L-04-094, License Amendment Request No. 184

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License Amendment Request No. 184
ML042100493
Person / Time
Site: Beaver Valley
Issue date: 07/23/2004
From: Pearce L
FirstEnergy Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
L-04-094
Download: ML042100493 (35)
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Text

Beaver Valley Power Station Route 168 P.Q . Box 4 FirstEnergy Nuclear Operating Company Shippingport, PA 15077-0004 L. William Pearce 724-G82-5234 Site Vice President Fax: 724-G43-8069 July 23, 2004 L-04-094 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001

Subject:

Beaver Valley Power Station, Unit No . 2 Docket No. 50-412, License No. NPF-73 License Amendment Request No. 184 Pursuant to 10 CFR 50 .90, FirstEnergy Nuclear Operating Company (FENOC) hereby requests an amendment to the above license in the form of changes to the Technical Specifications (TS) . The proposed amendment would eliminate periodic response time testing requirements on selected sensors and selected protection channel components.

The proposed amendment modifies TS Section 1 .0 Definitions for "REACTOR TRIP SYSTEM RESPONSE TIME" and "ENGINEERED SAFETY FEATURE RESPONSE TIME" to provide for verification of response time for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC . Surveillances 4.3 .1 .1 .3 and 4.3 .2 .1 .3 are modified consistent with the new definitions.

These changes are in conformance with changes approved in WCAP-13632-P-A, Revision 2, and WCAP-14036-P-A, Revision l . These are proprietary documents developed by Westinghouse and approved by the NRC in August 1995, and October 1998, respectively .

The reason for the requested changes is to permit the option of either measuring or verifying the response time for specific components in the Reactor Trip System and Engineered Safety Feature Actuation System instrumentation. WCAP-13632-P-A, Revision 2, is for specific pressure sensors and WCAP-14036-P-A, Revision 1, is for instrument loop channels. This option will allow the use of allocated response times in lieu of measured response times for selected sensors and protection channel components .

The FENOC evaluation of the proposed changes are presented in the Enclosure . The proposed Beaver Valley Power Station (BVPS) Unit No. 2 Technical Specification changes are presented in Attachment A. The proposed Technical Specification Bases changes are presented in Attachment B. The proposed Technical Specification Bases changes are provided for information only. New commitments contained within this submittal are described in Attachment C.

Beaver Valley Power Station, Unit No. 2 License Amendment Request No. 184 L-04-094 Page 2 The Beaver Valley Power Station review committees have reviewed the changes. The changes were determined to be safe and do not involve a significant hazard consideration as defined in 10 CFR 50 .92 based on the attached safety analysis and no significant hazard evaluation .

FENOC requests approval of the proposed amendment by February 28, 2005, so that sufficient time will be provided to include consideration of the proposed amendment in the work planning process far the next BVPS Unit No . 2 refueling outage, which is scheduled for the spring of 2005, Once approved, the amendment shall be implemented within 60 days .

If there are any questions concerning this matter, please contact Mr. Larry R. Freeland, Manager, Regulatory Affairs/Performance Improvement at 724-682-5284 .

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 23 , 2004.

L'. William Pearce

Enclosure:

FENOC Evaluation of the Proposed Changes Attachments :

A Proposed BVPS Unit 2 Technical Specification Changes B Proposed BVPS Unit 2 Technical Specification Bases Changes (for information only)

C Commitment Summary c: Mr. T. G. Colburn, NRR Senior Project Manager Mr. P. C. Cataldo, NRC Sr. Resident Inspector Mr. H. J. Miller, NRC Region I Administrator Mr . D. A. Allard, Director BRP/DEP Mr. L. E. Ryan (BRP/DEP)

ENCLOSURE FENOC Evaluation of the Proposed Changes Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184

Subject:

Proposed Modification to Technical Specifications Requirements Associated with Response Time Testing of Selected Pressure Sensors and Selected Protection Channel Components Table of Contents Section Title Page

1.0 DESCRIPTION

.......................................................................... 1

2.0 PROPOSED CHANGE

S ........................................................... 1

3.0 BACKGROUND

........................................................................ 3

4.0 TECHNICAL ANALYSIS

........................................................ 4 5.0 REGULATORY SAFETY ANALYSIS.................................. 17 5.1 No Significant Hazards Consideration..................................... 17 5.2 Applicable Regulatory Requirements/Criteria......................... 19

6.0 ENVIRONMENTAL CONSIDERATION

............................. 20

7.0 REFERENCES

......................................................................... 20 Attachments Number Title A Proposed BVPS Unit 2 Technical Specification Changes B Proposed BVPS Unit 2 Technical Specification Bases Changes (for information only)

C Commitment Summary i

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184

1.0 DESCRIPTION

This is a request to amend Operating License NPF-73 (Beaver Valley Power Station Unit 2).

