generic basis, the NRC stated that the use of alternate instrumentation should continue to be reviewed on a plant specific basis. Therefore, additional justification for the use of the alternate PAM instrumentation identified in WCAP-15981-NP-A is provided to support plant specific implementation of alternate instrumentation.

Page 1 of 28

TSTF-525, Rev. 0 WCAP-15981-NP-A also proposed that the number of required Core Exit Thermocouples (CETs) specified in Table 3.3.3-1 be revised from two channels per quadrant to two channels (total) with a Note specifying the acceptable core locations of the required channels. The proposed change in the number of required CETs was not approved by the NRC in the SE based on WCAP-15981-NP-A' However, this traveler provides additional information beyond what was included in the WCAP to justify the change in the required number of CETs.

The current PAM Technical Specification is applicable in Modes 1, 2, and 3 for all Functions listed on Table 3.3.3-1. The proposed changes include the addition of a Note to provide an exception for the Operability of the Power Range Neutron Flux PAM Function in Mode 3. The proposed change is consistent with the discussion of this PAM function in WCAP-15981-NP-A andwith the Applicability of the Power Range Neutron Flux instrumentation in Technical Specification 3.3.1, "Reactor Trip Instrumentation."

In addition to the changes addressed in WCAP-15981-NP-A, the following changes are included:

" In order to support the use of the proposed alternate monitoring instrumentation that were discussed in WCAP-15981 -NP-A, the Actions are simplified and the reference to Actions in Table 3.3.3-1 is deleted.

• The current PAM Technical Specification requires a Channel Calibration be performed on all PAM instrument Functions including the Penetration Flow path Containment Isolation Valve (CIV) Position Function. In order to provide a more appropriate Surveillance Requirement for the Penetration Flow path CIV Position indication, a Trip. Actuating Device Operational Test (TADOT) is added to Technical Specification 3.3.3 for this PAM Function.

2.0 Proposed Changes I) The current PAM Technical Specification Table 3.3.3-1 contains three columns.

The table columns contain the PAM Functions, the required channels for each Function, and the Condition referenced from Required Action D. I for each Function in the Table. The proposed change contains only two columns. The first column contains a revised list of PAM Functions based on the application of the methodology in WCAP-15981-NP-A and the second column specifies the required number of channels for each Function. The third column, referencing the Required Action Condition is deleted.

2) The Actions in the current PAM Technical Specification that may be applicable when a Function or Functions have two inoperable channels are Action Conditions C, D, and E. Action Condition C addresses two required channels inoperable and requires restoration of one channel to operable status in 7 days. If the Required Action andassociated Completion Time of Condition C are not met, Condition D directs that the Condition referenced (for the affected PAM Function) in Table 3.3.3-1 be entered. Table 3.3.3-1 lists eitherCondition E or Page 2 of 28

TSTF-525, Rev. 0 Condition F for each PAM Function listed on the table. Condition E requires a unit shutdown and Condition F allows continued operation but specifies that action be initiated immediately in accordance with Specification 5.6.5, "Post Accident Monitoring Report." Specification 5.6.5 requiresthat a report be submitted to the NRC within 14 days outlining the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to operable status.

The proposed change simplifies the Actions in the current PAM Technical Specification associated with two required channels inoperable. Condition D is revised with a Note such that the Condition only applies to those PAM Functions in Table 3.3.3-1 that have identified preplanned alternate methods of monitoring.

If applicable, the Bases for the PAM Function describe the acceptable alternate means of monitoring consistent with the guidance in WCAP-15981-NP-A. As such, if the Required Action and associated Completion Time of Condition C (for two inoperable channels) are not met), the proposed Condition D provides the Action for those Functions with preplanned alternate methods of monitoring without the need to reference Condition F from Table 3.3.3-1. The Required Action for Condition D is replaced with the Required Action F.I. Required Action F.1 specifies action be initiated immediately in accordance with Specification 5.6.5. Thus, Condition D effectively replaces the current combination of Conditions D and F.

Condition E in the current PAM Technical Specification is also revised to state "Required Action and associated Completion Time of Condition C not met" and Condition E is modified by a Note that specifies the applicability of the Condition to those PAM Functions not addressed by Condition D. The Required Actions of proposed Condition E remain the same (i.e., to be in Mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). As such, if the Required Action and associated Completion Time of Condition C (for two inoperable channels) are not met, the proposed Condition E provides the appropriate Action for those Functions without preplanned alternate methods of monitoring defined in the Bases. Thus, ,the proposed Condition E effectively replaces the combined Conditions D and E of the current PAM Technical Specification.

The revisions to Conditions D and E effectively provide the same guidance as current PAM Conditions D, E and F. The difference between the current and proposed Actions is that it is no longer necessary to reference separate Actions for each Function on Table 3.3.3.-1. In addition, with proposed Conditions D and E as described above, current Condition F is no longer necessary and is deleted.

The proposed changes to Conditions D, E, and F described above also affect Specification 5.6.5. This is due to the proposed new Condition D which references Specification 5.6.5. Each PAM Condition that references Specification 5.6.5 is listed in Specification 5.6.5. Therefore,. Specification 5.6.5 is revised to reference Condition D.

Page 3 of 28

TSTF-525, Rev. 0

3) The current PAM Technical Specification is applicable in Modes 1, 2, and 3 for all Functions listed in Table 3.3.3-1. One of the proposed changes includes the addition of a Note that provides an exception for the Operability of the Power Range Neutron Flux PAM Function in Mode 3.

4) The current PAM Technical Specification requires a Channel Check to be performed on each PAM Function every 31 days (SR 3.3.3.1) and a Channel Calibration be performed on each PAM instrument every 18 months (SR 3.3.3.2).

A Note to the SRs states that SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrument Function in Table 3.3.3-1. The proposed changes include the addition of a new Surveillance Requirement to the PAM Technical Specification. The new Surveillance (SR 3.3.3.3) requires the performance of a TADOT every 18 months.

In addition, the Surveillance Requirement Note is revised such that the Channel Calibration Surveillance Requirement is applicable to all PAM Functions except the Penetration Flow paih CIV Position and that the new TADOT Surveillance Requirement is only applicable to the Penetration Flow path CIV Position PAM Function.

5) The current PAM Technical Specification allows the use of alternate indications for the Containment Area Radiation monitors (which typically include the use of one or more portable radiation monitors as an alternate) and the Reactor Vessel Water Level Indication System. The proposed change would include the provision to use acceptable alternate indications for the following additional PAM Functions:

• Power Range Neutron Flux

• High Head Safety Injection Flow

• SG Water Level (Wide Range)

• Auxiliary Feedwater Flow The use of alternate PAM indications has been incorporated in the proposed PAM Technical Specification and Bases. The provision to use alternate PAM instrumentation is discussed in the Technical Specification Bases for each of the affected PAM Functions. The individual PAM Functions that would allow the, use of alternate indications are also identified in the proposed Note modifying Condition D.

6) The current PAM Technical Specification requires two channels of CETs per core quadrant. The two required channels in each quadrant specified in Table 3.3.3-1 are modified by Note (c) which specifies that a channel consists of two CETs.

This results in eight required channels with a total of sixteen required CETs. The proposed change would reduce the number of required CET channels specified in Table 3.3.3-1 to a total of two. Table 3.3.3-1 Note (c) would be revised to specify that a channel consists of two CETs in the nine central core rows and columns.

Page 4 of 28

TSTF-525, Rev. 0 The proposed change excludes all CETs in the three outer rows of the core and does not include any requirements regarding core quadrants. Thus the proposed change would result in two required channels with a total of four required CETs in the core locations specified in the proposed Note.

Due to the extensive nature of the changes, a retyped version of the revised TS and Bases is included to Facilitate the review.

3.0 Background The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during an accident. This information provides the necessary support for the control room operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions during Design Basis Accidents (DBAs).

The PAM instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and assess the unit status and behavior following an accident.

The availability of the PAM instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. These essential instruments were originally identified by addressing the recommendations of Regulatory Guide 1.97 as required by Supplement I to NUREG-0737 (Reference-4)..

Regulatory Guide 1.97 Type A variables provide the primary information required for the control room operator to take specific manual actions for which no automatic control is provided, and that are required for safety systems to accomplish their safety functions as assumed in the DBA analyses.

In addition to Type A variables, Regulatory Guide 1.97 identified Category 1 variables as significant to safety. Regulatory Guide 1.97 Category I variables were provided to determine whether other systems important to safety are performing their intended functions.

Typically, Regulatory Guide 1.97 Type A variables are also Category I variables.

However, not all Category I variables are also classified as Type A.

Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation,"

specifies (in a reviewers note) that a plant should include all of their Regulatory Guide 1.97 Type A and all of their Regulatory Guide 1.97 Category I *instrumentation in the PAM Technical Specification. The list of generic PAM Functions identified in Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," in NUREG-1431 was developed in the late 1980's based on design basis accident requirements and generic insights from Probabilistic Risk Assessments (PRA) available at that time.

The PAM instrumentation was included in the Technical Specifications to ensure that the instrumentation required by the operators to respond to an accident and bring the plant to Page 5 of 28

TSTF-525, Rev. 0 a safe stable state is operable if required during an accident. The inclusion of PAM instrumentation functions in NUREG- 1431 was determined based on the Technical Specification Criteria in 10 CFR 50.36 (c)(2)(ii).

The four Criteria for determining Technical Specification content were codified in the Federal Regulations by an amendment to 10 CFR 50.36 on July 19, 1995 (60 CFR 36953). The 10 CFR 50.36 (c)(2)(ii) 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 risk assessment has shown to be significant to public health and safety.

The fourth Criterion reflects the insights obtained from PRA studies. As discussed below, the PAM instrumentation contained in Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," in NUREG- 1431 is based primarily on the first three Criteria of 10 CFR 50.36(c)(2)(ii). Insights from PRA studies were not widely known, or available when Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," was issued in Revision 0 ofNUREG-1431 in September 1992.

The original basis for determining the instrumentation to be included in Technical Specification 3.3.3, "PAM Instrumentation," contained in NUREG-1431 is defined in WCAP-1 1618 (Reference 5). WCAP-1 1618 was submitted to the NRC in November 1987, and identified certain PAM Instrumentation that satisfied 10 CFR 50.36(c)(2)(ii)

Criterion 3. The justification discussed in WCAP-1 1618 for satisfying Criterion 3 for the PAM Tech Spec is as follows:

"Specific Accident Monitoring Instrumentation provides the operator with the information needed to perform the required manual actions to bring the plant to a stable condition following an accident. This instrumentation is part of the primary success path and which functions or actuates to mitigate a DBA or Transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. Specific Accident Monitoring Instrumentation satisfies criterion 3."

Page 6 of 28

TSTF-525, Rev. 0 Therefore, WCAP-1 1618 limited the content of the proposed NUREG-1431 PAM Technical Specification to Regulatory Guide 1.97 Type A instruments. Non Type A Category I instrumentation was not identified as satisfying any of the criteria for inclusion in the Technical Specifications.

The NRC letter to the Owners Groups (Reference 6), which documented the review of WCAP- 11618, stated that PAM Instrumentation satisfies the definition of Type A variables in Regulatory Guide 1.97, and meets Criterion 3. The NRC justification for retaining Type A variables states: "Type A variables provide primary information (i.e.,

information that is essential for the direct accomplishment of the specified manual actions (including long-term recovery actions) for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for DBAs or transients)." It also discusses that since only Type A variables meet Criterion 3, the Standard Technical Specifications should contain a narrative statement that indicates that individual plant Tech Specs should contain a list of PAM Instrumentation that includes Type A variables.

However, regarding the non-Type A Category I variables, the NRC stated in Reference 6,: "the staff is unable to confirm the Owners Groups' conclusion that Category I Post-Accident Monitoring Instrumentation'is not of prime importance in limiting risk

[Criterion 4]. Recent risk assessments have shown the risk significance of operator recovery actions which would require knowledge of Category I variables. Furthermore, recent severe accident studies have shown significant potential for risk reduction from accident management. The Owners Groups' should develop further risk-based justification in support of relocating any or all Category 1 variables from the Standard Technical Specifications." The Owners Groups' participating in the development of NUREG-1431 chose not to evaluate the inclusion of Regulatory Guide 1.97 Non Type A, Category I instrumentation in the PAM Technical Specification at that time. Therefore, the NUREG- 1431 PAM Technical Specification was issued with the requirement that all plant specific Regulatory Guide 1.97 Type A, and all plant specific Regulatory Guide 1.97 Category I instrumentation be included in the PAM Technical Specification.

WCAP-1 5981-NP-A was developed to specifically address the NRC request to further evaluate the inclusion of Regulatory Guide 1.97 Category 1 variables in the PAM Technical Specification. In addition, WCAP-15981-NP-A provides a generic methodology for developing a technical basis for relocating certain Post Accident Monitoring instruments from the Technical Specifications. The conclusions contained in WCAP-15981-NP-A are based on generic risk insights (i.e., evaluations against 10 CFR 50.36 (c)(2)(ii) Criterion 4) and a re-evaluation of the overall basis for Accident Monitoring instrumentation with respect to the first three Criteria of 10 CFR 50.36(c)(2)(ii). WCAP-15981-NP-A also includes the consideration of the reliance on the instrumentation not specifically evaluated when the list of PAM instrumentation was originally developed for NUREG-1431. These additional considerations include instrumentation required to mitigate the consequences of beyond design basis accidents, such as those that are important for Severe Accident Management (e.g., SAMG), and offsite emergency radiological protection actions (e.g., Emergency Action Level declarations and Offsite Dose Calculations). A plant specific confirmation, Page 7 of 28

TSTF-525, Rev. 0 using the methodology contained in WCAP-15981-NP-A is required to ensure that the generic conclusions are applicable to each plant implementing the methodology in the WCAP.

Regulatory Guide 1.97 provides guidance on the classification of plant parameters (that are indicated by instrumentation) according to their importance. There are different classes of variables identified in Regulatory Guide 1.97 according to the type of information that is provided by that variable. Type A variables provide primary information needed to permit control room operating personnel to take the specified manually controlled actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for design basis events.

Regulatory Guide 1.97 does not include a listing of Type A variables, since they are to be determined on a plant specific basis. However, Regulatory Guide 1.97 includes a list of Category I variables. The instrumentation associated with those Category I variables is identified in Table I below:

[ Table 1: Regulatory Guide 1.97 Category 1 Variables Power Range Neutron Flux Hydrogen Monitors Source Range Neutron Flux Pressurizer Level Reactor Coolant System Hot Leg Steam Generator Water Level (Wide Temperature Range)

Reactor Coolant System Cold Leg Condensate Storage Tank Level Temperature Reactor Coolant System Pressure (Wide Core Exit Temperature - Quadrant [I]

Range)

Reactor Vessel Water Level Core Exit Temperature - Quadrant [2]

Containment Sump Water Level (Wide Core Exit Temperature -- Quadrant [3]

Range)

Containment Pressure (Wide Range) Core Exit Temperature - Quadrant [4]

Containment Isolation Valve Position RCS Radiation (no instrumentation available for direct measurement)

Containment Area Radiation (High Range)

With the exception of the Reactor Coolant Systems (RCS) Radiation indication, these instruments are included in the Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," in NUREG-1431 based on the NRC's 1987 conclusion contained in Reference 6 that these instruments may be important in limiting risk, based on a limited perspective of available PRA results. The results of the assessment contained in WCAP-15981-NP-A show that some of the Category I non-Type A variables are not important in limiting risk as indicated from a wide-reaching survey of Westinghouse plant PRA results. Thus, WCAP-15981-NP-A recommends that only instrumentation that satisfies Criterion 3 or 4 of 10 CFR 50.36 (c)(2)(ii) be included in Page 8 of 28

TSTF-525, Rev. 0 the Technical Specifications.

A similar re-evaluation of the basis and classification of the hydrogen monitor PAM Function was performed in the rulemaking to revise 10 CFR 50.44, where the NRC determined that the hydrogen monitors no longer met the definition of Category I in Regulatory Guide 1.97. The NRC concluded that Category 3 as defined in Regulatory Guide 1.97 is an appropriate categorization for the hydrogen monitors because they are only required to diagnose the course of beyond design basis accidents.

The NRC approved the elimination of the Post Accident Sampling System (PASS) on June 14, 2000. The PASS requirements were based on the knowledge of severe accidents shortly after the accident at Three Mile Island Unit 2 in 1979. The justification for eliminating the PASS was based on a better understanding of severe accidents due to significant research and analysis after the requirements for the PASS were developed.

The basis for the NRC approval for PASS elimination is contained in WCAP-14986-A, Revision 2 (Reference 7). Eliminating the PASS was based on the accident progression as implemented in the Abnormal and Emergency Operating Procedures, Severe Accident Management Guidelines, Core Damage Assessment Guidelines, Emergency Plan, and Emergency Plan Implementing Procedures. The WCAP-15981-NP-A methodology utilizes a similar approach to evaluate the required PAM instruments based on these procedures.

The proposed changes in the PAM instrumentation Functions in NUREG-1431, Technical Specification 3.3.3, Table 3.3.3-1 have been evaluated in accordance with the screening criteria contained in WCAP-15981-NP-A. The screening criteria were used to identify the PAM instrumentation important to safety (i.e., monitor plant parameters that are the basis for important operator actions to bring the unit to a safe stable state in the event of an accident). The selected instrument Functions satisfy Criterion 3 and/or 4 of 10 CFR 50.36(c)(2)(ii), and include Regulatory Guide 1.97 monitoring instrumentation for parameters identified as important to safety in accordance with the methodology contained in WCAP-15981-NP-A.