The proposed changes will revise the Beaver Valley Power Station Unit No. 2 Technical Specifications to eliminate periodic response time testing requirements on selected sensors and selected protection channel components and permit the option of either measuring or verifying, by means other than testing, the response times.

2.0 PROPOSED CHANGE

S The proposed Technical Specification changes, which are submitted for NRC review and approval, are provided in Attachment A. The changes proposed to the Technical Specification Bases are provided in Attachment B.

The proposed Technical Specification Bases changes do not require NRC approval. The Beaver Valley Power Station (BVPS) Technical Specification Bases Control Program controls the review, approval and implementation of Technical Specification Bases changes. The Technical Specification Bases changes are provided for information only. Attachment C provides a list of commitments associated with this License Amendment Request (LAR).

The proposed changes to the Technical Specifications and Technical Specification Bases have been prepared electronically. Deletions are shown with a strike-through and insertions are shown double-underlined. This presentation allows the reviewer to readily identify the information that has been deleted and added.

To meet format requirements the Technical Specifications and Bases pages will be revised and repaginated as necessary to reflect the changes being proposed by this LAR.

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Changes to the following Technical Specifications (TS) are being proposed to eliminate periodic response time testing requirements on selected sensors and selected protection channel components.

Affected Technical Specifications Specification Title Definition 1.22 REACTOR TRIP SYSTEM RESPONSE TIME Definition 1.23 ENGINEERED SAFETY FEATURE RESPONSE TIME 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The following provides a description of the proposed changes and a basis for the changes.

Proposed Change The proposed change to the technical specifications will allow use of allocated response times in lieu of measured response times for selected sensors and selected protection system components. The proposed amendment modifies TS Section 1.0 Definitions for REACTOR TRIP SYSTEM RESPONSE TIME and ENGINEERED SAFETY FEATURE RESPONSE TIME to be consistent with TS definition changes approved in WCAP-14036-P-A, Revision 1. The definitions are revised to provide for verification of response time for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC. Surveillances 4.3.1.1.3 and 4.3.2.1.3 are modified consistent with the new definitions. The associated Bases for specifications 3/4.3.1 and 3/4.3.2 will be revised to clarify that allocations for sensor response times may be obtained from: 1) historical records based on acceptable response time tests; 2) in place, onsite, or offsite (e.g., vendor) test measurements; or 3) utilizing vendor engineering specifications. In addition, the Bases revision will clarify that allocations for signal processing Page 2

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 and actuation logic times may also be used in the verification of the overall protection system channel response times.

Basis for Proposed Change The current BVPS Technical Specifications require periodic measurement of response times of reactor protection and engineered safety features instrumentation channels. The proposed change would eliminate the requirement to actually measure the response times on selected sensors and selected protection channel components. Instead, the response times could be verified by summing allocated times for sensors, the process protection system, the nuclear instrumentation system, and the logic system. These allocated values will be added to the measured times for the actuated devices and compared to the overall analysis limits. The evaluations and conclusions contained in WCAP-13632-P-A Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, WCAP-14036-P-A Revision 1, Elimination of Periodic Protection Channel Response Time Tests, and WCAP-15413-A Revision 0, Westinghouse 7300A ASIC-Based Replacement Module Licensing Summary Report, form the basis for the proposed change. These WCAPs are discussed further in Sections 3 and 4 of this enclosure.

3.0 BACKGROUND

IEEE Standard 338-1977, Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems, defines a basis for eliminating periodic response time testing. Section 6.3.4 of the Standard states:

Response time testing of all safety-related equipment, per se, is not required if, in lieu of response time testing, the response time of safety system equipment is verified by functional testing calibration checks, or other tests, or both.

The NRC accepted this Standard with Regulatory Guide 1.118, Periodic Testing of Electric Power and Protection Systems, Revision 2. WCAP-13632 and WCAP-14036 provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component into operational service and re-Page 3

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 verified following maintenance that may adversely affect the response time.

WCAP-15413 provides supplemental data for allocation of signal processing times.

4.0 TECHNICAL ANALYSIS

Basis for Proposed Change for Pressure Sensors WCAP-13632-P-A contains the technical basis and methodology for eliminating response time testing (RTT) requirements on sensors identified in the WCAP. The technical basis and methodology were approved by letter dated September 5, 1995 from Bruce A. Boger (NRC) to Roger A. Newton (WOG). The NRC safety evaluation for WCAP-13632-P-A requires confirmation by the licensee that the generic analysis in the WCAP is applicable to their plant, and that licensees take the following actions:

1. Perform a hydraulic RTT prior to installation of a new transmitter/switch or following refurbishment of the transmitter/switch (e.g., sensor cell or variable damping components) to determine an initial sensor-specific response time value.
2. For transmitters and switches that use capillary tubes, perform a RTT after initial installation and after any maintenance or modification activity that could damage the capillary tubes.
3. If variable damping is used, implement a method to assure that the potentiometer is at the required setting and cannot be inadvertently changed, or perform a hydraulic RTT of the sensor following each calibration.
4. Perform periodic drift monitoring of all Model 1151, 1152, 1153, and 1154 Rosemount pressure and differential pressure transmitters, for which RTT elimination is proposed, in accordance with the guidance contained in Rosemount Technical Bulletin No. 4 and continue to remain in full compliance with any prior commitments to Bulletin 90-01, Supplement 1, "Loss of Fill-Oil in Transmitters Manufactured by Rosemount." As an alternative to performing periodic drift monitoring of Rosemount transmitters, licensees may complete the following actions: (a) ensure that operators and technicians are aware of the Rosemount transmitter loss of fill-oil issue and make provisions Page 4

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 to ensure that technicians monitor for sensor response time degradation during the performance of calibrations and functional tests of these transmitters, and (b) review and revise surveillance testing procedures, if necessary, to ensure that calibrations are being performed using equipment designed to provide a step function or fast ramp in the process variable and that calibrations and functional tests are being performed in a manner that allows simultaneous monitoring of both the input and output response of the transmitter under test, thus allowing, with reasonable assurance, the recognition of significant response time degradation.

FENOC has reviewed the plant data for BVPS Unit 2. The sensors installed in Unit 2 that are bounded by the generic analysis contained in WCAP-13632-P-A are identified in Table 1 of this enclosure.

The response time to be allocated in place of response times obtained through actual measurement during the period of verification may be obtained from:

(1) Historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests)

(2) In place, onsite, or offsite (e.g., vendor ) test measurements (3) Utilizing vendor engineering specifications BVPS Unit 2 will use allocated sensor response times based on historical records (as described in Table 1, Note 1).

FENOC responses to the conditions of the NRC SER contained in WCAP-13632-P-A are as follows:

Response to Item 1 Consistent with the proposed TS changes (including the associated Bases for TS 3/4.3.1. and TS 3/4.3.2.), the applicable plant procedures will include revisions which stipulate that pressure sensor response times must be verified by performance of an appropriate response time test prior to placing a new sensor into operational service and re-verified following maintenance that may adversely affect sensor response time.

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Response to Item 2 BVPS Unit 2 has no pressure sensors (transmitters or switches) that use capillary tubes in any Reactor Trip System (RTS) or Engineered Safety Features Actuation System (ESFAS) application for which RTT is required. Therefore, no procedure changes or enhanced administrative controls are required. If BVPS Unit 2 replaces any of these sensors in the future with sensors using capillary tubes, then BVPS Unit 2 will implement plant procedure changes (and/or other appropriate administrative controls) prior to application of the WCAP methodology to assure the sensors are response time tested after initial installation and after any maintenance or modification activity that could damage the capillary tubes.

Response to Item 3 BVPS Unit 2 has no pressure transmitters with variable damping installed in any RTS or ESFAS application for which RTT is required; therefore, no procedure changes or enhanced administrative controls are required. If BVPS Unit 2 replaces any transmitters in the future with variable damping capability, then BVPS Unit 2 will implement procedure changes and/or establish appropriate administrative controls prior to application of the WCAP methodology to assure the variable damping potentiometer cannot be inadvertently changed.

Response to Item 4 BVPS Unit 2 has no Rosemount transmitters that are installed in any RTS or ESFAS application that requires response time testing, and therefore no periodic drift monitoring of Rosemount transmitters, for which response time testing elimination is proposed, is required.

Basis for Proposed Change for Protection Channels WCAP-14036-P-A, Revision 1 contains the technical basis and methodology for RTT requirements on protection channels identified in the WCAP. The basic justification for the elimination of periodic response time testing is based on a Failure Modes and Effects Analysis (FMEA) that: 1) determined Page 6

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 that individual component degradation had no response time impact; or

2) identified components that may contribute to trip system response time degradation. Where potential response time impact was identified, testing was conducted to determine the magnitude of the response time degradation, or a bounding response time limit for the system or component was determined. As a result of the FMEA, the only components that were tested were the Westinghouse 7100 and 7300 Process Protection System circuit boards and modules. For the remainder of the hardware types shown in segments 2 and 3 of Figure 1 of the WCAP (e.g., Nuclear Instrumentation System, Eagle 21, Solid State Protection System and relay logic), bounding response time allocations were determined. In these cases the bounding response time allocation is derived from design response time specifications for the component.

For the 7100 and 7300 process protection system circuit boards and modules, the FMEA was performed by having a circuit designer review the circuits and identify those components that may increase the response time if they degrade from their nominal value. The time response of dynamic function (i.e., lead-lag, etc.) cards is verified during periodic calibration testing and, therefore, these cards were not included in the program. Where it was necessary to provide a response time limit with component degradation, the conclusions of the FMEA were quantified by testing card and module response times with degraded components.

The FMEA does the following:

- identifies response time sensitive components on the cards and modules via circuit analysis;

- evaluates the impact on the response time if a component fails or degrades;

- identifies detectability of degraded component via calibration; and

- identifies components that impact calibration but not response time.