The PAM instrumentation selected in accordance with the methodology in WCAP-I15981-NP-A provides the capability to monitor plant parameters necessary for safety significant operator actions so that the control room operating staff can:

• Perform the diagnosis specified in the EOPs (these variables are restricted to preplanned actions for the primary success path of DBAs), e.g., loss of coolant accident (LOCA),

• Take the specified, pre-planned, manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety function,

• Implement procedures or guidance that has been shown to have an important role in preventing core damage or early fission product releases, Page 9 of 28

TSTF-525, Rev. 0

" Determine the likelihood of a gross breach of the barriers that prevent radioactivity release,

" Determine if a gross breach of a barrier has occurred, and

" Initiate action necessary to protect the public and to estimate the magnitude of any impending threat.

4.0 Technical Analysis Change 1 The revisions made to PAM Technical Specification Table 3.3.3-1.

The proposed change is the revision of the instrument Functions listed in PAM Technical Specification Table 3.3.3-1. The revised list of Functions represents what a typical Westinghouse plant may include in the PAM Technical Specification after a plant specific application of the WCAP-15981-NP-A methodology. As such, proposed Table 3.3.3-1 serves the same purpose as current Table 3.3.3-1, which contains a "typical" list of Type A and Category I PAM instrument Functions. Plant specific versions of the current Table 3.3.3-1 are expected to vary somewhat from the typical list provided in NUREG-1431. The technical justification for the content of proposed Table 3.3.3-1 (selected instrument Functions and number of required channels) is contained in WCAP-15981 -NP-A.

The remaining change to current PAM Technical Specification Table 3.3.3-1, is the deletion of the third column of the Table. The third column of Table 3.3.3-1 lists the appropriate Condition for each PAM Function referenced from Required Action D. 1.

Condition D is a "steering" Condition that is applicable when the Required Action and Associated Completion Time of Condition C (two required channels inoperable) are not met. Condition D directs the user to the Table whichspecifies either Condition E or F for each instrument Function listed in the Table. If the affected Function is assigned Condition E, shutdown Actions are applicable. If the affected Function is assigned Condition F, continued operation is permitted in accordance with Specification 5.6.5. As discussed below in Change 2, Conditions D, E, and F are revised to accomplish the same function as before without the need to reference Table 3.3.3-1 for the appropriate Condition. Therefore, the elimination of the Conditions referenced in Table 3.3.3-1 represents an administrative change due to a different (simpler) method of presenting the same information to the user.

Change 2: The revision of Conditions D and E and the deletion of Condition F..

Current Conditions D, E, and F provide the appropriate Action when the Required Action and associated Completion Time of Action Condition C (for two inoperable channels) are not met. This was accomplished by Condition D which references Table 3.3.3-1 and the Table directs the user to either Condition E of Condition F. As previously discussed, the proposed change modifies Conditions D and E with Notes such that referencing Table 3.3.3-1 is no longer necessary to determine the applicable Condition for each PAM Function. The proposed Notes modifying Conditions D and E, and the deletion of Page 10 of 28

TSTF-525, Rev. 0 Condition F, simplify the Actions without changing the technical intent of the Actions.

Therefore, these changes are administrative that only affect the format and presentation

• ofthe PAM Technical Specification.

The revision of the Actions described above also includes moving Required Action F.I

.(initiate Action in accordance with Specification 5.6.5) to Required Action D.1. This change is part of the reformatting of the Conditions that allows proposed Condition D to replace the combination of Conditions D and F contained in the current PAM Technical Specification. The proposed Condition D provides the same Required Action as the combination of Conditions it replaces. The relocation of Required Action F. 1 to Action D.I also affects the content of Specification 5.6.5, "Post Accident Monitoring Report.

Specification 5.6.5 lists the Conditions that contain references to Specification 5.6.5.

Therefore, Specification 5.6.5 is revised to list Condition D as well as Conditions B and F. This change makes the list of Conditions in Specification 5.6.5 complete and is necessary due to the re-arrangement of Action Conditions D, E, and F discussed above.

As such, the relocation of Required Action F.1 to Required Action D. l and the addition of Condition D to Specification 5.6.5 do not represent technical changes.

The list of PAM Functions included in the proposed Note to Condition D defines the Functions for which an alternate monitoring method has been identified and justified based on WCAP-15981-NP-A and these proposed changes. The Bases for each of these PAM Functions describes the acceptable alternate method(s) available to monitor the Function. Similar to the list of Functions on Table 3.3.3-1, the proposed list of Functions with alternate monitoring method(s) is typical of what may be expected for Westinghouse plants, but may vary due to plant specific considerations when applying the methodology of WCAP-I15981-NP-A.

It should be noted that the list of PAM Functions that have alternate monitoring instrumentation are listed in the Note to Condition D (as discussed above) and as such, are part of the PAM Technical Specification. Therefore, the PAM instrumentation allowed to use alternate instrumentation continues to be subject to NRC review and approval.

Change 3: The addition of an LCO Note stating that the Power Range Neutron Flux PAM Function is not required to be operable in Mode 3.

The PAM Technical Specification is applicable in Modes 1, 2, and 3.

WCAP-15981-NP-A discusses that power range neutron flux is a key indication for accident management operator actions to initiate a manual reactor trip to bring the reactor to a subcritical condition. The key indication of the power range neutron flux is consistent with the keff > 0.99 specified for Mode 2 and for power operation in Mode 1.

Subsequent operator actions (in Mode 3 after a reactor trip) to assure that the reactor remains in a subcritical state, where the power range neutron flux monitor may be inoperable such as during RCS depressurization, were not determined to be important for long term core cooling. Therefore, for the required PAM indication function (i.e.,

confirming reactor trip from Modes I and 2) performed by the Power Range Neutron Flux monitor is only required to be operable in Modes I and 2. The proposed change Page I1 of 28

TSTF-525, Rev. 0 addresses this Mode of Applicability by the addition of a Note that excludes the requirement for the Power Range Neutron Flux PAM Function to be operable in Mode 3.

This proposed change makes the PAM Technical Specification for the Power Range Neutron Flux instrumentation consistent with the applicability for this instrumentation in Technical Specification 3.3.1, "Reactor Trip System Instrumentation."

The proposed change (which limits the applicability of the power range neutron flux indication to Modes I and 2) is acceptable because it is necessary to properly define the applicable Modes in which the power range neutron flux indication is required to perform its PAM function and to exclude Mode 3 where the power range instrumentation may not be operable and is not required for its key indication Function (i.e., confirming a reactor trip from Modes I and 2).

Change 4: The addition of a TADOT Surveillance Requirement for the Penetration Flow Path CIV Position PAM Function.

A new surveillance is added to the PAM Technical Specification. Proposed SR 3.3.3.3 specifies that a TADOT be performed every 18 months. The Surveillance Requirement Note is revised to describe that the new TADOT surveillance is only applicable to the Penetration Flow Path CIV Position PAM Function. The proposed new surveillance is necessary because the existing PAM Channel Calibration surveillance (which is intended for process monitoring instrument channels containing sensors, setpoints and signal processing) is not the appropriate surveillance to be performed on the containment isolation valve position indication channels. The TADOT is a defined Technical Specification surveillance more appropriate to the verification of-valve position indication. The TADOT does not contain requirements for "...adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to knownvalues of the parameter that the channel monitors" as does the Channel Calibration surveillance. The TADOT requires "...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". As such, the TADOT is the appropriate surveillance to verify the operation of the required valve position indication circuits and to assure that these required PAM indications are operable. The proposed change is acceptable because. it specifies the appropriate surveillance to ensure the continued operability of the required containment isolation valve position indication. It should be noted that this change was previously approved by the NRC in Amendment 278 to Renewed Facility Operating License No.

DPR-66 and Amendment 161 to Renewed Facility Operating License No. NPF-73 for Beaver Valley Power Station Units I & 2, respectively (see BVPS Unit 1 & 2 Technical Specification Surveillance Requirement 3.3.3.3).

Page 12 of 28

TSTF-525, Rev. 0 Change 5:, The addition of provisions to allow the use of alternate PAM indications for selected PAM Functions.

The following Table is taken from WCAP-1 5981-NP-A (Ref. 2) and summarizes the alternate PAM instrumentation proposed and discussed in the WCAP.

Table 13 Summary of Alternate PAM Instrumentation Primary Instrumentation Alternate Instrumentation SG Water Level (Wide Range) SG Narrow Range Level AND Auxiliary Feedwater Flow Rate Power Range Neutron Flux Intermediate or Source Range Indications AND either the Rod Position Indicators OR Rod Bottom Lights Containment Area Radiation (High Range) Portable Radiation Monitors High Head Safety Injection Flow High Head Safety Injection Pump Amperage AND SI Pump Discharge or Header Pressure AND Automatic SI valve position Auxiliary Feedwater Flow Motor Driven Pumps: Pump Amperage AND Pump Discharge Pressure OR flow control valve (SG supply) position Turbine Driven Pump: Pump Discharge Pressure OR steam supply valve position AND flow control valve (SG supply) position In Section 3.2.21, "Proposed Alternate Instrumentation," of the NRC SE for WCAP-15981-NP-A (Ref. 2) (note the SE is included in Reference 2), the NRC addressed the use of the alternate PAM instrumentation proposed in WCAP-15981-NP-A. In Section 3.2.21 of the SE the NRC stated:

"Alternate instrumentation should meet the same RG 1.97 category as the primary instrumentation. RG 1.97 recommends two channels of Category I instrumentation for each Type A or Category 1 variable.

TR WCAP-1598 1-NP recommends the use of alternate instrumentation for various PAM instrumentation."

In Sections 3.2.21.1 through 3.2.21.5 of the NRC SE, the NRC addressed the individual instrumentation proposed for use as alternate PAM instrumentation. In these SE sections the NRC stated that either:

• The WCAP did not discuss the qualification of the proposed alternate instrumentation, or Page 13 of 28

TSTF-525, Rev. 0

• The acceptability of the alternate instrumentation should be reviewed on a plant specific basis.

In Section 4.0, "Limitations and Conditions" of the NRC SE for Reference 2 the NRC concluded that:

"The NRC staff does not agree with the proposed use of alternate instrumentation on a generic basis. The use of instruments as alternates should continue to be reviewed on a plant-specific basis."

The use of pre-planned alternate instrumentation is not intended to change the design of the current PAM instrumentation; rather it is provided as an alternative to prevent a plant shutdown by utilizing other indications that are available to provide similar information during the time period until the required PAM instrumentation is restored to operable status. The "Post Accident Monitoring Report" required to be submitted to the NRC by Specification 5.6.5 will identify the preplanned alternate method of monitoring the function, the cause of the inoperability, and the plans and schedule for restoring the inoperable instrumentation to operable status. Therefore, the alternate instrumentation does not have to meet the same design and qualification criteria as the PAM instrumentation during this short period of time.

As discussed in the Background Document for the Westinghouse Emergency Response Guidelines (ERGs), other instrumentation is assumed to be available to permit the operator to:

a. Operate plant safety systems employed for mitigating the consequences of an accident and subsequent plant recovery to attain a safe shutdown condition, including verification of the automatic actuation of safety systems; and

b. Operate other systems normally employed for attaining a safe shutdown condition.

The instrumentation that provides this plant information to the operator is typically not required to meet the same design, qualification and display requirements as the key instrumentation. For example, this instrumentation is not required to be redundant, it may be powered only from a highly reliable power source (battery backed), and is not needed to be either accessible on demand or recorded.

The following generic discussion would be utilized to develop plant specific justifications for the use of alternate PAM Instrumentation. The proposed model application requires the licensee to providethe plant-specific justification for the use of alternate PAM instruments.

The following generic discussion contains information regarding plant procedures and design that would be used to justify the use of alternate PAM Instrumentation.

The use of alternate instrumentation is acceptable because it takes into account the passive nature of the PAM Function and that there are alternate Functions available to Page 14 of 28

TSTF-525, Rev. 0 accomplish the required PAM Function, and the low likelihood that the PAM Function would be required during the time that an alternate method of monitoring the PAM Function is utilized. The allowance to temporarily rely on an alternate PAM Function is also acceptable because in most cases, the alternate Functions identified for a given PAM Function are Regulatory Guide 1.97 qualified or associated with a safety related system with a reliable power supply (e.g., pump or header discharge pressure or motor amperes) and are displayed in the main control room. Additionally, the alternate indications that may not be Regulatory Guide 1.97 qualified (e.g., rod bottom lights, rod position indicators, pump/header pressures, and motor amperage) may also be utilized in the EOPs to confirm protective system actuation.

Regarding the use of alternate indications in the EOPs, the following information is contained in Westinghouse ERGs including the Executive Volume and the Background Information for Individual Guidelines:

• SG Water Level (Wide Range)

Wide Range SG Level is used as part of the bleed and feed initiation parameter in FR-H.1, "Response to Loss of Secondary Heat Sink," to determine if SG inventory is approaching a dryout condition. Note that a dryout condition would not be approached if (minimum) SG feed flow is indicated or level is in the narrow range.

When performing FR-H.1, if feed flow or narrow range level is restored, the operator is directed out of the guideline. Therefore, consistent with the ERGs, the alternate indications of SG Narrow Range Level or Auxiliary Feedwater Flow may be used in lieu of Wide Range SG level. Inaddition, the ERG that uses Wide Rafige SG level would not be applicable if the alternate indications confirm that SG dryout will not occur.

" Power Range Neutron Flux Neutron flux decreasing is one of the indications used to determine if the reactor is tripped in the first step of E-0, "Reactor Trip Or Safety Injection," ECA-0.0, 'Loss Of All AC Power," and FR-S. 1, "Response to Nuclear Power Generation/ATWS."

The other indications used are rod bottom lights lit, reactor trip and bypass breakers open, and rod position indicators at zero. The ERGs also confirm the viability of using the intermediate and source indications to confirm a reactor trip. The ERGs assume that following a reactor trip it is expected that neutron flux will decrease rapidly to low in the power range then continue to decrease at a relatively steady rate until stabilizing in the source range. The intermediate range detectors are used to monitor the decreasing neutron flux until the source range detectors are energized.

• High Head Safety Injection Flow The ERGs contain the following information related to this indication. If high-head SI flow is not verified after SI is actuated, the operator is directed to manually start high-head SI pumps (if necessary) and align high-head SI valves to the proper Page 15 of 28

TSTF-525, Rev. 0 position for injection into the RCS (if necessary). Separate actions are provided in E-0, "Reactor Trip Or Safety Injection," to verify the high-head pumps are running and the SI valves are in proper emergency alignment. Separate actions are also provided in FR-C.1, "Response to Inadequate Core Cooling," FR-C.2, "Response to Degraded Core Cooling," and FR-C.3, "Response to Saturated Core Cooling," to verify the SI valves are in proper emergency alignment.

Auxiliary Feedwater (AFW) Flow AFW flow is used in many ERGs, along with Steam Generator (SG) narrow range level, to determine if an adequate secondary heat sink is provided. Note that "total" feed flow is generally prescribed; therefore, other sources of feed flow may be used if available. The related ERG assumptions provide the following additional information: If AFW flow (or "total" feed flow) is not greater than the specified minimum value, the operator is directed to check for adequate SG narrow range level to maintain a secondary heat sink. If narrow range level is not adequate, the operator is directed to manually start AFW pumps (if necessary) and align AFW valves to the proper position for injection into the SGs (if necessary). Separate actions are provided in E-0, "Reactor Trip Or Safety Injection," to verify the AFW pumps are running and the AFW valves are in proper emergency alignment. If AFW flow is not adequate in ECA-0.0, "Loss Of All AC Power," when AC power is not available to the motor-driven AFW pumps, the operator is directed to manually open the steam supply valves to the turbine-driven AFW pump and align AFW valves to the proper position for injection into the SGs (if necessary).

In general, the use of additional control room indications such as pump motor current, pump discharge pressure, and valve position to monitor pump or system operation are considered operator "skill-of-the-craft"; that is, no special operator knowledge or training requirements beyond normal EOP or other procedure training are considered necessary to ensure the proper understanding and performance of the procedure requirements.

Therefore, the use of this type of alternate indication to determine system status and infer equipment operation for PAM purposes is acceptable with minimal additional training.

In addition, it should be noted that the use of the alternate PAM indicatibns would only be required temporarily under the provisions of the PAM Technical Specification Actions. Therefore, it is expected that control room personnel would be aware of the requirements of the Technical Specification Actions, including the required alternate indications to be used.

The applicable PAM Technical Specification Actions also require that a report be submitted to the NRC that outlines the alternate method being used to accomplish the PAM function and the duration that the alternate method will be employed, as well as the cause of the inoperability, and the plans and schedule for restoring the affected channel(s) to operable status. Therefore, the use of an alternate method to accomplish a required PAM Function and the duration that alternate method may be employed and the plans and schedule for restoring the affected PAM Function to operable status are subject to NRC review. This reporting requirement provides the NRC with the opportunity to judge the Page 16 of 28

TSTF-525, Rev. 0 acceptability of the proposed alternate method(s) and, if necessary, question the continued use of the proposed methods.