The analysis identified capacitors and resistors as the dominant response time sensitive components. Other tested components included diodes, zener diodes, inductors, and potentiometers. Increased capacitance tends to lead to Page 7

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 increased response time. Manufacturers of sensitive capacitors on the printed circuit cards identified the failure mechanism and the maximum change in capacitance which could be reached before the capacitor failed.

One manufacturer stated that the capacitance will not increase beyond 25%

of the nominal value. All of the responses of the manufacturers provided gross estimates that capacitors identified in the 7300 circuits do not have a failure mechanism that will double the nominal capacitance. Based on this information, a conservative increase of 50% in capacitance was used to determine the maximum change in response time for capacitor degradation.

Resistors were assumed to degrade to as much as 200% of the nominal resistance, which is a conservative increase based on engineering judgement.

Actual testing was used to verify and further quantify the FMEA results.

The test procedures were used to verify and/or determine actual response time of the card or module with a degraded capacitor or resistor.

Components of different values were substituted to simulate various degrees of degradation. The procedures required calibration checks on the card and module after each component change to determine if the calibration could or could not detect the degraded component. If the post-component change calibration inaccuracy exceeded 0.5% of span, then the degradation was considered detectable.

An input step change was used to obtain step response traces. The response time was defined as the time to reach 63% of the final output. This time is equal to the time constant of a dynamic system with a characteristic first order lag. For the 7300 cards, a slightly more conservative limit of 67% was used. In summary, the tests:

- measured the response time of calibrated production modules and provided response time base-line data;

- verified the analysis by measuring response times and obtaining calibration data for the card or module when the component(s) identified by analysis as having an impact on response time were degraded;

- verified that similar results would be obtained if testing was done at a temperature that more closely modeled the rack environment; and Page 8

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184

- measured the response time of a simulated protection channel from input to output with components degraded.

Sections 4.2 - 4.5 of WCAP-14036-P-A, Revision 1 present the results of the FMEA and testing with degraded components. Testing verified that the FMEA was conservative and provided a baseline response time value for each card and module tested. Testing components with simulated degradations was deemed necessary to precisely quantify the increase in response time, because the Westinghouse 7100 and 7300 process protection system FMEAs show that components can degrade and impact response time without a corresponding calibration or functional test failure. Because the degradation would be undetectable by routine calibration testing, bounding response times with a degraded component were determined. In cases where more than one component impacted the response time, the individual response time degradation increments were summed to estimate the total response time degradation for the card. The bounding response time is justified because of its small magnitude when compared to the total response time limit for the protection channel and because the simulated degradations were conservatively exaggerated as described above.

Sections 4.6 - 4.9 of WCAP-14036-P-A, Revision 1 present the results of the FMEA for the Nuclear Instrumentation System, EAGLE 21, Solid State Protection System and relay logic protection system. These systems did not require testing with degraded components. In some cases, the FMEA did not identify any response time sensitive components that are subject to degradation, and in other cases the effects of component degradation are accounted for in the overall response time allocation for the system.

In Section 8 of WCAP-14036-P-A, Revision 1, the methodology to integrate the component response time results into the determination of the limit for protection channels is presented. This information is then combined with the results of the actuated component periodic response time tests to ensure that the response time limits are verified.

Westinghouse 7300A ASIC-Based Replacement Modules (ABRMs)

WCAP-15413, Westinghouse 7300A ASIC-Based Replacement Module Licensing Summary Report Section 9 provides the details supporting Page 9

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 response time test deletion for ASICs type cards. The same methodology used in WCAP-14036 was used to analyze the ABRMs. The FMEA circuit analysis determined which components on the Main Board and Personality Modules were critical to response time. In lieu of testing, due to the less complex ABRM, the analysis took into account catastrophic component failure and degraded component performance to determine a bounding response time for the ABRM modules. This response time bounds the limit to which response time can be increased by degraded or failed components without that degradation or failure affecting calibration. The FMEA shows that component degradation will not increase the response time beyond the bounding response time without that degradation being detectable by other periodic surveillance tests, such as channel check, functional tests and/or calibrations.

Verification of Plant-Specific Configuration FENOC has reviewed the plant equipment installed in BVPS Unit 2.

Table 1 provides the listing of equipment installed in Unit 2 that is applicable to the generic analysis contained in WCAP-14036-P-A, Revision 1. The FEMA documented in WCAP-14036-P-A, Revision 1 is applicable to the equipment installed in BVPS Unit 2 and the analysis is valid for the versions of the boards used in the protection system with the following clarification. Some versions of 7300 cards installed (or planned to be installed) in BVPS Unit 2 are cards that have been re-designed since the original FEMA was performed for WCAP-14036-P-A (see Table 1, Note 3).