One of the primary concerns regarding the generic approval of the use of alternate PAM instruments in WCAP-15981-NP-A was the qualification of the proposed alternate instrumentation. However, it should be noted that the allowance to use an alternate method to accomplish a required PAM Function is not intended to change the design requirements for the PAM instrumentation or the alternate instruments utilized, but rather to avoid a plant shutdown by the temporary use of available and highly reliable alternate main control room indications while the associated PAM Function is restored to operable status. In addition, this allowance includes the Technical Specification requirement that a report be submitted to inform the NRC of the use of an alternate method to accomplish a required PAM Function and the duration that the alternate method may be employed and the plans and schedule for restoring the affected PAM Function to operable status. Under the circumstances described above, the temporary use of an alternate (non-Regulatory Guide 1.97 qualified) method of accomplishing a PAM Function with NRC cognizance is acceptable in lieu of requiring a plant shutdown for the loss of indication instrumentation that provides no protective actuation functions and that only would be required in the unlikely event of a Design Basis Accident.

The allowance for continued unit operation with an alternate method of accomplishing the PAM function based on submitting a report to the NRC was previously approved on a generic bases by the NRC in the current PAM Technical Specification for the Containment Area Radiation monitors (which typically includes the use of one or more portable radiation monitors) and the Reactor Vessel Water Level Indication System. In addition to these two PAM Functions, the NRC has previously approved (on a generic bases) the temporary use of "unqualified alternates" in the following Technical Specifications:

3.4.14, "RCS Leakage Detection Instrumentation." The Actions for this Technical Specification allow for the performance of an RCS inventory balance (SR 3.4.13.1) on an accelerated basis while the affected 1E qualified leakage detection,'

instrumentation is being restored to operable status. The RCS inventory balance is typically performed using non-IE indications such as the Pressurizer Relief Tank (PRT) level, PRT temperature, and reactor coolant drain tank level.

3.6.9, "Hydrogen Mixing System." The Actions for Condition B in this Technical Specification address two inoperable hydrogen mixing trains and allow for the temporary reliance on an alternate method of hydrogen mixing while one train of the required system is restored to operable status. The alternate methods listed in the 3.6.9 Bases may not be equally qualified as the required I E hydrogen mixing system.

3.7.6, "Condensate Storage Tank (CST)." Required Action A.1 allows the temporary reliance on a backup water supply to the auxiliary feedwater pumps when the safety grade CST is inoperable. The backup water supply may not meet the same qualifications as the CST.

Page 17 of 28

TSTF-525, Rev. 0 As such, an established precedence in the Technical Specifications exists for allowing the temporary use of alternate methods of accomplishing the affected safety function in lieu of requiring a unit shutdown.

Based on the above discussions and the plant-specific information to be provided when adopting the proposed change, the use of alternate PAM indication continues to assure that the unit is operated in a safe manner (i.e., a forced unit shutdown would not always be required for the loss of instrumentation that only provides an indication function, and for which there are acceptable alternative monitoring methods). The additional safety benefit provided by the option to use an alternate PAM Function in lieu of a unit shutdown is based on avoiding the additional transition risk of a plant shutdown/restart transient. In addition, the proposed change (via the required NRC report) continues to assure that adequate regulatory control is maintained when alternate PAM methods are employed.

Change 6: The revision in the number of required Core Exit Thermocouples (CET) channels.

In Section 3.2.21.6, "Core Exit Temperature Channels," of the NRC SE for WCAP-1 5981-NP-A, the NRC addressed the proposed revision to the PAM CET requirement. In Section 3.2.21.6 of the SE the NRC stated in part:

"However, TR WCAP-15981-NP and TR WCAP-14696-A did not discuss the quadrants or how many channels should be required per quadrant. Therefore, based on the information provided, the NRC staff does not agree with the proposed change for the number of required channels for Core Exit Temperature in NUREG-143 1."

WCAP- 15981-NP-A proposed that the number of required CETs specified in Table.3.3.3-1 in Technical Specification 3.3.3 be changed from two required channels per quadrant to two required channels (total), with a Note specifying the acceptable core locations of the required channels.

The proposed change would revise the requirement for two operable CETs per quadrant to a requirement to have two operable CETs in the core central area; i.e., in the nine central core rows and columns. The proposed change was based on'the evaluation documented in WCAP-15981-NP-A. However, the staff did not concur with the evaluation because the evaluation did not relate the required number of core exit thermocouples to core quadrants. Therefore, this traveler presents additional justification for the proposed CET change in WCAP-15981-NP-A.

In order to present a-more complete discussion of the CETs, the applicable information contained in WCAP-15981-NP-A is presented first, and is followed by the additional justifications proposed to address the NRC concern regarding core quadrants.

Page 18 of 28

TSTF-525, Rev. 0 The following discussion is contained in WCAP- 15981 -NP-A:

"Any of the CETs can provide the required information for operator actions related to RCS subcooling when the core is covered with water. The risk importance of the CETs is associated with the operator actions to respond to inadequate core cooling conditions from the PRA and from the Emergency Plan.

notifications of plant conditions that may influence offsite emergency radiological protective actions. An inadequate core cooling condition is assumed in the PWROG ERGs if the highest reading CETs are indicating greater than 1200 degrees F. The peripheral rows of CETs are excluded from consideration of inadequate core cooling in the WOG ERGs.. The WOG ERG, Westinghouse Owners Group Emergency Response Guidelines," Rev. I C, (Ref. 8), Background Document for FR-0.2 identifies that the CETs in the outer two rows of assemblies should be excluded from determinations of inadequate core cooling because they can receive significant cooling from SG drainage due to refluxing. The ERG Background Document also identifies that RCS hot leg temperature indications are not recommended for use in determining an inadequate core cooling condition, since the RCS hot leg temperature reacts significantly slower than the core exit temperature to uncovery of the core for some scenarios. The major reason is that the water draining from the SGs to the core can affect the RCS hot leg temperature indication.

For the core damage assessment, the core heatup assessment in WCAP-14696-A, Rev. 1, "Westinghouse Owners Group Core Damage Assessment Guidance (Reference 9) (pages 5-1 through 5-7) shows that there is a radial temperature gradient in the core during core heatup due to inadequate core cooling. For the purpose of timely diagnosis of an inadequate core cooling condition, the central core exit thermocouple locations provide the timeliest indications. The assessment in WCAP-14696-A also shows that non-central core exit thermocouple locations can provide a rapid indication of inadequate core cooling if the thermocouple locations in the outer-most assemblies are not used. For example, a comparison of WCAP-14696-A, Figures 2b and 2c (and 3b vs. 3c) shows that there would be a delay of less than 5 minutes in the diagnosis of inadequate core cooling between the use of the central and non-central/non-peripheral CET locations. Thus, the minimum CET locations to provide information for risk significant operator actions in the EOPs and SAMG are not limited to the most central locations. Two CETs provide adequate feedback based on the relative uniformity of a core heatup during an inadequate core cooling episode.

The conditions at the RCS hot leg resistance temperature detectors (RTDs) would represent the bulk temperature of the fluid flow from the core under inadequate core cooling conditions. The bulk temperature of the fluid at the RCS hot leg RTD locations would also be significantly reduced from the fluid conditions at the exit of the core, since there would be significant heat loses to structures in the upper core plenum region and the RCS piping between the reactor vessel and the RTD location during the initial phases of the an accident with inadequate core Page 19 of 28

TSTF-525, Rev. 0 cooling. Also, since the upper indicated range of the RCS hot leg RTDs is 700 degrees F, they may be indicating off-scale high shortly after the "centrally located" CETs indicate an inadequate core cooling condition.

In defining the non-acceptable locations of the CETs in the PAM Technical Specification, the three outer rows were chosen based on the information in WCAP-14696-A, as opposed two outer rows from the ERG basis to provide additional margin for the inadequate core cooling indication. Based on the information in WCAP-14696-A and the discussion above, the required number of CET channels proposed to be included in the PAM Technical Specification is two. The recommendation of the required number of CET channels of two, and the exclusion of the CETs in the three outer rows are applicable to all two, three, and four loop Westinghouse NSSS plants.

The only alternate indication used in the WOG ERGs for the indication of inadequate core cooling is the reactor vessel level indication. However the reactor vessel level indication is not used to indicate, the need to transition from the EOP to the SAMG; only the CET indications provide an operator cue for this transition. Since the CETs are used for important operator actions in the SAMG, it is concluded that there are no appropriate alternate indications for the CETs."

Based on the WCAP-15981-NP-A CET discussion (above), the proposed change in the number of required CETs was not approved by the NRC in the SE for WCAP-15981-NP-A.

However, the following discussion provides additional information (beyond that contained in WCAP 15981-NP-A) to justify the change in the required number of CETs proposed in WCAP-15981-NP-A and to address the NRC concern regarding core quadrants.

The basis for the relation between the core quadrants and the CET channels is discussed in NUREG-0737, Item II.F.2, "Design and Qualification Criteria for Pressurized Water Reactor Incore Thermocouples." The NUREG recommends that thermocouples be:

"...located at the top of each core quadrant, of sufficient number to provide indication of radial distribution of coolant enthalpy (temperature) rise across representative regions of the core. Power distribution symmetry should be considered when determining the specific number and location of thermocouples to be provided for diagnosis of local core problems."

and "A backup display (or displays) should be provided with the capability for selective reading of a minimum of 16 operable thermocouples, 4 from each core quadrant, all within a time interval no greater than 6 minutes".

At the time the NUREG was developed immediately after the Three Mile Island Unit 2 accident, it was thought that both the absolute value of temperature in the core, as well as Page 20 of 28

TSTF-525, Rev. 0 the temperature gradient across the core, would be useful information for diagnosis of an inadequate core cooling condition. However, it was later determined that the core temperature radial gradient is not a valid indicator of inadequate core cooling based on the information in WCAP-14696-A. The most accurate indication of inadequate core cooling would be based on the absolute value of temperature obtained from non-peripheral core locations irrespective of the core quadrant.

The proposed change, based on the evaluation in WCAP-15981-NP-A, eliminates the connection between the number of operable channels and core quadrants for the purpose of Technical Specification operability. The design of the typical CET system retains its relation to the core quadrants for the purpose of the overall system capability. However, the Technical Specification operability is based on the ability of the operators to use the system to respond to the accident monitoring indications described in the EOPs. The use of quadrant-related gradient information is not required for operability if: 1) the operators do not depend on the information for the mitigation of accidents, and 2) if the algorithms applied by the display system for the CETs do not require the quadrant/gradient information to provide adequate information for the operators.

The plant EOPs rely on the CETs to monitor the core cooling critical safety function on the Safety Parameter Display System (SPDS) during and after an accident. The Core Cooling Critical Safety Function Status Tree (CSFST) (F-0.2 in the EOPs) directs the operator to different Function Restoration Procedures, depending on the CET indication or Subcooling Margin Monitor indication. The CSFST direction is based on specific absolute temperature indications and does not require the operators to evaluate core temperature gradients. None of the Function Restoration Procedures require the operators to monitor or evaluate core temperature gradients.

The EOPs also use the Subcooling Margin indication in conjunction with other RCS parameters to determine if Safety Injection (SI) may be terminated or is required to be re-initiated. The Core Cooling CSFST also uses the Subcooling Margin indication for the diagnosis of an inadequate core cooling (ICC) condition in the core. The EOPs do not require the operators to monitor or evaluate core temperature gradient.

In summary, there are no EOP requirements for the operators to monitor or evaluate core temperature gradients and the EOP action is acceptable and appropriate for response to accident conditions. Consequently, the first condition identified above regarding no operator dependency on quadrant-related gradient information is met. Additionally, the algorithms applied by the display system for the CETs do not include a temperature gradient, and only absolute values are used for temperature display. Therefore, the second condition identified above regarding quadrant/gradient information in the display system for the CETs is also met.

Page 21 of28

TSTF-525, Rev. 0 WCAP-15981-NP-A and its reference to WCAP-14696-A provide the basis that the algorithms applied by the display system for the CETs do not require the quadrant/gradient information to provide adequate information for the operators and that a restriction of a CET in each core quadrant is overly restrictive. Paraphrasing from Section 6 of the WCAP- 14696-A:

For the core damage assessment, the core heatup assessment in WCAP-14696-A (pages 5-1 through 5-7) shows that there is a radial temperature gradient in the core during core heatup due to inadequate core cooling. For the purpose of diagnosis of an inadequate core cooling condition, the central core exit thermocouple locations provide the timeliest indications. However, the assessment in WCAP-14696-A also shows that non-central core exit thermocouple locations can provide a rapid indication of inadequate core cooling if the thermocouple locations in the outer-most assemblies are not used.

Thus, the minimum CET locations to provide information for risk significant operator actions in the EOPs and SAMGs are not limited to the most central locations. Two CETs provide,'adequate feedback based on the relative uniformity of a core heatup during an inadequate core cooling condition.

With respect to a LOCA and core temperatures, WCAP-14696-A states:

"The core heatup is generally greatest in the region of the core that had the highest local power density just prior to the accident due to the greater inventory of decay heat generating fission products in this region. Although the local power densities change throughout the life

• ofa core cycle, the highest power densities are generally found near the center of the core and the lowest power densities are generally found near the periphery of the core. With regard to heat sinks, near the center of the core, all of the fuel assemblies are nearly the same temperature and pre-accident power level so that there.is little heat sink impact on the core heatup rate. On the other hand, near the periphery of the core, heat transfer to the core baffle/barrel can be significant."

With respect to a non-LOCA Transient event and the impact of natural circulation after core uncovery, WCAP- 14696-A states:

"The natural circulation flow is upward in the center of the core and downward near the core periphery. Under these conditions, the core exit thermocouples near the periphery of the core may indicate the temperature of the downward recirculation flow rather than upward steam flow from the fuel assemblies."

Page 22 of 28

TSTF-525, Rev. 0 With respect to the behavior of CETs during core overheating leading to core damage, WCAP- 14696-A states:

"Referring to Figures 2 and 3, if it is assumed that gap release occurs at a core exit thermocouple indication of 1400°F and grain boundary release occurs just after the core exit thermocouple indication exceeds 2500'F, the periphery fuel assemblies would not be at a high enough temperature to experience fuel rod clad rupture when the central fuel rods are close to their melting temperatures. Thus, in a core damage accident such as this, a portion of the core would contribute fission products from fuel rod melting, a portion of the core from grain boundary release, a portion of the core from gap release and a portion of the core would contribute no release at all."

With respect to the progression of core damage accidents and the CETs, WCAP- 14696-A states:

"Analyses of the progression of core damage accidents shows that the difference in readings amongst the core exit thermocouples near the center of the core are likely to be within 50 'F of one another. In other words, the fuel assemblies in the center of the core heatup together with a very small radial temperature gradient (i.e., the temperature difference between adjacent assemblies near the center of the core is very small). As the distance from the center of the core becomes greater, the temperature gradient becomes more pronounced until a very large gradient is observed at the periphery of the core. Thus, it is quite likely that several core exit thermocouple readings will approach the high temperatures indicative of cladding damage and significant fission product release from the fuel pellets at about the same time.

This provides confidence that there is not an erroneous core exit thermocouple reading and that the diagnosis according to this guidance is correct."

WCAP-14696-A demonstrates that there is no quadrant bias for the rapid diagnosis of inadequate core cooling conditions.

Therefore, the need for a large number of CETs representing a uniform radial distribution in the core is not supported by the analyses discussed in WCAP-14696-A.

In defining the non-acceptable locations of the CETs in the proposed PAM Technical Specification, the three outer rows were chosen based on the information in WCAP-14696-A, as opposed to two outer rows from the ERG basis to provide additional margin for the inadequate core cooling indication. Based on the information in WCAP-14696-A and the discussion above, the required number of CET channels proposed to be included in the PAM Technical Specification for the CET inadequate core cooling function is two and the number of CETs per channel is two, provided they are located in the 9 central core rows and columns.

Page 23 of 28

TSTF-525, Rev. 0 The only basis for representing radial locations in the core is for localized fuel blockages in a single quadrant of the core due to flow maldistribution such as a large cold leg break where the core quadrant closest to the broken cold leg may theoretically exhibit an uneven heatup rate compared to the remainder of the core. However, this is not shown in the design basis analyses. Additionally, if there is localized fuel blockage in one quadrant that causes the CETs in that quadrant to read above the F-0.2 setpoint, the response is to continue providing water to the core. Therefore in this case, the diagnosis of an inadequate core cooling condition does not alter the strategy to provide water to the core.

Based on the evaluation above, the operators would have adequate CET information to make a determination of inadequate core cooling with two channels of CETs, with a channel consisting of two CETs, in the central 9 rows and columns of the core. A single failure of a channel would still result in two CETs being available.

5.0 Regulatory Analysis 5.1 No Significant Hazards Consideration The proposed change revises the recommended Post Accident Monitoring (PAM) instrument functions, Actions, and Surveillance Requirements based on approved Topical Report WCAP-15981-NP-A, Revision 0, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants," and additional justification.

The TSTF has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the instruments, Actions, and Surveillance Requirements for PAM instrumentation. PAM instrumentation is not an initiator to any accident previously evaluated, and as a result, the proposed change does not involve a significant increase the probability of an analyzed event. The PAM instruments provide information for manual operator actions for which no automatic control is provided and, as a result, function to mitigate the consequences of an accident. The proposed change retains in the Technical Specifications the PAM instrumentation functions needed for manual operator actions for which no automatic control is provided. The proposed change to the Actions is administrative and does not affect the actions taken when a PAM instrument function is not operable. The proposed change to the Surveillance Requirement adds a new testing requirement to the containment isolation valve position indication function which is a more appropriate test to confirm the function is operable. As a result, the proposed change does not affect the ability Page 24 of 28

TSTF-525, Rev. 0 of the PAM instrumentation to provide information to the operator to mitigate an accident.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change revises the instruments, Actions, and Surveillance Requirements for PAM instrumentation. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements that could initiate an accident. The changes do' not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises the instruments, Actions, and Surveillance Requirements for PAM instrumentation. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by these changes. The proposed change will not revise the indication provided by the affected instruments, and all operator actions based on these indications that are credited in the safety analyses will remain the same. As such, the proposed change will not result in plant operation in a configuration outside the design basis or assumptions of the design basis accident analyses.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, the TSTF concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements/Criteria In the following paragraphs, the applicable regulatory requirements/criteria related to the proposed changes are discussed.

Page 25 of 28

TSTF-525, Rev. 0 10 CFR, Appendix A, General Design Criteria (GDC)

GDC 13 includes a design requirement that instrumentation be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions as appropriate to ensure adequate safety. There is no impact on the requirement of GDC 13, since the proposed change does not include any plant design changes. The proposed change does not eliminate or otherwise alter any existing instrumentation.

GDC 19 includes a requirement that a control room be provided from which actions can be taken to maintain the nuclear power unit in a safe condition under accident conditions, including loss-of-coolant accidents, and that equipment, including the necessary instrumentation, at appropriate locations outside the control room be provided with a design capability for prompt hot shutdown of the reactor. In addition, GDC 19 requires that adequate radiation protection be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of the specified limits. There is no impact on the requirements of GDC 19, since the proposed change does not include any design changes or plant modifications.

In addition, the change does not introduce any changes that could adversely affect the potential radiation exposure of control room personnel.

GDC 64 requires that means be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations and postulated accidents. There is no impact on the requirement of GDC 64, since the proposed changes do not modify the design or otherwise alter the current plant instrumentation used for monitoring radioactive releases.

No monitoring instrumentation is being eliminated or modified due to the proposed changes.

10 CFR 50.49 specifies design and performance requirements for safety-related instrumentation exposed to adverse environments during accident conditions. The proposed changes do not impact the requirements of 10 CFR 50.49. The proposed changes do not introduce changes that affect the design or performance of the safety-related instrumentation subject to 10 CFR 50.49. There are no plant design changes or modifications associated with the proposed changes.

10 CFR 50.36 contains requirements applicable to the content of a plant's Technical Specifications. The proposed changes utilize the criteria of 10 CFR 50.36(c)(2)(ii) to evaluate the content of the PAM Technical Specification: The proposed change includes the relocation of certain instruments from the Technical Specifications that do not satisfy any of the criteria of 10 CFR 50.36(c)(2)(ii). In addition, the proposed changes include the addition of several instruments to the PAM Technical Specifications that have been determined to meet one or more criteria of 10 CFR 50.36(c)(2)(ii). As such, the proposed changes are consistent with the requirements of 10 CFR 50.36.

• Page26 of 28

TSTF-525, Rev. 0 10 CFR 50.67, "Accident source term," contains requirements and dose limits associated with the use of an alternate source term for design basis radiological consequence analyses. The proposed changes do not affect the requirements or dose limits specified in 10 CFR 50.67. In addition, the proposed changes do not affect the assumptions of any radiological consequence analysis.

10 CFR 100.11, "Determination of exclusion area, low population zone, and population center distance," contains requirements and dose limits associated with the design basis radiological consequence analyses. The proposed changes do not affect the requirements or dose limits specified in 10 CFR 100.11. In addition, the proposed changes do not affect the assumptions of any radiological consequence analysis.

In summary, the proposed changes do not involve any design changes to the PAM instrumentation, or changes to the physical arrangement of PAM instrumentation. The proposed changes result in modifying the scope of the existing PAM Technical Specification consistent with the requirements of 10 CFR 50.36. Therefore, the proposed changes provide enhanced assurance that the required PAM Functions remain capable of performing their accident monitoring function. Thus, the proposed changes do not adversely impact the design or performance characteristics, of the PAM system, or any other system.

In conclusion, based on the considerations discussed above, (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.

6.0 Environmental Considerations A review has determined that the proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

,However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types, or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

7.0 References

1. Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980.

Page 27 of 28

TSTF-525, Rev. 0

2. WCAP-15981 -NP-A, Revision 0, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants," September 2008.

3. NUREG-1431, "Standard Technical Specifications Westinghouse Plants,"

Revision 3.1, December 2005.

4. Supplement I to NUREG-0737, "Requirements for Emergency Response Capability," December 1982.

5. WCAP-1 1618, "Methodically Engineered, Restructured and Improved, Technical Specifications, MERITS Program - Phase II Task 5 Criteria Application,"

November 1987.

6. NRC letter from T. E. Murley (NRC) to W. S. Wilgus (B&W Owners Group),

"NRC Staff Review of Nuclear Steam Supply System Vendor Owners Groups' Application of the Commission's Interim Policy Statement Criteria to Standard Technical Specifications," May 1988.

7. WCAP-14986-A, "Westinghouse Owners Group Post Accident Sampling System Requirements: A Technical Basis," Revision 2, July 2000.

8. Westinghouse Owners Group Emergency Response Guidelines," Revision IC, 1997 9; WCAP-14696-A, "Westinghouse Owners Group Core Damage Assessment Guidance," Revision 1, November 1999.

Page 28 of 28

TSTF-525, Rev. 0 PAM Instrumentation 3.3.3 3.3 INSTRUMENTATION 3.3.3 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3 The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

---

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

The Power Range Neutron Flux PAM Function is not required in MODE 3.

ACTIONS

---

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

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required channel to 30 days with one required OPERABLE status.

channel inoperable.

B. Required Action and B.1 Initiate action in accordance Immediately associated Completion with Specification 5.6.5.

Time of Condition A not met.

C. One or more Functions C.1 Restore one channel to 7 days with two required OPERABLE status.

channels inoperable.

---

- NOTE -- -------

Condition D is only D.1 Initiate action in accordance Immediately applicable to PAM Functions with Specification 5.6.5.

[1, 4, 8, 10, and 121 in Table Enter the Condition referenced in 3.3.3-1. Table 3.3.3 1 for the

-ha4-el--

D. Required Action and associated Completion Time of Condition C not met.

WOG STS 3.3.3-1 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation 3.3.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

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

Condition E is only 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 /> applicable to those PAM Functions in Table 3.3.3-1 AND not addressed by Condition D. E.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 /> E. Required Action and associated Completion Time of Condition C not ..

met.

As required by Required ctoD.1 -An F. Asb required by Required F.I Initia;te action- in accodance Immediately Ac-tio D.1I anPAWith Specification 5.6.5.

Ta;ble 334.31 SURVEILLANCE REQUIREMENTS

---

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

SR 3.3.3.1 and SR 3.3.3.2 apply@applies to each PAM instrumentation Function in Table 3.3.3-1.

SR 3.3.3.2 applies to each PAM instrumentation Function in Table 3.3.3-1 except for the Penetration Flow Path Containment Isolation Valve Position. SR 3.3.3.3 aDDlies only to the Penetration Flow Path Containment Isolation Valve Position.

SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.

SR 3.3.3.2 --------------------- NOTE---------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. [18] months WOG STS 3.3.3-2 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation 3.3.3 SURVEILLANCE FREQUENCY SR 3.3.3.3 Perform TADOT. [181 months WOG STS 3.3.3-3 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 1)

Post Accident Monitbring Instrumentation CONDIT ON REFERENCED FROM REQUIRED FUNCTION REQUIRED CHANNELS ACTIOQ DN

1. Power Range Neutron Flux 2

2. Sou'r" Range Neutron FluxSteam Generator (SG) 2 Pressure a) SG "A" 2 b) SG "B" 2 c) SG "C" 2 d) SG "D" 2

3. RQeator Coolant SlerRSt* HotL*og 2-1perrop E.

TeeýatiweRefuelinq Water Storaqe Tank Level (Wide Range)

4. RCS Cold Log TemperaturoHigh Head Safety 2 E-Injection (SI) Flow L1 per leelSI train)

5. R-G-Reactor Coolant System Pressure (Wide 2 Range)

(j. Keacto vessel WaWe 6ceV9 2

7. 1Containment Sump Water LeY9l (Wide Range) 2

86. Containment Pressure (Wide Range) E_

97. Penetration Flow Path Containment Isolation Valve 2 per penetrton flow E_

Position path 408. Containment Area Radiation (High Range) 2 F-1-1-9. Pressurizer Level 2 E-4-210. Steam- GeneratoG Water Level (Wide Range) 2 per steam generator a) SG "A" 2 b) SG "B" 2 c) SG "C" 2 d) SG "D" 2

13. Cende.nsate Storage Tank Level. E-4411. Core Exit Temperature ,Qua.d*a.rt[4 2(c)

15. Co*r

_ Eit Temperature Quadrant [21

16. Core Exit Tempe*ature Q,-drant[3] E-

17. Cre rEXit Temperature Quadrant [41 E-4812. Auxiliary Feedwater (AFW) Flow 2 E-(11ner AFW train)

(a) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

WOG STS 3.3.3-4 WRev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation 3.3.3 (b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

(c) A channel consists of two core exit thermocouples (CETs) in the nine central core rows and columns.

---

--- REVIEWERS NOTE - .....-------------.--------.-................-----------

Table 3.3.3-1 shall be amended for each unit as necessary to list the plant specific instrument Functions identified by the application of the~methodology contained in WCAP-15981-P-A, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants".

REVI EWVE 'S NT Table 3.3.3 1 rhall be amnedfreah u-not asncesr to lSt

1. All Regulatowy Guide 1.97, Type A isrmnsand 2.Alegulatory Guide 1.97, Catogor,' I non Type A ntum tsiacodceWith the unit's Rogulator; Guide 1.97, Safet EyaluatiGn Repept.

WOG STS 3.3.3-5 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Post Accident Monitoring Report When a report is required by Condition B or F-D of LCO 3.3.[3], "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.6 [Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days. The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.]

5.6.7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, "Steam Generator (SG) Program." The report shall include:

a. The scope of inspections performed on each SG,

b. Active degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism,

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged [or repaired] during the inspection outage for each active degradation mechanism,

f. Total number and percentage of tubes plugged [or repaired] to date,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

[h. The effective plugging percentage for all plugging [and tube repairs] in each SG, and]

[i. Repair method utilized and the number of tubes repaired by each repair method.]

WOG STS 5.6-4 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBAs)- or that the Probabilistic Risk Assessment (PRA) has shown to be significant to the public health and safety.

The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and to assess unit status and behavior following an accident.

The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. These essential instruments are identified by unit specific documents (Ref. 1) addressing the recommendations of Regulatory Guide 1.97 (Ref. ) as required by Supplement 1 to NUREG-0737 (Ref. 2) and by evaluatinq the instrumentation consistent with the methodology contained in WCAP-1 5981 (Ref. 3). This methodology considers the use of the accident monitorinq instrumentation in the PRA, Emer-gency Operating Procedures (EOPs), Severe Accident Management Guidance (SAMG) procedures, and Emergency Plan (E-Plan).

The instrument channelscontrol room monitoring instrumentation Functions required to be OPERABLE by th& L-tOhave been evaluated and selected in accordance with the screening criteria contained in WCAP-15981. The screening criteria were used to identify the PAM instrumentation important to safety (i.e., monitor plant parameters that are the basis for important operator actions to bring the unit to a safe stable state in the event of an accident). incud t, o cla..e. of parameters d,

identifiedr dring unit sp*cfic* *iplemeRtatioRnof Regulator; Guide 1.97 as Type A and Categor,' 1variables.

Type A variables. are finclu-ded inthis ILC) because they provideth pirimary information requiredd foar th~e con-trol roomn operator to take spec ,;+';t-manually controlled atin fo-r %Whichno auomtc otrol is provided, and- that are required fogr safety systems to accomplish their 6afety f-rfuncions forlDBAs. BeAuP se the list of Type A variablesA differswiel

-be-t;eenunits Table 3.. nthe accomnpanying LCO contains no examples of Type A variables, except for those that may also be Category I VAriAbleS.

B 3.3.3-1 Rev. 3.1, 12/01/05 WOG STS B 3.3.3-1 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 Category 1variables are the key variables deemned risk Significant becauSe they TR-09 d-leedd toG:

a- Determine whether othoe *y*tems important. to safety are peif.oFrni..g their intended functions,-

Provide information to the operators that will enable them to determine the lilkelihood of a gross breach of the barriers to radoativtyrelease, and WOG STS B 3.33-2 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES BACKGROUND (continued)

• Provide infor~m.ation regarding the release of rad-ioac-tive Materials to alwfor early indication of the need to initiate actfion enecessary to protect the pubicIO, and to es-timate the magnitude of anyimedn threat=

The6e key variables are-identifi by the unit specific Regulato.Yt G-uidel9 1.97 analy6e6 (Ref. 1). These analyses, identify the unit specific Type A and Category : variables and provide justification for d&.iat..g from the NRC proposed lfist of Category 1variables.-

REV"EWE R'S NOTE Table 3X3.3-4 provides a list of vaOriabes typical of those identifiedb h uit specific Regulatory Guide 1.97 analyses. Table 3.3.2-4i ui specific Technical Specifications (TS) shall list all Type A and Cateq*ry I var;iales den;tified by the unit specific Regulatory Guide 1 97 aRalyse*s, aso amnend-e-d by the NRC's Safety E=valuation Repopt (SER).

The specific ntrmn Func-tions fis-te-d inTable 3.3.3- 1-are discuissed i the LCO section.

APPLICABLE SAFETY ANALYSES The PAM iu o specification ensures the operability of RegulatoryGuide 1.97 Type A and Catgor I r.iables instrumentation to monitor plant parameters necessary for safety. significant operator actions so that the control room operating staff can:

e Perform the diagnosis specified in the em.ergency operating PFeedu esEOPs (these variables are restricted to preplanned actions for the primary success path of DBAs), e.g., loss of coolant accident (LOCA),

0 Take the specified, pre-planned, manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety function,

• Determine whether systemsimplement procedures or guidance that has been shown to have an important impo.tant to safety are peFo-rming their intended functionsrole in preventinq core damage or early fission product releases, Determine the likelihood of a gross breach of the barriers tethat prevent radioactivity release, WOG STS B 3.3.3-3 Rev. 3.1, 12/01/65

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 Determine if a gross breach of a barrier has occurred, and B 3.3.3-4 Rev. 3.1, 12/01/05 WOG STS WOO STS B 3.3.3-4 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES APPLICABLE SAFETY ANALYSES (continued)

Initiate action necessary to protect the public and to estimate the magnitude of any impending threat.

The PAM instrumentation that meets the definition of Type Aselected in Reguatory; Guide 1 .97 accordance with WCAP-1 5981 is used to monitor plant parameters necessary for safety significant operator actions and satisfies Criterion 3 and/or 4 of 10 CFR 50.36(c)(2)(ii).

Categor,' 1,non Type A,ins6tru-mentationR mus-rt be retained in TS becauswe it iineddto assist ope-ra-torsF in minimizing the consequence o accdens.Therefore, Category 1,non Type A, variables are ipratfor reducGing public risk.

LCO The PAM instrumentation LCO provides OPERABILITY requirements for Regulato"y Guide 1.97 Typo A monitors, Which .prFvid.e inform.at;i Fequiied bthe control room monitoring instrumentation Functions important to safety (i.e., monitor plant parameters that are the basis for important operator actions to bring the centrol reeom operators to perform certain manual actions G196G.fiedplant to a safe stable state in the uait Emergenc. O*peating Procedu-res These mna acin ensure that a systemFncan accomvplish it saft* fuction, and are credited in the safety analyses. Additionally, this LOad~dresses Regulatory Guide 1.7 in..s....ruments that .havbn designated Categrey 1,non Type A.event of an accident).

The OPERABILITY of the PAM instrumentation ensures there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident. This capability is consistent with the recommendations of Refer-AIee4Regulatory Guide 1.97 (Ref. 1) and the quidance provided in WCAP-1 5981 (Ref. 3) for selecting the appropriate instrumentation.

LCO 3.3.3 requires two OPERABLE channels for most Functions. Two OPERABLE channels ensure no single failure prevents operators from getting the information necessary for them to determine the safety status of the unit, and to bring the unit to and maintain it in a safe condition following an accident. Therefore where plant design permits, the two channels required OPERABLE by the LCO should be supplied from different trains of electrical power.

Furthermore, OPERABILITY of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information. More than +h to c.hannels may be required ;at som, unit ingf the untpecific Regulatory Guide 1.97 analyses, (Ref. 1)determined that failu-re of one a.ccident monitoring channel results in information ambiguity (that is, the redundant displays disagree) that could lead operators to defeat or fail to accomplish a required safety function B 3.3.3-5 Rev. 3.1, 12/01/05 WOG STS WOG STS B 3.3.3-5 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 The exception to the two channel requirement is Penetration Flow Path Containment Isolation Valve (CIV) Position. In this case, the important information is the status of the containment penetrations. The LCO requires one position indicator for each active CIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of a passive B 3.3.3-6 Rev. 3.1, 12/01/05 WOG STS WOG STS B 3.3.3-6 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued) valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE.

Table 3.3.3-1 provides a list of var.iables typio of- thosethe control room indications identified by the unit Specific Regu!ateo.' Guide 1*97as important to safety in accordance with the methodology of WCAP-15981 (Ref.-1-aRaly6.- 3). The monitoring functions specified in Table 4-.3.3-1 inUnit spec-fific- IS shou ld list all Topo A anRd Categor,' I variable identified by the unit specific Regulator,' Guide 1.97 analyses, as amended by the NSRC's SZER-.