These newer cards have been evaluated by Westinghouse for RTT elimination. The results of this evaluation for the newer NLP, NSA and NAL cards concluded that the differences in the re-designed cards are insignificant with respect to the conclusions of the original FEMA and that the FEMA and bounding response times documented in WCAP-14036-P-A, Revision 1 is applicable to these newer cards. A new FMEA was performed for the re-designed NRA cards. This FEMA was based on the original NRA FEMA methodology used for WCAP-14036-P-A, Revision 1. The Westinghouse evaluation concluded that the newer NRA cards meet the bounding response times listed in WCAP-14036-P-A, Revision 1.

Also note that there are no ABRMs currently installed in BVPS Unit 2. If ABRMs are installed in the future as a replacement for 7300 cards, BVPS Unit 2 will verify the affected protection function system response time Page 10

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 requirement by adjusting the allocated response times provided in Table 1 to account for the ABRM replacement card, using the bounding response times listed in Table 9-1 of WCAP-15413.

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Table 1 Beaver Valley Power Station Unit 2 Process Channel & Actuation Logic Response Time Allocations Reactor Trip System FUNCTION SENSOR TIME 7300/NIS STRING TIME SSPS TIME (Note 1) (Note 3) (Note 5) RELAYS (Note 7)

(Note 6)

Power Range, Neutron Detectors Exempt N/A NIS FMEA 65 ms Input 20 ms Flux Power Range, Neutron Detectors Exempt N/A NIS FMEA 200 ms Input 20 ms Flux, Hi-Negative Rate OTDT (Vary Tavg) Weed N9004E-2B/ (Note 2) 400 ms Input 20 ms Weed N9004S-2B NRA+NSA+NSA+NSA

+NAL (Note 4)

OTDT (Vary Delta T) Weed N9004E-2B/ (Note 2) NRA+NSA+NSA+NAL 400 ms Input 20 ms Weed N9004S-2B (Note 4)

OTDT (Vary Press) ITT Barton 763/763A 400 ms NLP+NSA+NSA+NAL 400 ms Input 20 ms OTDT (Vary Flux) Detectors Exempt N/A NIS (1 ms)+NSA +NCH 401 ms Input 20 ms

+NSA+NAL OPDT (Vary Tavg) Weed N9004E-2B/ (Note 2) 400 ms Input 20 ms Weed N9004S-2B NRA+NSA+NSA+NSA

+NSA+NAL (Note 4)

OPDT (Vary Delta T) Weed N9004E-2B/ (Note 2) NRA+NSA+NSA+NAL 400 ms Input 20 ms Weed N9004S-2B (Note 4)

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Table 1 Beaver Valley Power Station Unit 2 Process Channel & Actuation Logic Response Time Allocations Reactor Trip System FUNCTION SENSOR TIME 7300/NIS STRING TIME SSPS TIME (Note 1) (Note 3) (Note 5) RELAYS (Note 7)

(Note 6)

PZR. PRESS. LO ITT Barton 763/763A 400 ms NLP+NAL 100 ms Input 20 ms PZR. PRESS. HI ITT Barton 763/763A 400 ms NLP+NAL 100 ms Input 20 ms RCS FLOW LO ITT Barton 752 200 ms NLP+NAL 100 ms Input 20 ms SG LEVEL LO-LO ITT Barton 764 400 ms NLP+NAL 100 ms Input 20 ms RCP UNDER-VOLT. Gould 211N6171 (47D) (Note 2) N/A N/A Input 20 ms RCP UNDER-FREQ. E-MAX SFR-2/59-12A (Note 2) N/A N/A Input 20 ms Page 13

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Table 1 Beaver Valley Power Station Unit 2 Process Channel & Actuation Logic Response Time Allocations Engineered Safety Features Actuation System FUNCTION SENSOR TIME 7300/NIS TIME SSPS RELAYS TIME (Note 1) STRING (Note 5) (Note 6) (Note 7)

(Note 3)

CONTAINMENT PRESS. HI ITT Barton 764 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms PZR. PRESS. LO ITT Barton 763/763A 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms STEAM LINE PRESS. LO ITT Barton 763/763A 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms CONT. PRESS HI-HI ITT Barton 764 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms SG LEVEL HI-HI (FEEDWATER ITT Barton 764 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms ISOLATION)

CONT. PRESS. INT HI-HI ITT Barton 764 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms STEAM LINE PRESS RATE -HI ITT Barton 763/763A 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms NEG SG LEVEL LO-LO ITT Barton 764 400 ms NLP+NAL 100 ms Input + Master + Slave 88 ms UNDERVOLTAGE RCP Gould 211N6171 (Note 2) N/A N/A Input + Master (Note 8) 52 ms (47D)

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Table 1 Notes

1. Allocated sensor response times for the ITT Barton 752, 763 (763A) and 764 pressure sensors specified in Table 1 are based on historical records (method 1) of acceptable RTT obtained from the BVPS response time testing program. The sensor response times for all sensors except the containment pressure sensors were obtained using the Analysis and Measurement Services (AMS) noise analysis method. The containment pressure sensor response times were obtained using a hydraulic RTT method. The historical response time data used as the bases for the allocated response times were the AMS sensor response time measurement tests performed on October 1998, August 2000, January 2002 and August 2003 and the containment pressure sensors hydraulic RTT performed between February 1995 and August 2003. The highest response time measured for the ITT Barton sensors were as follows: Barton 752 -

160 ms; Barton 763 (763A) - 370 ms; Barton 764 - 380 ms.