Type A and Categeory variables.3-11 are required to.. et Regulato,, y Guide 1.97 Categoy (Ref. 2) des*ig a q atn reuirements for seismic and environmental qualificatio, sigl failure criteFio*, u1,tiizatio of em.ergency standby power, immediately accessible diplay, contfinu ous11; readout, and recording of *di*rpQaPERABLE by LCO 3.3.3. The following discussions describe the instrument monitoring Functions identified in Table 3.3.3-1.

---

Reviewer's Note ---------------------------------------

Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1. These discussions are intended as examples of what should be provided for each Function when the.unit specific list is prepared.

1,-2. Power Range and Su'rce Range Neutron Flux The Power Range Neutron Flux indication is used to confirm a reactor shutdown following an accident or other receipt of a reactor trip. The PRA shows that operator actions to manually shutdown the reactor in the event of a failure of the automatic actions, as determined from the Power Range a:;dSRe, -eNeutron Flux indication, can be important to safety.

The Power Range Neutron Flux indication is previded to verify reacter shu-tdo':Wn. The ve rFangessatisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

The Applicability is modified by a Note that provides an exception to the OPERABILITY requirement for Power Range Neutron Flux indication in MODE 3. The Power Range Neutron Flux indication is used to confirm an automatic reactor shutdown from power operation. Therefore, the PAM Power Range Neutron Flux indication requirement is -aFe-only applicable in MODES 1 and 2 when the WOG STS B 3.3.3-7 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 power range instrumentation functions to provide the ,nees aPy-to GOVoPAM indication (i.e., reactor trip).

Alternate Monitoring Methods Ifthe power range neutron flux indications are not available, an alternate method of verifying a reactor trip is a combination of:

1. A neutron flux indication:

i. The intermediate range neutron flux indication, or ii. source ranqe neutron flux indication: and

2. A rod position indication:

i. The rod bottom lights, or ii. The rod position indicators.

These alternate indications can also provide the information necessary for operators to determine the need to initiate a manual reactor trip.

the full raRng of flux that may occu1F,r p ccide Got NIAetron flu.x i u.ed for accident diagnosis, verifi'cation of snbnriticality, an,d diagnOsis Of positiVe reactiVity ine"ion.

3, 4. Reactor Coolant S*ntem (RGC Hot a- d ColdlegI TeArar, r" Steam Generator (SG) Pressure RGC Hot and Cold Leg Temperatuiresi are Category I variabl rided for nerification of core cooling and long termn WeaRn ~~0.

nRCS lirg margind lea RCS colingarmargin illi allowtermination ot Safety injoction (S1), if still in progress, Or reinitiatiOn Of SI if it has hbe*en*stopped. RCS SIubslning margin is also used for unit stabieli;zation And cooldoWn control SG Pressure indication is used followinq an accident or receipt of a reactor trip signal to indicate secondary side integrity. It is also used to indicate the target pressure for RCS depressurization following a SG tube rupture accident to terminate the RCS inventory loss. In the event of a SG tube rupture accident, the EOPs instruct the operators to depressurize the RCS to a pressure below the secondary side pressure in the ruptured SG. RCS depressurization to a pressure less than the SG pressure terminates the RCS inventory loss and WOG STS B 3.3.3-8 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 terminates the SG inventory gain, preventing overfill of the SG. The termination of the break flow is an operator action assumed in the design basis SG tube rupture analysis for which no automatic action is provided. The PRA shows that failure to depressurize the RCS to a pressure less than the secondary side pressure in the ruptured SG is a risk significant operator action.

Due to the number and redundancy of indications provided by 2 channels per SG and multiple SGs, the indications for each SG are listed as individual PAM Functions on Table 3.3.3-1.

SG Pressure indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

B 3.3.3-9 Rev. 3.1. 12/01/05 WOG STS WOG STS B 3.3.3-9 Rev. 3.1. 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

In addition, RCS cold leg tem.perature *R*..u in coj Oed tion with RGS hot leg temporature to verify,the unit condiltions nt~eessry to ostablish natu~ral circumlation in the RCS.

R etaco outlet pra pt t th a io s are provded.by two l*at respon*e* Fre.VVtancev Ev3.M,,vt.s -,An asociated tranRsmitters in eah lGOP. The channels provide

.dindiation over a range of 320F to 70 0 F.

3. Refueling Water Storage Tank Level (Wide Ran-ge)

Refuelinq Water Storage Tank (RWST) Level provides an indication of the water inventory remaining for use by containment spray and safety iniection for core cooling and containment cooling. [No operator actions in the design basis accident analysis are based on the RWST level indication. The switchover from the RWST to the containment sump is performed automatically.1 The PRA shows that in the event of an accident in which the RCS inventory losses are outside of containment (e.g., SG tube rupture and interfacing system LOCA), the remaining RWST level is an important indication in the EOPs for choosing the appropriate operator actions to maintain core cooling. The PRA shows the importance of diagnosing the need for implementing RWST refill'to maintain a sufficient inventory for long term core cooling following these events.

RWST Level indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

B 3.3.3-10 Rev. 3.1, 12/01/05 WOG STS B 3.3.3-10 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

4. High Head Safety Injection (SI) Flow High Head SI Flow indication is used to confirm automatic safety injection initiation following a design basis accident. Therefore, the required flow indicator for this PAM Function is the one installed in the automatic High Head SI flow path. The results of the PRA show that this is a risk significant operator action. Failure to manually initiate SI flow when the automatic initiation fails can lead to a significant increase in core damage frequency. The operator action is based on the Emergency Core Cooling System flow indication in the control room. The PRA shows that only high head safety iniection is important for all accident sequences except the unlikely double-ended guillotine rupture of the largest reactor coolant pipe.

Therefore, only the High Head SI Flow indication is required.

High Head SI Flow indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods If the High Head SI Flow indication is not available, an alternate method of verifying SI initiation can be provided by either:

1. The High Head SI pump motor amperage and the High Head SI pump discharge pressure indications; or

2. The header pressure and the automatic SI valve position indications.

5. Reactor Coolant System Pressure (Wide Range)

The PRA indicates that Reactor Coolant System (RCS-wide-rami).

pressure is a GategGei-l-variable previdedimportant to safety for verification of core cooling and RCS itgtylong term 6urweillance.

cooldown and depressurization following a SG tube rupture. RCS Wide Range Pressure indication provides the information necessary for RCS depressurization for the SG tube rupture accident to terminate the RCS inventory loss. In the event of a SG tube rupture accident, the EOPs instruct the operators to depressurize the RCS to a pressure is used to verif, delivery of SI f*eo to R, S from at least one tra;in* when the RGS pressure is- below the pump ,shueff head*secondary side pressure in the ruptured SG. RCS depressurization to a pressure less than the SG pressure terminates the RCS inventory loss and terminates the SG inventory gain, preventing overfill of the SG. The termination of the break flow is an B 3.3.3-11 Rev. 3.1, 12/01/05 WOG STS WOC STS B 3.3.3-11 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 operator action assumed in the design basis SG tube rupture analysis for which no automatic action is provided. RCS pressure is also used to Ve-if' clo..ure Of mnalcoed pray In.e valves and for operator action to terminate SI in the event of a steamline break to prevent pressurizer power operated relief val. es POR/s).

In addition to these verificationsoverfill for which no automatic actuation is provided. Additionally, RCS pressure is used for determining RCS subcooling marg.iRS 6ubG.G.i.' margin will allow.. term.ination, of SI, if still in progre.., following an accident or eW 1iftireceipt of a reactor trip sigqnal. SI has been stopped.

RG p tS s;:---re canmalso b eAused:

rrenssu

" to determine whether to terminate acV-;tuatd S;I or to reinitiate 6tepped SI,

" to deterrmine whenp to rease~t SI and shut off low head SI,

--" L',.; ;;;*;;L;*;;V  ;*.L--*-;L ;V;." ;;*--*-*; ,31.

as raFctor cOolant pump (RCP) trip criteria, and

- to make a determination on the nature of the acc,,ident in

-progress, and where to go Re~d in the parocedure.

RCS su-bcooling margin is also'wed for unit stabilization and cooldown control.

B 3.3.3-12 Rev. 3.1, 12/01/05 STS WOG STS B 3.3.3-12 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASER RCS p~eresure is alsorel E4ate-d to_three eciion about depressurization. They ar**

' to der.rterminAWhether to proceed with primand system depressu ization, tovrfy termination of depressurization, and 0 to determine whether to losbe accumnuatGor islto 4avs during' ai GORI1L-IAfA GIUUGuIuunG upttn1L10

.A.final use of RCS pressur_,_,e is- to d-eteFrmine wvhether to Operate the ppressurizegr heate~r.

In some units, RCS prvssure is a Type A variable beca-use th operator user-this idctiont monitor the cooldown o~f the4 RCS following a steamn geneerator tube rupture (SGT-R) or small bra LOC)A.. Operator actions to maintain a controlled cooldown, such as_

adju6sting steam generator

, (SG) pressurne or level, would use thos indication-. Furthermore, RCS pressure is one factor that may-be uswed- in decisions; to termin;at RCP operation.

RCS pressure indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

6. Rean-tor \ese WAter Lev Reactor Vessel Water Level is provided for ver"infation and long teFm surveillanc of core cooling. It is also used for accident diagnosis and- to d~eteprm.ine.A ractor coolant inVentor,' adequacy.

The Reac-tor Vessel W-ater Level Monitoring System providesr a; dirret mneasuerement Of thes collapsed liquid-le-vel Rabve the fulz alignm:ent plate. The collapsed level represents the amoeunt of liquid mass that is in the reactor v:essel aboe the core. Measurement of the collapsed water level is selected the Mfater netr'

7. Containment Surnp Water Level (Wide Range)

Cortainment Sump Water Level s provided fr verification* and long term surveillanc~e of RGS integrit, C-ontafinment Sump W-ater Level isusbe-d to determfine:

B 3.3.3-13 Rev. 3.1. 12/01/05 WOG STS WOO STS B 3.3.3-13 Rev. 3.1. 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 sum lV8 accidetdanss containmentIM Rev. 3.1. 12101105 B 3.3.3-14 WOG STS WOO STS B 3.3.3-14 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued) whetobein hoRecGircu-lA-tion procedure, and

. whe-ther to terminate SI, if still in progress.-

8

6. Containment Pressure (Wide Range)

Containment Pressure (Wide Range) is provided for ver-fication of RCS and containment OPERARiLITYindication is provided for assessing containment cooling and containment integrity. No operator actions in the design basis accident analysis are based on the containment pressure indication. Containment pressure is an indicator of the potential loss of a fission product boundary in the Emergency Action Levels in the E-Plan. Containment pressure is a key indicator in the declaration of a General Emergency level and the potential need for offsite radiological protection actions. Containment pressure may also be used in post accident conditions to determine when to vent the containment to prevent overpressurization.

ContainmenRt prsUrs, e to verify clo"ure of mnaiR Ste'm-is-olationR valves (IVS Vs), and containment spray Phase 1 soato wNhen High 3 containment pressure .fis- reached.

-- gContainment Pressure (Wide Range) indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

7. Penetration Flow Path Containment Isolation Valve Position Penetration Flow Path CIV Position is provided for verification of Containment OPERABILITY, and Phase A and Phase B isolation.

The CIV position indication provides a direct indication of a failure to completely isolate the containment following the receipt of a containment isolation signal. The E-Plan identifies that an elevated emergency action level should be declared following an accident in the event of a failure of the automatic containment isolation.

When used to verify Phase A and Phase B isolation, the important information is the isolation status of the containment penetrations.

The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active CIV in a containment penetration flow path, i.e., two total channels of CIV position indication for a penetration flow path with two active valves. For containment penetrations with only one active CIV having control WOG STS. B 3.3.3-15 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve, as applicable, and prior knowledge of a passive valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE. Note (a) to the Required Channels states that the Function is not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

Each penetration is treated separately and each penetration flow path is considered a separate function. Therefore, separate Condition entry is allowed for each inoperable penetration flow path.

CIV Position indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

B 3.3.3-16 Rev. 3.1. 12/01/05 STS WOG STS B 3.3.3-16 Rev. 3.1. 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued) 4-08.Containment Area Radiation (High Range)

Containment Area Radiation is provided t. m...ontor for the potential of Gignifia*nt radwifion release, and to provide release as6esment for use by operators in determining the need- to- invo-ke- site9 emergency plans. Containmeant radi~ation level isusd to determine ifahigh energy lne break (HELB) has, occurred, and -w'hetherthe eVent i inside or ountide of containmentHiqh Range provides an indication of a loss of one or more fission product barriers. The Emergency Action Levels in the E-Plan utilize the Containment Area Radiation High Range monitor as an indication of the potential loss of one or more fission product barriers in the assessment of the declaration of a General Emergency level and the potential need for offsite radiological protection actions. The post accident Core Damage Assessment also uses the Containment Area Radiation High Range monitor as an input to the determination of core damage.

---4The Containment Area Radiation High Range indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods A portable radiation monitor (with appropriate multiplier if necessary) can be used as an alternate method of indication for Containment Area Radiation High Range.

9. Pressurizer Level Pressurizer Level is used to determine wihether to terminate SI, if still i progress, or to reinitiate SI if it has been stopped. Knowledge of presurier ater level ialous-ed to verify' the unit conditions necesrw to establish natu ral niwr*

c ation in the RCR;ad toe that the unit ir maintained in a safe sh'utdow-n conditigoindication is used for the SI termination criteria to prevent pressurizer overfill. The termination of SI to prevent pressurizer overfill is an operator action assumed in the design basis steamline break analysis for which no automatic actuation is provided. The PRA. also indicates that SI termination in the event of a steam generator tube rupture is required for long term core cooling.

12. Steam Generator WaterPressurizer Level indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

10. SG Water Level (Wide Range)

WOG STS B 3.3.3-17 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 SG Water Level (Wide Range) indication is provided to monitor operation of decay heat removal via the SGs. The Gategepy4 indicration of SGQ level is-the extdend-ed startup range level instru mentation. The ekdended s~tartup range leVel covers a span oa

-6*inches to 9 394 inches above the lower tubesheet. The m diferential p.resure i displayed

,easured , in inchesof water ,t Temperature comrpens-ation- of thisiniatoni peiformed manuWally by the operator. Redun~dant mon-iteorin capability is proVided by two ,%

trains of insrmentation. The uncopnae level signalis nu to the-unicmputer, a control room iniaoand the EmFergency 1=feedwbAatQr ConrolI System.

SG Water Level (Wide Range) is used to:

• identify the faulted SG following a tube rupture,

• verify that the intact SGs are an adequate heat sink for the reactor,

• determine the nature of the accident in progress (e.g., verify an SGTR), aid

• verify unit conditions for termination of SI during secondary unit HELBs outside containment, and

• verify SG tubes are covered before terminatinq Auxiliary Feedwater to the faulted SG to assure iodine scrubbing and design basis iodine partitioning in the event of a steam gqenerator tube rupture.

Controlling SG level to maintain a heat sink and the diagnosis of a steam generator tube rupture based on SG level are operator actions assumed in the design basis accident analysis for which no automatic actuation is provided. In addition, the PRA shows that SG Wide Range Level indication can be important to safety by providing information for the initiation of operator actions to establish bleed and feed for a loss of heat sink event.

Due to the number and redundancy of indications provided by 2 channels per SG and multiple SGs, the indications for each SG are listed as individual PAM Functions on Table 3.3.3-1.

SG Water Level (Wide Range) indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods If a channel of wide range SG level instrumentation is not available, an alternate method of monitoring the SG level is a combination of B 3.3.3-18 Rev. 3.1, 12/01/05 WOG STS WOC STS B 3.3.3-18 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 one channel of SG narrow range instrumentation and the Auxiliary Feedwater flow rate indication to that SG.

B 3.3.3-19 Rev. 3.1, 12/01/05 STS WOG STS B 3.3.3-19 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

At so Me unit, operatoar actionA is based On the control roo-m indic-ationn of S-G level. The RCS Freponse during a desfign basis esmail brea;k ILOCGAl depends ORn the bhrea;-k s*e F*r a certan rn-ge of break sizes, the boiler o-d-ersGer mode'- of heat tranfer is necessary to remove decay heat. Extended startup range level is a Type A variable b-ec-ause the operator mnust manually raise and-control SG level to establitsh boiler condenser heat transfer. Operator action "isinitiated on a less of subcooled margin. Feedw-ater flow is increased until the indicated extended starup range le-el reaches the boiler condenser setpeG~t

13. Condensate Steraa-e Tank (OST-) Level CST Level is provided to ensure water supply for auxiliary foedwater (AF\ý. The CST provid-es- the ensured safet' grade water supply for the AFW Systemn. T-he CST consists Of tWO identical tanks cnnec~ted by a co~mmon- ou-tlet h~eadter. Inven~tory is monitored by a 0 inc-h to cOnAtrol room; inictortip chart recorder, and unit computer. 'n addition, a control room annunciautor alarms on low level.

At so-me units, CST- Level is. considered a Type A.variable because the control roomn mete-r .;And-anncao arecnsidered the primary ind-icOation used by the operator; The -DBAsthat require AFWA are the loss Of electric power, steam line break (91=B), and small break LOCA.

The CST is the initial sou rag of wa9ter for the AFWV System. However, as the hST is depleted, m anual eopwator acti i nceFry -s to replenish the- CST Or align suc~tion to the AFWV PUMPS from th 14, 15, 16, 717-1. Core Exit Temperature Core Exit Temperature indication is provided for verification and long term surveillance of core cooling.

An evalu'ation was made Of the minimum number of valid core exit

....... teooe (GET)- necessary for measring core cooling The e-valuation deAtermined- the-reduceG-d com;pl9emet Of CETsO n9eesary to detect initial core recover,' and tre~nd the ensuing core heatup.

The evaluations account... for core....uiformies, 'includingncre WOG STS B 3.3.3-20 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 effects of the; radal decay power distribution, e-XcoreP e-ffePc-tso cogndensate runhack in the hot legs, and nonuniorine B 3.3.3-21 Rev. 3.1. 12/01/05 WOG STS WOO STS B 3.3.3-21 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LC (co r.nti'.nued) temperatur*. Barsed ORn these eValuations, adequate core Goolg*IsI ensured moith bmo ualid The Core Exit Temperature Ghannets-peF quadrant With tWO CET... per required n-hannel. The GET pair are rientedFradially to permit evaluation of coere radial decay power distrb'tien. indication provides information for the operators to initiate RCS depressurization following a steam generator tube rupture. The PRA shows that Core Exit Temperature is. use to determffine w%÷hter tn terminate SI, if still in prOFr6e, Or to reintte 25I1if it h been stopped. Thermocouple indication is important to safety by providing information necessary to maintain subcooling for RCS cooldown and depressurization following steam generator tube rupture and other small LOCA events. It is also used as an indication for the transfer from the EOPs to the Severe Accident Management Guidance, where a greater focus is maintained on preserving the remaining fission product barriers.

Table 3.3.3-1 requires two OPERABLE channels of Core Exit Temperature. Footnote (c) to Table 3.3.3-1 requires a Core Exit Temperature is also u-sed for unit stabilization and coldo,,n control.

-we.0 OUEL A' FLcn~an~;,;nti o"ore E~xt Te emerarure ame maulmde in each quadrant to provide indication of radial distribution of the coo-wlant temperature rise across representative regions, of thea coere.

Pow rditribution symmet~' wats considered indetermininthe spcfcnmber and locations, provided forF diagnosiS Of local core problems. Therefore, two randomly selected thermocuples arel not sufficient twomcet the tWo thermocouples per chaenneral requiremt i any quadrant. The two thermocouples in each c ehannelmut mnee the additional requifement that one islocyated Rnear the -enter of the core and the othe.r near the core perimneter, such that the pair of Core Exit Temnperatures indiate the radial temperature gradient acrossr TheIr quadrant.

core Uneit pecific e o esponse to Ite~m ll.F. 2of NilUREG 0E37- (Ref. 3) should h~aveidnife the thermnocouple pairings that satisfy ths rqiemets. channel to consist of two core exit thermocouples in the nine central core rows and columns. Two sets of two thermocouples ensure that a single failure will not dis;abheaffect the ability to determine the Fadial temperature gradientwhether an inadequate core cooling condition exists.

Two OPERABLE channels of Core Exit Temperature in the nine central core rows and columns are required to provide the most timely indication of the coolant temperature rise across the core exit.

The acceptable central core exit thermocouples can be identified as B 3.3.3-22 Rev. 3.1, 12/01/05 WOG STS WOO STS B 3.3.3-22 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 beingq within the core area consisting of up to four fuel assemblies from the center fuel assembly (not counting the center assembly).

Severe accident analyses documented in WCAP-14696-A (Ref. 4) demonstrate that the coolant temperature increase at a central core location (i.e., not in the three outermost rows of fuel assemblies) provides the most rapid indication of inadequate core cooling.

Therefore, in order to get the most rapid indication of coolant temperature rise in the core, the two thermocouples in each channel used to meet the LCO requirement must be in the nine central core rows and columns.

Core Exit Temperature indication satisfies Criteria 3 and 4 of 10 CFR 50.36('c)(2)(ii).

12. Auxiliary Feedwater (AFW) Flow AFW Flow is provided to m.nitor operation Of decay heat removal via the-SG&-

The AA' FWlFow to each SG is determined from a differential pressure measurement ralikrated for a rFaRge of 0 gpn to 1200 gpm.

ReduIndaont monitoring capability is provided by tWO independnt trains Of i*nstrumeAnA-"tation for each "G.Each differential pressuFe transm;itter provides, an input to a controlI ro indcaor.an the unit computer. Since the primary indi.ation used indication is used by the operator todurn--an..... ac t is th co..ntrol, roo-m indicator,

, the PAAA specification deals specGifially with this porrion of tha instrument Gha-R-eI.

NWAI flW is-*u-,S-d thr*ee ways:

" to ;gerif' delivery of AFWA floW to the SGS,

" to deteFr-ine whether to) terminate SI ifstill in progress, in conjunction,,with SG, a...ter level (naro rne, an B 3.3.3-23 Rev. 3.1, 12/01/05 WOG STS WOG STS B 3.3.3-23 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 RASES

. to r*eulate AEr flo, go thaIt the SG tubhes remain covered.

AtI- ,,mr nt~n.f A EWA flo'wA i; Tve,, A vai,"rb'Jle beue*

• , oneratr',fr action is required to throttle flow durin an S.B accide-nt to. prev*et the-4A.FWpumnpsfrom operating in ru-nou-t conditions. AFW flow is also. used by the operator to verifyconfirm that the AFW System is in operation and delivering the correct flow to each SG. However, the primary indication used by the operator to ensure an adequate inventory is SG level. The PRA shows that AFW flow indication can be important to safety by providing information necessary for operator action to manually initiate AFW or initiate an alternate feedwater source in the event of a failure of the AFW system.

AFW Flow indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods Alternate methods of determining the need for operator action in the event that the AFW flow rate indication is not available can be provided by:

For the motor-driven AFW Pumps.

1. The AFW pump motor amperage indication; and

2. Either:

i. The AFW pump discharge pressure, or ii. The AFW flow control valve (SG supply) position indications.

For the turbine-driven AFW pump-:

1. The AFW pump discharge pressure or the steam supply valve position indication: and

2. The AFW flow control valve (SG supply) position indication.

B 3.3.3-24 Rev. 3.1, 12/01/05 WOO STS WOG STS B 3.3.3-24 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1, 2, and 3.

These variables are related to the diagnosis and pre-planned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3. In MODES 4, 5, and 6, unit conditions are such that the likelihood of an event that would require PAM instrumentation is low; therefore, the PAM instrumentation is not required to be'OPERABLE in these MODES.

The APPLICABILITY is modified by a Note that provides an exception for Power Range Neutron Flux indication in MODE 3. Power Range Neutron Flux indication is used to confirm an automatic reactor shutdown from power operation. Therefore, the Power Range Neutron Flux indication is only applicable in MODES 1 and 2 when the power range instrumentation functions to provide the necessary PAM indication (i.e., reactor trip).

ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. The Completion Time(s) of the-inoperable channel(s) of a -Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A.1 Condition A applies when one or more Functions have one required channel that is inoperable. Required Action A.1 requires restoring the inoperable channel to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a Function that has only one required channel, other non Regulatory Guide 1.97 instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

WOG STS B 3.3.3-25 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued)

B.1 Condition B applies when the Required Action and associated Completion Time for Condition A are not met. This Required Action specifies initiation of actions in Specification 5.6.5, which requires a written report to be submitted to the NRC immediately. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.

C.1 Condition C applies when one or more Functions have two inoperable required channels (i.e., two channels inoperable in the same Function).

Required Action C.1 requires restoring one channel in the Function(s) to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate- means to obtain the required information. Centinuoeu, operation '-with v-o required channels inoerable in a Fumnction; is not acceptablebeaus the- alItern-ate ind-icationsr. may not fully meet all peorfemaRG8 qualification requiremnents applied to the P.AM nsrmettin Therefore, requiring restoration o!

one ineperabhe channel of the F'unction limits the risk that the PAM Funrctio-n 1,ill be in a degrad-ed rond-ition s.hould. an accident oGc*ur.

D.1 Condition D applies when the Required Action and associated Completion Time of Condition C is not met. Required A*tion D.* requires entering the appopriate CGendition referenced in TbhlA 2-3.33 1 for the chamnnel immediately. The applicable Cond-itio*n reference-d* in the Table is Required ActionA of Coniio., and the associated Completion Time has expired, Condition D is entered for that chanmnel and provides for transfer to the appropriate subsequent ConditionContinuous operation with two or more required channels inoperable in a Function is acceptable beyond the initial 7 days provided for restoration in Condition C, provided that acceptable alternate means of monitoring are available to monitor the Function(s) with the inoperable channels and a report is submitted to the NRC.

B 3.3.3-26 Rev. 3.1. 12/01/05 STS WOG STS B 3.3.3-26 Rev. 3.1. 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued)

E.4-an&d.2 if the Required Ac*thn and associated ComFplefietIn Tipm o-f Cond*i*ti*o*n C i-notmetandTable 3.3.3 1 dfirects entrY into Condition E, the unit must be bruh oa.4MODE_ w.Ahere the requirements of this LCO_ doe not apply. To achiev,,,e t.is status, the unit Mu, t be brought to at least M*ODE 3F= w

,,ithin 6r hou---rsS and M A.4 E 0t.h *lA1-A.9 hin hoQrs.'*

1 ý:2 o The allowed Completion Times are reasonable, based on operating exporience, to FAroach the_ required unit conditions from full power coand-itions- inan orderly mannr_4 and Withut challengin untssems.

At this Limalternate moans Of MQiaigReactor Vessel ý.ate+r Level and Containmen~t ArFea Radi*ation. have been developed and tested These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time.

If these alternate means are used, the Required Action is not to shut down the unit but rather to follow the directions of Specification 5.6.5, in the Administrative Controls section of the TS. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

Condition D is modified by a Note that only allows the Condition to be used for certain PAM Functions. The Note lists the PAM Functions (by Function number) that have a preplanned alternate method of monitoring described in the Bases for that Function. Condition D is only applicable to those PAM Functions with a preplanned alternate method of monitoring described in the Bases for the Function. Condition E is applicable for PAM Functions without a preplanned alternate method of monitoring described in the Bases for the Function.

WOG STS B 3.3.3-27 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued)

E.1 and E.2 Ifthe Required Action and associated Completion Time of Condition C is not met, the unit must be brought to a MODE where the requirements of this LCO do not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Condition E is modified by a Note that only allows the Condition to be used for certain PAM Functions. The Note states that Condition E is only applicable to those PAM Functions not addressed by Condition D. As such, Condition E is only applicable to those PAM Functions without a preplanned alternate method of monitoring described in the Bases for the Function. Condition D is applicable for PAM Functions with a preplanned alternate method of monitoring described in the Bases for the Function.

SURVEILLANCE A Note has been added to-the SR Table to clarify that.SR 3.3.3.1-aid REQUIREMENTS SR 3.3.3.3 7 ,.,ap.,ppDles to each PAM instrumentation Function in Table 3.3.3-1 and that SR 3.3.3.2 applies to each PAM instrument Function in Table 3.3.3-1 except for the Penetration Flow Path CIV Position Function. In addition, the Note specifies that SR 3.3.3.3 applies only to the Penetration Flow Path CIV Position Function.

SR 3.3.3.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar unit instruments located throughout the unit.

B 3.3.3-28 Rev. 3.1, 12/01/05 STS WOG STS B 3.3.3-28 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued)

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.

The Frequency of 31 days is based on operating experience that demonstrates that channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels.

SR 3.3.3.2 A CHANNEL CALIBRATION is performed every [18] months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS)

Instrumentation." Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The Frequency is based on operating experience and consistency with the typical industry refueling cycle.

SR 3.3.3.3 A TADOT is required for the Penetration Flow Path CIV Position Function.

This test is required to be performed [at least once every 18-months, or approximately at every refuelinqi. The TADOT verifies the OPERABILITY of the required containment isolation valve position indication instrumentation.

A Note modifies the Surveillance Requirements and specifies that the TADOT surveillance is only applicable to the Penetration Flow Path CIV Position Function. Due to the relatively simple instrument circuits involved and the lack of a conventional process sensor to adiust, the TADOT. rather than the CHANNEL CALIBRATION. Drovides the more WOG STS B 3.3.3-29 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 PAM Instrumentation B 3.3.3 appropriate OPERABILITY verification of these channels.

The Frequency of 18-months is consistent with the typical industry refueling cycle.

REFERENCES f-1i. Uit specific doc..um.ent (e.g., FSAR, NRC Regulatory Guide 1.97 SR Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Followinq an Accident," December 1980..

2. Rogulatory Guide 1.97, [date].

32. NUREG-0737, Supplement 1, "TMI Action Items."

3. WCAP-1 5981-NP-A, Revision 0, "Post Accident. Monitorinq Instrumentation Re-Definition for Westinghouse NSSS Plants,"

September 2008."

4. WCAP-14696-A, Revision 1, "Westinghouse Owners Group Core Damaae Assessment Guidance."

......... l .........................

WOG STS B 3.3.3-30 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation 3.3.3 3.3 INSTRUMENTATION 3.3.3 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3 The PAM instrumentation for each Function in Table 3.3.3-1 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

---

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

The Power Range Neutron Flux PAM Function is not required in MODE 3.

ACTIONS

---

NOTE Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions A.1 Restore required channel to 30 days with one required OPERABLE status.

channel inoperable.

B. Required Action and B.1 Initiate action in accordance Immediately associated Completion with Specification 5.6.5.

Time of Condition A not met.

C. One or more Functions C.1 Restore one channel to 7 days with two required OPERABLE status.

channels inoperable.

---

- NOTE -- -------

Condition D is only D.1 Initiate action in accordance Immediately applicable to PAM Functions with Specification 5.6.5.

[1,4, 8, 10, and 12] in Table 3.3.3-1.

D. Required Action and associated Completion*

Time of Condition C not met.

WOG STS 3.3.3-1 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation 3.3.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

-NOTE Condition E is only 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 /> applicable to those PAM Functions in Table 3.3.3-1 AND not addressed by Condition D. E.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 /> E. Required Action and associated Completion Time of Condition C not met.

SURVEILLANCE REQUIREMENTS NO r------------------------------------------------------------

SR 3.3.3.1 applies to each PAM instrumentation Function in Table 3.3.3-1. SR 3.3.3.2 applies to each PAM instrumentation Function in Table 3.3.3-1 except for the Penetration Flow Path Containment Isolation Valve Position. SR 3.3.3.3 applies only to the Penetration Flow Path Containment Isolation Valve Position.

SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.

SR 3.3.3.2 - ------------------ NOTE ---------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION. [18] months SR 3.3.3.3 Perform TADOT. [18] months WOG STS 3.3.3-2 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation 3.3.3 Table 3.3.3-1 (page 1 of 1)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS

1. Power Range Neutron Flux 2

2. Steam Generator (SG) Pressure a) SG "A" 2 b) SG "B" 2 c) SG "C" 2 d) SG "D" 2

3. Refueling Water Storage Tank Level (Wide Range) 2

4. High Head Safety Injection (SI) Flow 2 (1 per SI train)

5. Reactor Coolant System Pressure (Wide Range) 2

6. Containment Pressure (Wide Range) 2

7. Penetration Flow Path Containment Isolation Valve 2 per peneraon flow Position path a

8. Containment Area Radiation (High Range) 2

9. Pressurizer Level 2

10. SG Water Level (Wide Range) a) SG "A" 2 b) SG "B" 2 c) SG "C" 2 d) SG "D" 2

11. Core Exit Temperature 2 (c)

12. Auxiliary Feedwater (AFW) Flow 2 (1 rer AFW train)

(a) Not required for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

(b) Only one position indication channel is required for penetration flow paths with only one installed control room indication channel.

(c) A channel consists of two core exit thermocouples (CETs) in the nine central core rows and columns.

---

REVIEWERS NOTE-----------------------------------

Table 3.3.3-1 shall be amended for each unit as necessary to list the plant specific instrument Functions identified by the application of the methodology contained in WCAP-15981-P-A, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants".

WOG STS .3.3.3-3 Rev. 3.1, .12/01/05

Retyped Version for Referen'ce TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 B 3.3 INSTRUMENTATION B 3.3.3 Post Accident Monitoring (PAM) Instrumentation BASES BACKGROUND The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Accidents (DBAs) or that the Probabilistic Risk Assessment (PRA) has shown to be significant to the public health and safety.

The OPERABILITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and to assess unit status and behavior following an accident.

The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed and the need for, and magnitude of, further actions can be determined. These essential instruments are identified by addressing the recommendations of Regulatory Guide 1.97 (Ref. 1) as required by Supplement 1 to NUREG-0737 (Ref. 2) and by evaluating the instrumentation consistent with the methodology contained in WCAP-1 5981 (Ref. 3). This methodology considers the use of the accident monitoring instrumentation in the PRA, Emergency Operating Procedures (EOPs), Severe Accident Management Guidance (SAMG) procedures, and Emergency Plan (E-Plan).

The control room monitoring instrumentation Functions required to be OPERABLE have been evaluated and selected in accordance with the screening criteria contained in WCAP-15981. The screening criteria were used to identify the PAM instrumentation important to safety (i.e., monitor plant parameters that are the basis for important operator actions to bring the unit to a safe stable state in the event of an accident).

B 3.3.3-1 Rev. 3.1, 12/01/05 WOG STSSTS B 3.3.3-1 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES The PAM specification ensures the operability of instrumentation to monitor plant parameters necessary for safety significant operator actions so that the control room operating staff can:

• Perform the diagnosis specified in the EOPs (these variables are restricted to preplanned actions for the primary success path of DBAs), e.g., loss of coolant accident (LOCA),

• Take the specified, pre-planned, manually controlled actions, for which no automatic control is provided, and that are required for safety systems to accomplish their safety function,

• Implement procedures or guidance that has been shown to have an important role in preventing core damage or early fission product releases,

• Determine the likelihood of a gross breach of the barriers that prevent radioactivity release,

• Determine if a gross breach of a barrier has occurred, and B 3.3.3-2 Rev. 3.1, 12/01/05 WOG STS WOG STS B 3.3.3-2 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES APPLICABLE SAFETY ANALYSES (continued)

• Initiate action necessary to protect the public and to estimate the magnitude of any impending threat..