2. Allocated response times not used for these variables. The components will continue to be tested as required.
3. 7300 cards installed are 4NCH, 4NRA (5NRA and 6NRA may be installed later in 2004), 11NLP, 6NSA and 10NAL or older artwork levels. The WCAP-14036-P-A R1 evaluation for RTT elimination was based on 4NCH, 4NRA, 6NLP, 4NSA and 9NAL or older artwork levels versions of the 7300 cards. The newer versions of the 7300 cards used at BVPS Unit 2 have been evaluated by Westinghouse and the bounding response times reported in WCAP-14036-P-A R1 are valid for these versions of cards. The NIS components installed were evaluated in Section 4.6 of WCAP-14036-P-A R1.
4. Card string includes a Lead/Lag card set to zero. Therefore, this card will continue to be periodically RTT and response time contribution included in the total channel response time in accordance with WCAP-14036-P-A R1, Section 8.0.
5. The allocated response times are derived from Table 8-1 of WCAP-14036-P-A R1. If ABRMs are installed in the future, response times listed will be adjusted to account for the ABRM installed using the response times listed in Table 9-1 of WCAP-15413.
6. SSPS Input and Master relays are Midtex Series 156 and Potter & Brumfield KH series relays. SSPS Slave relays are Westinghouse AR relays and Potter & Brumfield MDR relays. Values are tabulated from Section 4.8 of WCAP-14036-P-A R1.
7. The allocated response times for the SSPS reactor trip functions (input relay) and ESFAS functions (input relay, master relay and slave relay) are derived from Table 8-1 of WCAP-14036-P-A R1. For ESFAS functions, the time shown only accounts for one slave relay in the circuit. For circuits containing two slave relays in series an additional 36 msec. must be added.

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Table 1 Notes (cont.)

8. Slave relay actuation is time delayed, and therefore, will continue to be tested to verify timer operation.

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Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 5.0 REGULATORY SAFETY ANALYSIS The proposed license amendment modifies the Beaver Valley Power Station (BVPS) Unit 2 Technical Specifications, Section 1.0 definitions for Reactor Trip System Response Time and Engineered Safety Feature Response Time and Surveillance Requirements 4.3.1.1.3 and 4.3.2.1.3. These changes would revise the definition and surveillance requirements for response time testing of the Reactor Trip System (RTS) and Engineered Safety Features Actuation System (ESFAS). These changes are in conformance with changes approved in WCAP-13632-P-A, Revision 2 and WCAP-14036-P-A, Revision 1. The reason for the proposed changes are to eliminate the requirement for periodic response time testing for specific sensors and protection channel components and allow the response time to be verified based on the results of WCAP-13632-P-A, Revision 2 and WCAP-14036-P-A, Revision 1. This option will give BVPS the opportunity to eliminate redundant measurement of channel performance without reducing the reliability of these systems.

5.1 No Significant Hazards Consideration FirstEnergy Nuclear Operating Company (FENOC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10CFR50.92, 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.

This change to the Technical Specifications does not result in a condition where the design, material, and construction standards that were applicable prior to the change are altered. The same RTS and ESFAS instrumentation is being used; the time response allocations/modeling assumptions in the Updated Final Safety Analysis Report (UFSAR) Chapter 15 analyses are still the same; only the method of verifying time response is changed. The proposed change will not modify any system interface and could not increase the likelihood of an accident since these events are independent of this change. The proposed activity will not change, degrade or prevent Page 17

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 actions or alter any assumptions previously made in evaluating the radiological consequences of an accident described in the UFSAR.

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.

This change does not alter the performance of the pressure and differential pressure transmitters, process protection racks, Nuclear Instrumentation, and logic systems used in the plant Reactor Trip and Engineered Safety Features Actuation Systems. All sensors, process protection racks, Nuclear Instrumentation, and logic systems will still have response time verified by test before placing the equipment into operational service and after any maintenance that could affect the response time. Changing the method of periodically verifying instrument response times for certain equipment (assuring equipment operability) from time response testing to calibration and channel checks will not create any new accident initiators or scenarios.

Periodic surveillance of these instruments will detect significant degradation in the equipment response time characteristics.

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

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

Response: No.