The PAM instrumentation selected in accordance with WCAP-15981 is used to monitor plant parameters necessary for safety significant operator actions and satisfies Criterion 3 and/or 4 of .10 CFR 50.36(c)(2)(ii).

LCO The PAM instrumentation LCO provides OPERABILITY requirements for the control room monitoring instrumentation Functions important to safety (i.e., monitor plant parameters that are the basis for important operator actions to bring the plant to a safe stable state in the event of an accident).

The OPERABILITY of the PAM instrumentation ensures there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97 (Ref. 1) and the guidance provided in WCAP-15981 (Ref. 3) for selecting the appropriate instrumentation.

LCO 3.3.3 requires two OPERABLE channels for most Functions. Two OPERABLE channels ensure no single failure prevents operators from getting the information necessary for them to determine the safety status of the unit, and to bring the unit to and maintain it in a safe condition following an accident. Therefore where plant design permits, the two channels required OPERABLE by the LCO should be supplied from different trains of electrical power.

Furthermore, OPERABILITY of two channels allows a CHANNEL CHECK during the post accident phase to confirm the validity of displayed information.

The exception to the two channel requirement is Penetration Flow Path Containment Isolation Valve (CIV) Position. In this case, the important information is the status of the containment penetrations. The LCO requires one position indicator for each active CIV. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve and prior knowledge of a passive B 3.3.3-3 Rev. 3.1, 12/01/05 WOG STS WOG STS B-3.3.3-3 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued) valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for valves in this state is not required to be OPERABLE.

Table 3.3.3-1 provides a list of the control room indications identified as important to safety in accordance with the methodology of WCAP-15981 (Ref. 3). The monitoring functions specified in Table 3.3.3-1 are required OPERABLE by LCO 3.3.3. The following discussions describe the instrument monitoring Functions identified in Table 3.3.3-1.

---

Reviewer's Note ---------------------------------------

Listed below are discussions of the specified instrument Functions listed in Table 3.3.3-1. These discussions are intended as examples of what should be provided for each Function when the unit specific list is prepared.

1. Power Rangqe Neutron Flux The Power Range Neutron Flux indication is used to confirm a reactor shutdown following an accident or other receipt of a reactor trip. The PRA shows that operator actions to manually shutdown the reactor in the event of a failure of the automatic actions, as determined from the Power Range Neutron Flux indication, can be important to safety.

The Power Range Neutron Flux indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

The Applicability is modified by a Note that provides an exception to the OPERABILITY requirement for Power Range Neutron Flux indication in MODE 3. The Power Range Neutron Flux indication is used to confirm an automatic reactor shutdown from power operation. Therefore, the PAM Power Range Neutron Flux indication requirement is only applicable in MODES 1 and 2 when the power range instrumentation functions to provide the PAM indication (i.e.,

reactor trip).

Alternate Monitoringq Methods If the power range neutron flux indications are not available, an alternate method of verifying a reactor trip is a combination of:

1. A neutron flux indication:

i. The intermediate range neutron flux indication, or WOG STS B 3.3.3-4 Rev. 3.1. 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 ii. source range neutron flux indication; and

2. A rod position indication:

i. The rod bottom lights, or ii. The rod position indicators.

These alternate indications can also provide the information necessary for operators to determine the need to initiate a manual reactor trip.

2. Steam Generator (SG) Pressure SG Pressure indication is used following an accident or receipt of a reactor trip signal to indicate secondary side integrity. It is also used to indicate the target pressure for RCS depressurization following a SG tube rupture accident to terminate the RCS inventory loss. In the event of a SG tube rupture accident, the EOPs instruct the operators to depressurize the RCS to a pressure below the secondary side pressure in the ruptured SG. RCS depressurization to a pressure less than the SG pressure terminates the RCS inventory loss and terminates the SG inventory gain, preventing overfill of the SG. The termination of the break flow is an operator action assumed in the design basis SG tube rupture analysis for which no automatic action is provided. The PRA shows that failure to depressurize the RCS to a pressure less than the secondary side pressure in the ruptured SG is a risk significant operator action.

Due to the number and redundancy of indications provided by 2 channels per SG and multiple SGs, the indications for each SG are listed as individual PAM Functions on Table 3.3.3-1.

SG Pressure indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

B 3.3.3-5 Rev. 3.1, 12/01/05 WOG STS WOC STS B 3.3.3-5 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523,.Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

3. Refueling Water Storage Tank Level (Wide Range)

Refueling Water Storage Tank (RWST) Level provides an indication of the water inventory remaining for use by containment spray and safety injection for core cooling and containment cooling. [No operator actions in the design basis accident analysis are based on the RWST level indication. The switchover from the RWST to the containment sump is performed automatically.]

The PRA shows that in the event of an accident in which the RCS inventory losses are outside of containment (e.g., SG tube rupture and interfacing system LOCA), the remaining RWST level is an important indication in the EOPs for choosing the appropriate operator actions to maintain core cooling. The PRA shows the importance of diagnosing the need for implementing RWST refill to maintain a sufficient inventory for long term core cooling following these events.

RWST Level indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

B 3.3.3-6 Rev. 3.1, 12/01/05 WOG STS WOG STS B 3.3.3-6 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

4. High Head Safety Injection (SI) Flow High Head SI Flow indication is used to confirm automatic safety injection initiation following a design basis accident. Therefore, the required flow indicator for this PAM Function is the one installed in the automatic High Head SI flow path. The results of the PRA show that this is a risk significant operator action. Failure to manually initiate SI flow when the automatic initiation fails can lead to a significant increase in core damage frequency. The operator action is based 6n the Emergency Core Cooling System flow indication in the control room. The PRA shows that only high head safety injection is important for all accident sequences except the unlikely double-ended guillotine rupture of the largest reactor coolant pipe.

Therefore, only the High Head SI Flow indication is required.

High Head SI Flow indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods If the High Head SI Flow indication is not available, an alternate method of verifying SI initiation can be provided by either:

1. The High Head SI pump motor amperage and the High Head SI pump discharge pressure indications; or

2. The header pressure and the automatic SI valve position indications.

5. Reactor Coolant System Pressure (Wide Range)

The PRA indicates that Reactor Coolant System (RCS) pressure is a variable important to safety for RCS cooldown and depressurization following a SG tube rupture. RCS Wide Range Pressure indication provides the information necessary for RCS depressurization for the SG tube rupture accident to terminate the RCS inventory loss. In the event of a SG tube rupture accident, the EOPs instruct the operators to depressurize the RCS to a pressure below the secondary side pressure in the ruptured SG. RCS depressurization to a pressure less than the SG pressure terminates the RCS inventory loss and terminates the SG inventory gain, preventing overfill of the SG. The termination of the break flow is an operator action assumed in the design basis SG tube rupture analysis for which no automatic action is provided. RCS pressure is also used for operator action to terminate SI in the event of a steamline break to prevent pressurizer overfill for which no automatic actuation is provided. Additionally, WOG STS B 3.3.3-7 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 RCS pressure is used for determining RCS subcooling following an accident or receipt of a reactor trip signal.

RCS pressure indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

6 WOG STS B 3.3.3-8 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

6. Containment Pressure (Wide Ranqe)

Containment Pressure (Wide Range) indication is provided for assessing containment cooling and containment integrity. No operator actions in the design basis accident analysis are based on the containment pressure indication. Containment pressure is an indicator of the potential loss of a fission product boundary in the Emergency Action Levels in the E-Plan. Containment pressure is a key indicator in the declaration of a General Emergency level and the potential need for offsite radiological protection actions. Containment pressure may also be used in post accident conditions to determine when to vent the containment to prevent overpressurization.

Containment Pressure (Wide Range) indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

7. Penetration Flow Path Containment Isolation Valve Position Penetration Flow Path CIV Position is provided for verification of.

Containment OPERABILITY, and Phase A and Phase B isolation.

The CIV position indication provides a direct indication of a failure to completely isolate the containment following the receipt of a containment isolation signal. The E-Plan identifies that an elevated emergency action level should be declared following an accident in the event of a failure of the automatic containment isolation.

When used to verify Phase A and Phase B isolation, the important information is the isolation status of the containment penetrations.

The LCO requires one channel of valve position indication in the control room to be OPERABLE for each active CIV in a containment penetration flow path, i.e., two total channels of CIV position indication for a penetration flow path with two active valves. For containment penetrations with only one active CIV having control room indication, Note (b) requires a single channel of valve position indication to be OPERABLE. This is sufficient to redundantly verify the isolation status of each isolable penetration either via indicated status of the active valve, as applicable, and prior knowledge of a passive valve, or via system boundary status. If a normally active CIV is known to be closed and deactivated, position indication is not needed to determine status. Therefore, the position indication for' valves in this state is not required to be OPERABLE. Note (a) to the Required Channels states that the Function is notrequired for isolation valves whose associated penetration is isolated by at least one closed and deactivated automatic valve, closed manual valve, blind flange, or check valve with flow through the valve secured.

WOG STS B 3.3.3-9 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 Each penetration is treated separately and each penetration flow path is considered a separate function. Therefore, separate Condition entry is allowed for each inoperable penetration flow path.

CIV Position indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

B 3.3.3-10 Rev. 3.1, 12/01/05 STS WOG STS B 3.3.3-10 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 -

PAM Instrumentation B 3.3.3 BASES LCO (continued)

8. Containment Area Radiation (High Range)

Containment Area Radiation High Range provides an indication of a loss of one or more fission product barriers. The Emergency Action Levels in the E-Plan utilize the Containment Area Radiation High Range monitor as anindication of the potential loss of one or more fission product barriers in the assessment of the declaration of a General Emergency level and the potential need for offsite radiological protection actions. The post accident Core Damage Assessment also uses the Containment Area Radiation High Range monitor as an input to the determination of core damage.

The Containment Area Radiation High Range indication satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods A portable radiation monitor (with appropriate multiplier if necessary) can be used as an alternate method of indication for Containment Area Radiation High Range.

9. Pressurizer Level Pressurizer Level indication is used for the SI termination criteria to prevent pressurizer overfill. The termination of SI to prevent pressurizer overfill is an operator action assumed in the design basis steamline break analysis for which no automatic actuation is provided. The PRA also indicates that SI termination in the event of a steam generator tube rupture is required for long term core cooling.

Pressurizer Level indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

10. SG Water Level (Wide Range)

SG Water Level (Wide Range) indication is provided to monitor operation of decay heat removal via the SGs.

SG Water Level (Wide Range) is used to:

• identify the faulted SG following a tube rupture,

• verify that the intact SGs are an adequate heat sink for the reactor, WOG STS B 3.3.3-11 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 determine the nature of the accident in progress (e.g., verify an SGTR),

verify unit conditions for termination of SI during secondary unit HELBs outside containment, and

• verify SG tubes are covered before terminating Auxiliary Feedwater to the faulted SG to assure iodine scrubbing and design basis iodine partitioning in the event of a steam generator tube rupture.

Controlling SG level to maintain a heat sink and the diagnosis of a steam generator tube rupture based on SG level are operator actions assumed in the design basis accident analysis for which no automatic actuation is provided. In addition, the PRA shows that SG Wide Range Level indication can be important to safety by providing information for the initiation of operator actions to establish bleed and feed for a loss of heat sink event.

Due to the number and redundancy of indications provided by 2 channels per SG and multiple SGs, the indications for each SG are listed as individual PAM Functions on Table 3.3.3-1.

SG Water Level (Wide Range) indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods If a channel of wide range SG level instrumentation is not available, an alternate method'of monitoring the SG level is a combination of one channel of SG narrow range instrumentation and the Auxiliary Feedwater flow rate indication to that SG.

B 3.3.3-12 Rev. 3.1. 12/01/05 WOC STS WOG STS B 3.3.3-12 Rev. 3.1. 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

11. Core Exit Temperature Core Exit Temperature indication is provided for verification and long term surveillance of core cooling. The Core Exit Temperature indication provides information for the operators to initiate RCS depressurization following a steam generator tube rupture. The PRA shows that Core Exit Thermocouple indication is important to safety by providing information necessary to maintain subcooling for RCS cooldown and depressurization following steam generator tube rupture and other small LOCA events. It is also used as an indication for the transfer from the EOPs to the Severe Accident Management Guidance, where a greater focus is maintained on preserving the remaining fission product barriers.

Table 3.3.3-1 requires two OPERABLE channels of Core Exit Temperature. Footnote (c) to Table 3.3.3-1 requires a Core Exit Temperature channel to consist of two core exit thermocouples in the nine central core rows and columns. Two sets of two thermocouples ensure that a single failure will not affect the ability to determine whether an inadequate core cooling condition exists.

Two OPERABLE channels of Core Exit Temperature in the nine central core rows and columns are required to provide the most timely indication of the coolant temperature rise across the core exit.

The acceptable central core exit thermocouples can be identified as being within the core area consisting of up to four fuel assemblies from the center fuel assembly (not counting the center assembly).

Severe accident analyses documented in WCAP-14696-A (Ref. 4) demonstrate that the coolant temperature increase at a central core location (i.e., not in the three outermost rows of fuel assemblies) provides the most rapid indication of inadequate core cooling.

Therefore, in order to get the most rapid indication of coolant temperature rise in the core, the two thermocouples in each channel used to meet the LCO requirement must be in the nine central core rows and -columns.

Core Exit Temperature indication satisfies Criteria 3 and 4 of 10 CFR 50.36(c)(2)(ii).

12. Auxiliary Feedwater (AFW) Flow AFW Flow indication is used by the operator to confirm that the AFW System is in operation and delivering the correct flow to each SG.

However, the primary indication used by the operator to ensure an adequate inventory is SG level. The PRA shows that AFW flow indication can be important to safety by providing information WOG STS B 3.3.3-13 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 necessary for operator action to manually initiate AFW or initiate an alternate feedwater source in the event of a failure of the AFW system.

AFW Flow indication satisfies. Criterion 4 of 10 CFR 50.36(c)(2)(ii).

Alternate Monitoring Methods Alternate methods of determining the need for operator action in the event that the AFW flow rate indication is not available can be provided by:

For the motor-driven AFW Pumps.

1. The AFW pump motor amperage indication; and

2. Either:

i. The AFW pump discharge pressure, or ii. The AFW flow control valve (SG supply) position indications.

For the turbine-driven AFW pump:

1. The AFW pump discharge pressure or the steam supply valve position indication; and

2. The AFW flow control valve (SG supply) position indication.

B 3.3.3-14 Rev. 3.1, 12/01/05 WOG STS WOO STS B 3.3.3-14 Rev. 3.1,12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES LCO (continued)

APPLICABILITY The PAM instrumentation LCO is applicable in MODES 1, 2, and 3.

These variables are related to the diagnosis and pre-planned actions required to mitigate DBAs. The applicable DBAs are assumed to occur in MODES 1, 2, and 3. In MODES 4, 5, and 6, unit conditions are such that the likelihood of an event that would require PAM instrumentation is low; therefore, the PAM instrumentation is not required to be OPERABLE in these MODES.

The APPLICABILITY is modified by a Note that provides an exception for Power Range Neutron Flux indication in MODE 3. Power Range Neutron Flux indication is used to confirm an automatic reactor shutdown from power operation. Therefore, the Power Range Neutron Flux indication is only applicable in MODES 1 and 2 when the power range instrumentation functions to provide the necessary PAM indication (i.e., reactor trip).

ACTIONS A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed on Table 3.3.3-1. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A.1 Condition A applies when one or more Functions have one required channel that is inoperable. Required Action A.1 requires restoring the inoperable channel to OPERABLE status within 30 days. The 30 day Completion Time is based on operating experience and takes into account the remaining OPERABLE channel (or in the case of a Function that has only one required channel, other instrument channels to monitor the Function), the passive nature of the instrument (no critical automatic action is assumed to occur from these instruments), and the low probability of an event requiring PAM instrumentation during this interval.

WOG STS B 3.3.3-15 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued)

B.1 Condition B applies when the Required Action and associated Completion Time for Condition A are not met. This Required Action specifies initiation of actions in Specification 5.6.5, which requires a written report to be submitted to the NRC immediately. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. This action is appropriate in lieu of a shutdown requirement since alternative actions are identified before loss of functional capability, and given the likelihood of unit conditions that would require information provided by this instrumentation.

C..1 Condition C applies when one or more Functions have two inoperable required channels (i.e., two channels inoperable in the same Function).

Required Action C.1 requires restoring one channel in the Function(s) to OPERABLE status within 7 days. The Completion Time of 7 days is based on the relatively low probability of an event requiring PAM instrument operation and the availability of alternate means to obtain the required information.

D.1 Condition D applies when the Required Action and associated Completion Time of Condition C is not met. Continuous operation with two or more required channels inoperable in a Function is acceptable beyond the initial 7 days provided for restoration in Condition C, provided that acceptable alternate means of monitoring are available to monitor the Function(s) with the inoperable channels and a report is submitted to the NRC.

BASES ACTIONS (continued)

These alternate means may be temporarily installed ifthe normal PAM channel cannot be restored to OPERABLE status within the allotted time.