This change does not affect the total system response time assumed in the safety analysis. The periodic system response time verification method for selected pressure and differential pressure sensors and for process protection racks, Nuclear Instrumentation, and logic systems is modified to allow use of actual test data or engineering data. The method of verification still provides assurance that the total system response time is within that assumed in the safety analysis.

Page 18

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, FENOC concludes that elimination of periodic equipment response time testing for specific sensors and protection channel components is acceptable and the proposed license amendment presents no significant hazards consideration under the standards set forth in 10CFR50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

5.2 Applicable Regulatory Requirements/Criteria A review of the applicable requirements/criteria was conducted to assess the potential impact associated with the proposed changes. A summary of the applicable regulatory requirements and criteria are provided in the following Tables. In the following paragraphs these requirements/criteria as they relate to the proposed changes are discussed.

General Design Criteria Assessment 13 Instrumentation and Controls No Impact 20 Protection System Functions No Impact 21 Protection System Reliability and Testability No Impact Regulatory Guides Assessment 1.118 Periodic Testing of Electric Power and No Impact Protection Systems The proposed change would eliminate the requirement to periodically measure by test selected sensor and protection channel response times.

The proposed change will not alter the protection systems design function, reliability or testability. As discussed in Section 3.0 of this enclosure, IEEE Standard 338-1977 defines a basis for eliminating periodic response time testing. The NRC accepted this standard with Regulatory Guide 1.118. Therefore, the proposed change is consistent with Regulatory Guide 1.118.

In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not Page 19

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions 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 CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10CFR20, or would change an inspection or surveillance requirement. However, the proposed amendment 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 amendment meets the eligibility criterion for categorical exclusion set forth in 10CFR51.22(c)(9). Therefore, pursuant to 10CFR51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. WCAP-13632-P-A, Revision 2, Elimination of Pressure Sensor Response Time Testing Requirements, dated January 1996.
2. WCAP-14036-P-A, Revision 1, Elimination of Periodic Protection Channel Response Time Tests, dated October 6, 1998.
3. WCAP-15413-A, Revision 0, Westinghouse 7300A ASIC-Based Replacement Module Licensing Summary Report, dated March 2001.
4. IEEE Standard 338-1977, Criteria for the Periodic Surveillance Testing of Nuclear Power Generating Station Safety Systems.
5. NRC Regulatory Guide 1.118, Periodic Testing of Electric Power and Protection Systems, Revision 2, dated June, 1978.
6. 10CFR50, Appendix A, General Design Criteria for Nuclear Power Plants.
7. Precedent NRC Approved License Amendment

References:

7.1 Virgil C. Summer Nuclear Station, Unit No. 1 - Issuance of Amendment Re: Technical Specifications Changes Related to Page 20

Beaver Valley Power Station Unit No. 2 License Amendment Request No. 184 Response Time Testing Elimination (TAC No. MA8632) dated August 29, 2000.

7.2 Millstone Nuclear Power Station, Unit No. 3 - Issuance of Amendment Re: Pressure Sensor Response Time Verification (TAC No. MA9360) dated November 3, 2000.

7.3 Joseph M. Farley Nuclear Plant, Units 1 and 2 Re: Issuance of Amendments (TAC Nos. MB0100 and MB0101) dated June 7, 2001.

7.4 South Texas Project, Units 1 and 2 - Issuance of Amendments on Elimination of Response Time Testing (TAC Nos. MB1412 and MB1420) dated August 21, 2001.

7.5 Virgil C. Summer Nuclear Station, Unit No. 1 - Inclusion of Two Upgraded Instrument Cards Into Response Time Testing Elimination Category (TAC No. MB2236) dated March 12, 2002.

7.6 Catawba Nuclear Station, Units 1 and 2 - Issuance of Amendments Re: (TAC Nos. MB2108 and MB2109) dated April 22, 2002.

7.7 McGuire Nuclear Station, Units 1 and 2 Re: Issuance of Amendments Regarding Response Time Testing Requirements (TAC Nos. MB4676 and MB4677) dated August 23, 2002.

7.8 Shearon Harris Nuclear Power Plant, Unit 1 - Issuance of Amendment Re: Elimination of Periodic Pressure Sensor and Protection Channel Response Time Tests (TAC No. MB6230) dated March 7, 2003.

7.9 Comanche Peak Steam Electric Station, Units 1 and 2 -

Issuance of Amendments Re: Elimination of Periodic Protection Channel Response Time Testing (TAC Nos.

MB7984 and MB7985) dated September 25, 2003.

Page 21

Attachment A Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Changes License Amendment Request No. 184 The following is a list of the affected pages:

1-4 1-5 3/4 3-1 3/4 3-15

DEFINITIONS 1.15 THROUGH 1.17 (DELETED)

QUADRANT POWER TILT RATIO (QPTR) 1.18 QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

DOSE EQUIVALENT I-131 1.19 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The DOSE EQUIVALENT I-131 is calculated with the following equation:

C I132 C I133 C I134 C I135 C I131 C I131    

D.E. 170 6 1000 34 Where C is the concentration, in microcuries/gram of the iodine isotopes. This equation is based on dose conversion factors derived from ICRP-30.