If these alternate means are used, the Required Action is not to shut down the unit but rather to follow the directions of Specification 5.6.5, in the Administrative Controls section of the TS. The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

WOG STS B 3.3.3-16 Rev. 3.1, 12/01/05

Retyped Version for Reference , TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 Condition D is modified by a Note that only allows the Condition to be used for certain PAM Functions. The Note lists the PAM Functions (by Function number) that have a preplanned alternate method of monitoring described in the Bases for that Function. Condition D is only applicable to those PAM Functions with a preplanned alternate method of monitoring described in the Bases for the Function. Condition E is applicable for PAM Functions without a preplanned alternate method of monitoring described in the Bases for the Function.

B 3.3.3-17 Rev. 3.1, 12/01/05 STS WOG STS B 3.3.3-17 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES ACTIONS (continued)

E.1 and E.2 If the Required Action and associated Completion Time of Condition C is not met, the unit must be brought to a MODE where the requirements of this LCO do not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

Condition E is modified by a Note that only allows the Condition to be used for certain PAM Functions. The Note states that Condition E is only applicable to those PAM Functions not addressed by Condition D. As such, Condition E is only applicable to those PAM Functions without a preplanned alternate method of monitoring described in the Bases for the Function. Condition D is applicable for PAM Functions with a preplanned alternate method of monitoring described in the Bases for the Function.

SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 REQUIREMENTS applies to each PAM instrumentation Function in Table 3.3.3-1 and that SR 3.3.3.2 applies to each PAM instrument Function in Table 3.3.3-1 except for the Penetration Flow Path CIV Position Function. In addition, the Note specifies that SR 3.3.3.3 applies only to the Penetration Flow Path CIV Position Function.

SR 3.3.3.1 Performance of the CHANNEL CHECK once every 31 days ensures that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar unit instruments located throughout the unit.

B 3.3.3-18 Rev. 3.1, 12/01/05 WOG STS WOG STS B 3.3.3-18 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 BASES SURVEILLANCE REQUIREMENTS (continued)

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.

The Frequency of 31 days is based on operating experience that demonstrates that channel failure is rare. The CHANNEL CHECK supplements less .formal,,but more frequent, checks of channels during normal operational use of the displays associated with the LCO required channels.

SR 3.3.3.2 A CHANNEL CALIBRATION is performed every [18] months, or approximately at every refueling. CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. This SR is modified by a Note that excludes neutron detectors. The calibration method for neutron detectors is specified in the Bases of LCO 3.3.1, "Reactor Trip System (RTS)

Instrumentation." Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION of the Core Exit thermocouple sensors is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

The Frequency is based on operating experience and consistency with the typical industry refueling cycle.

SR 3.3.3.3 A TADOT is required for the Penetration Flow Path CIV Position Function.

This test is required to be performed [at least once every 18-months, or approximately at every refueling]. The TADOT verifies the OPERABILITY of the required containment isolation valve position indication instrumentation.

A Note modifies the Surveillance Requirements and specifies that the TADOT surveillance is only applicable to the Penetration Flow Path CIV Position Function. Due to the relativeiy simple instrument circuits involved and the lack of a conventional process sensor to adjust, the TADOT, rather than the CHANNEL CALIBRATION, provides the more WOG STS B 3.3.3-19 Rev. 3.1, 12/01/05

Retyped Version for Reference TSTF-523, Rev. 0 PAM Instrumentation B 3.3.3 appropriate OPERABILITY verification of these channels.

The Frequency of 18-months is consistent with the typical industry refueling cycle.

REFERENCES

1. Regulatory Guide 1.97, Revision 2, "Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," December 1980..

2. NUREG-0737, Supplement 1, "TMI Action Items."

3. WCAP-15981-NP-A, Revision 0, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants,"

September 2008."

4. WCAP-14696-A, Revision 1, "Westinghouse Owners Group Core Damage Assessment Guidance."

WOG STS B 3.3.3-20 Rev. 3.1, 12/01/05

TSTF-525, Rev. 0 Model Application for Adoption of TSTF-525

TSTF-525, Rev. 0

[DATE] 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

PLANT NAME DOCKET NO. 50-[xxx]

APPLICATION TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-525, "POST ACCIDENT MONITORING INSTRUMENTATION REQUIREMENTS (WCAP- 15981-NP-A)"

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, [LICENSEE] is submitting a request for an amendment to the Technical Specifications (TS) for [PLANT NAME, UNIT NOS.].

The proposed amendment would modify TS requirements regarding post accident monitoring instrumentation as described in TSTF-525, Revision 0, "Post Accident Monitoring Instrumentation Requirements (WCAP- 15981-NP-A)."

Attachment I provides a description and assessment of the proposed changes. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages for information marked to show the proposed changes.

Approval of the proposed amendment is requested by [date]. Once approved, the amendment shall be implemented within [ ] days.

In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being provided to the designated [STATE] Official.

Page I

TSTF-525, Rev. 0

[In accordance with 10 CFR 50.30(b), a license amendment request must be executed in a signed original under oath or affirmation. This can be accomplished by attaching a notarized affidavit confirming the signature authority of the signatory, or by including the following statement in the cover letter: "I declare under penalty of perjury that the foregoing is true and correct.

Executed on (date)." The alternative statement is pursuant to 28 USC 1746. It does not require notarization.]

If you should have any questions regarding this submittal, please contact [NAME, TELEPHONE NUMBER].

Sincerely,

[Name, Title]

Attachments: I. Description and Assessment

2. Proposed Technical Specification Changes (Mark-Up)

3. Revised Technical Specification Pages

4. Proposed Technical Specification Bases Changes (Mark-Up)

5. Justification for Alternate PAM Instrumentation cc: NRC Project Manager NRC Regional Office NRC Resident Inspector State Contact Page 2

TSTF-525, Rev. 0 ATTACHMENT 1 - DESCRIPTION AND ASSESSMENT

1.0 DESCRIPTION

The proposed change revises Technical Specification 3.3.3, "Post Accident Monitoring (PAM)

Instrumentation," to revise the PAM instrumentation included in Table 3.3.3-1 of Technical Specification 3.3.3 and to add the use of alternate monitoring instruments for certain PAM instrument Functions. In addition, a new Surveillance Requirement is added to Technical Specification 3.3.3. Table 3.3.3-1 in Technical Specification 3.3.3 is revised, as well as the associated Actions and Surveillance Requirements.

The proposed amendment is consistent with TSTF-525, Revision 0, "Post Accident Monitoring Instrumentation Requirements (WCAP- 15981 -NP-A)."

2.0 ASSESSMENT 2.1 Applicability of Published Safety Evaluation

[LICENSEE] has reviewed the model safety evaluation dated [DATE] as part of the Federal Register Notice of Availability. This review included a review of the NRC staff's evaluation, as well as the information provided in TSTF-525. [As described in the subsequent paragraphs,

][LICENSEE] has concluded that the justifications presented in the TSTF-525 proposal and the model safety evaluation prepared by the NRC staff are applicable to [PLANT, UNIT NOS.] and justify this amendment for the incorporation of the changes to the [PLANT] TS.

2.2 Optional Changes and Variations

[LICENSEE is not proposing any variations or deviations from the TS changes described in the TSTF-525, Revision 0, or the applicable parts of the NRC staff's model safety evaluation dated

[DATE].] [LICENSEE is proposing the following variations from the TS changes described in the TSTF-525, Revision 0, or the applicable parts of the NRC staff's model safety evaluation dated [DATE].]

[The [PLANT] TS utilize different [numbering][and][titles] than the Standard Technical Specifications on which TSTF-525 was based. Specifically, [describe differences between the plant-specific TS numbering and/or titles and the TSTF-525 numbering and titles.] These differences are administrative and do not affect the applicability of TSTF-525 to the [PLANT]

TS.]

[The [PLANT] TS include the adoption of changes in TSTF-425-A, Rev. 3 which relocate Surveillance Frequencies to a licensee controlled document. Therefore, the Surveillance Frequency of proposed SR 3.3.3.3 is "In accordance with the Surveillance Frequency Control Program" instead of 18 months. This change, although not specifically included in TSTF-425, is consistent with the changes approved in that Traveler.]

Page 3

TSTF-525, Rev. 0 2.3 Licensee Verifications and Plant-Specific Information

[LICENSEE] has utilized the methodology contained in WCAP-15981-NP-A, Revision 0, "Post Accident Monitoring Instrumentation Re-Definition for Westinghouse NSSS Plants," dated September 2008, and confirmed that the generic conclusions in WCAP-15981-NP-A and TSTF-525 are applicable to [PLANT].

Attachment 5, "Justification for Alternate PAM Instrumentation," provides a plant-specification justification for use of alternate PAM instrumentation, as discussed in the Safety Evaluation for WCAP-15981-NP-A.

3.0 REGULATORY ANALYSIS

3.1 No Significant Hazards Consideration Determination

[LICENSEE] requests adoption of TSTF-525, Revision 0, "Post Accident Monitoring Instrumentation Requirements (WCAP- 15981-NP-A)," which is an approved change to the standard technical specifications (STS), into the [PLANT NAME, UNIT NOS] technical specifications (TS). The proposed change revises Technical Specification 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," to revise the PAM instrumentation included in Table 3.3.3-1 of Technical Specification 3.3.3 and to add the use of alternate monitoring instruments for certain PAM instrument Functions. In addition, a new Surveillance Requirement isadded to Technical Specification 3.3.3. Table 3.3.3-1 in Technical Specification 3.3.3 is revised, as well as the associated Actions and Surveillance Requirements.

[LICENSEE] has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change revises the instruments, Actions, and Surveillance Requirements for PAM instrumentation. PAM instrumentation is not an initiator to any accident previously evaluated, and as a result, the proposed change does not involve a significant increase the probability of an analyzed event. The PAM instruments provide information for manual operator actions for which no automatic control is provided and, as a result, function to mitigate the consequences of an accident. The proposed change retains in the Technical Specifications the PAM instrumentation functions needed for manual operator actions for which no automatic control is provided. The proposed'change to the Actions is administrative and does not affect the actions taken when a PAM instrument function is not operable. The proposed change to the Surveillance Requirement adds a new testing requirement to the containment isolation valve position indication function which is a more appropriate test to confirm the function is operable. As a result, the proposed change does not affect the ability of the PAM instrumentation to provide information to the operator to mitigate an accident.

Page 4

TSTF-525, Rev. 0 Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change revises the instruments, Actions, and Surveillance Requirements for PAM instrumentation. The proposed change does not involve a physical alteration of the plant (i.e., no new or different type of equipment will be installed) or a change in the methods governing normal plant operation. In addition, the changes do not impose any new or different requirements that could initiate an accident. The changes do not alter assumptions made in the safety analysis. The proposed changes are consistent with the safety analysis assumptions.

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change revises the instruments, Actions, and Surveillance Requirements for PAM instrumentation. The proposed change does not alter the manner in which safety limits, limiting safety system settings, or limiting conditions for operation are determined. The safety analysis acceptance criteria are not affected by these changes. The proposed change will not revise the indication provided by the affected instruments, and all operator actions based on these indications that are credited in the safety analyses will remain the same. As such, the proposed change will not result in plant operation in a configuration outside the design basis or assumptions of the design basis accident analyses.

Therefore, it is concluded that this change does not involve a significant reduction in a margin of safety.

Based on the above, [LICENSEE] concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.0 ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Page 5

TSTF-525, Rev. 0 Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.

Page 6

TSTF-525, Rev. 0 - Proposed Technical Specification Changes (Mark-Up)

Page 7

TSTF-525, Rev. 0 - Revised Technical Specification Pages Page 8

TSTF-525, Rev. 0 - Proposed Technical Specification Bases Changes (Mark-Up)

(For Information Only)

Page 9

TSTF-525, Rev. 0 Attachment 5 - Justification for Alternate PAM Instrumentation

[Preparer's Note: The following generic discussion should be revised on a plant-specific basis to justify the use of alternate PAM Instrumentation.]

,The following discussion contains information regarding [PLANT] procedures and design that justify the use of alternate PAM Instrumentation.

The use of alternate instrumentation is acceptable because it takes into account the passive nature of the PAM Function and that there are alternate Functions available to accomplish the required PAM Function, and the low likelihood that the PAM Function would be required during the time that an alternate method ofmonitoring the PAM Function is utilized. The allowance to temporarily rely on an alternate PAM Function is also acceptable because in most cases, the alternate Functions identified for a given PAM Function are Regulatory Guide 1.97 qualified or associated with a safety related system with a reliable power supply (e.g., pump or header discharge pressure or motor amperes) and are displayed in the main control room. Additionally, the alternate indications that may not be Regulatory Guide 1.97 qualified (e:g., rod bottom lights, rod position indicators, pump/header pressures, and motor amperage) may also be utilized in the EOPs to confirm protective system actuation.

The following information refers to the [PLANT] Emergency Response Guidelines, including the Executive Volume and the Background Information for Individual Guidelines:

• SG Water Level (Wide Range)

Wide Range SG Level is used as part of the bleed and feed initiation parameter in FR-H.1, "Response to Loss of Secondary Heat Sink," to determine if SG inventory is approaching a dryout condition. Note that a dryout condition would not be approached if (minimum) SG feed flow is indicated or level is in the narrow range. When performing FR-H.1, if feed flow or narrow range level is restored, the operator is directed out of the guideline.

Therefore, consistent with the ERGs, the alternate indications of SG Narrow Range Level or Auxiliary Feedwater Flow may be used in lieu of Wide Range SG level. In addition, the ERG that uses Wide Range SG level would not be applicable if the alternate indications confirm that SG dryout will not occur.

• Power Range Neutron Flux Neutron flux decreasing is one of the indications used to determine if the reactor is tripped in the first step of E-0, "Reactor Trip Or Safety Injection," ECA-0.0, "LossOf All AC Power," and FR-S.1, "Response to Nuclear Power Generation/ATWS." The other indications used are rod bottom lights lit, reactor trip and bypass breakers open, and rod position indicators at zero. The ERGs also confirm the viability of using the intermediate and source indications to confirm a reactor trip. The ERGs assume that following a reactor trip it is expected that neutron flux will decrease rapidly to low in the power range then continue to decrease at a relatively steady rate until stabilizing in the source range. The intermediate range detectors are used to monitor the decreasing neutron flux until the source range detectors are energized.

Page 10

TSTF-525, Rev. 0 High Head Safety Injection Flow The ERGs contain the following information related to this indication. If high-head SI flow is not verified after SI is actuated, the operator is directed to manually start high-head SI pumps (if necessary) and align high-head SI valves to the proper position for injection into the RCS (if necessary). Separate actions are provided in E-0, "Reactor Trip Or Safety Injection," to verify the high-head pumps are running and the SI valves are in proper emergency alignment. Separate actions are also provided in FR-C.1, "Response to Inadequate Core Cooling," FR-C.2, "Response to Degraded Core Cooling," and FR-C.3, "Response to Saturated Core Cooling," to verify the SI valves are in proper emergency alignment.

Auxiliary Feedwater (AFW) Flow AFW flow is used, along with Steam Generator (SG) narrow range level, to determine if an adequate secondary heat sink is provided. Note that "total" feed flow is generally prescribed; therefore, other sources of feed flow may be used if available. The related ERG assumptions provide the following additional information: If AFW flow (or "total" feed flow) is not greater than the specified minimum value, the operator is directed to check for adequate SG narrow range level to maintain a secondary heat sink. If narrow range level is not adequate, the operator is directed to manually start AFW pumps (if necessary) and align AFW valves to the proper position for injection into the SGs (if necessary). Separate actions are provided in E-0, "Reactor Trip Or Safety Injection," to verify the AFW pumps are running and the AFW valves are in proper emergency alignment. If AFW flow is not adequate in ECA-0.0, "Loss Of All AC Power," when AC power is not available to the motor-driven AFW pumps, the operator is directed to manually open the steam supply valves to the turbine-driven AFW pump and align AFW valves to the proper position for injection into the SGs (if necessary).

[Preparer's Note: The following Table is included to summarize the proposed alternate monitoring methods and the TSTF-525 justifications for the use of instrumentation that is not RG 1.97 qualified. This Table should be revised on a plant-specific basis to describe the proposed alternate PAM Instrumentation and the basis for its acceptability.]

Alternate Indication Method Summary Table PAM Proposed Alternate RG 1.97 [Associated [Highly [Used in the Instrumentation Monitoring Method Qualified with Safety Reliablel EOPsl Related Systeml

[SG Water Level [SG Water Level [Yes or No] [Yes or No] [Yes or No] [Yes or No]

Wide Range] Narrow Range]

[Power Range [Intermediate Range [Yes or No] [Yes or No] [Yes or No] [Yes or No]

Neutron Flux] Indication and Rod Bottom lights]

Page II

TSTF-525, Rev. 0 PAM Proposed Alternate RG 1.97 JAssociated [Highly [Used in the Instrumentation Monitoring Method Qualified with Safety Reliablel EOPsI Related System]

[High Head [High Head Pump [Yes or No] [Yes or No] [Yes or No] [Yes or No]

Safety Injection Motor Amperage and Flow] Discharge Pressure]

[Auxiliary [Auxiliary Feedwater [Yes or No] [Yes or No] [Yes or No] [Yes or No]

Feedwater Flow Pump Motor Motor Driven] Amperage and Discharge Pressure]

[Auxiliary [Auxiliary Feedwater [Yes or No] [Yes or No] [Yes or No] [Yes or No]

Feedwater Flow Pump Discharge Steam Driven] Pressure and Flow Control Valve Position]

Page 12