STAGGERED TEST BASIS 1.20 A STAGGERED TEST BASIS shall consist of:

a. A test schedule for n systems, subsystems, trains or other designated components obtained by dividing the specified test interval into n equal subintervals;
b. The testing of one (1) system, subsystem, train or other designated component at the beginning of each subinterval.

FREQUENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

REACTOR TRIP SYSTEM RESPONSE TIME 1.22 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until loss of stationary gripper coil voltage.

The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

BEAVER VALLEY - UNIT 2 1-4 Amendment No. 101

DEFINITIONS ENGINEERED SAFETY FEATURE RESPONSE TIME 1.23 The ENGINEERED SAFETY FEATURE RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two-section excore neutron detector.

PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and 1) described in Chapter 14.0 of the FSAR,

2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

E - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average sum (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

SOURCE CHECK 1.27 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.

PROCESS CONTROL PROGRAM 1.28 The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

1.29 DELETED BEAVER VALLEY - UNIT 2 1-5 Amendment No. 101

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.

APPLICABILITY: As shown in Table 3.3-1.

ACTION: As shown in Table 3.3-1.(2)

SURVEILLANCE REQUIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel and interlock and automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements(1) during the MODES and at the frequencies shown in Table 4.3-1.

4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated verified to be within its limit at least once per 18 months. Neutron detectors are exempt from response time testing. Each test verification shall include at least one logic train such that both logic trains are tested verified at least once per 36 months and one channel per function such that all channels are tested verified at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

(1) For the automatic trip logic, the surveillance requirements shall be the application of various simulated input combinations in conjunction with each possible interlock logic state and verification of the required logic output including, as a minimum, a continuity check of output devices.

(2) Separate ACTION statement entry is allowed for each Function.

BEAVER VALLEY - UNIT 2 3/4 3-1 Amendment No. 120

INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1.1 Each engineered safety feature actuation system instrumentation channel and interlock and the automatic actuation logic with master and slave relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements(1) during the MODES and at the frequencies shown in Table 4.3-2.

4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the at power CHANNEL FUNCTIONAL TEST of channels affected by interlock operation. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.

4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESF function shall be demonstrated verified to be within the limit at least once per 18 months. Each test verification shall include at least one logic train such that both logic trains are tested verified at least once per 36 months and one channel per function such that all channels are tested verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESF function as shown in the Total No. Of Channels Column of Table 3.3-3.

(1) For the automatic actuation logic, the surveillance requirements shall be the application of various simulated input conditions in conjunction with each possible interlock logic state and verification of the required logic output including, as a minimum, a continuity check of output devices. For the actuation relays, the surveillance requirements shall be the energization of each master and slave relay and verification of OPERABILITY of each relay. The test of master relays shall include a continuity check of each associated slave relay. The test of slave relays (to be performed at least once per 92 days in lieu of at least once per 31 days) shall include, as a minimum, a continuity check of associated actuation devices that are not testable. The slave relay test frequency can be extended to once per 12 months provided a satisfactory contact loading analysis has been completed, and a satisfactory slave relay service life has been established, for the slave relay being tested.

BEAVER VALLEY - UNIT 2 3/4 3-15 Amendment No. 141

Attachment B Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Bases Changes License Amendment Request No. 184 The following is a list of the affected pages:

B 3/4 3-2 B 3/4 3-3

3/4.3 INSTRUMENTATION Provided for Information Only.

BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

OPERABILITY of the following trips in Table 3.3-1 provides additional diverse or anticipatory protection features and is not credited in the accident analyses:

Undervoltage - Reactor Coolant Pumps (Above P-7); Underfrequency Reactor Coolant Pumps (Above P-7); Turbine Trip (Above P-9); Reactor Coolant Pump Breaker Position Trip (Above P-7); Turbine First Stage Pressure, P-13.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, "Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and supplements to that report as approved by the NRC and documented in the SER (letter to J. J. Sheppard from Cecil O. Thomas dated February 21, 1985).

Jumpers and lifted leads are not an acceptable method for placing equipment in bypass as documented in the NRC safety evaluation report for this WCAP.

The surveillance requirements for the Manual Trip Function, Reactor Trip Breakers, and Reactor Trip Bypass Breakers are provided to reduce the possibility of an Anticipated Transient Without Scram (ATWS) event by ensuring OPERABILITY of the diverse trip features

(

Reference:

Generic Letter 85-09).

The measurementverification of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

ESF response times which include sequential operation of the RWST and VCT valves are based on values assumed in the non-LOCA safety analyses and are provided in Section 3 of the Licensing Requirements Manual. These analyses take credit for injection of borated water from the RWST. Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves. When sequential operation of the RWST and VCT valves is not included in the response times, the values specified are based on the LOCA analyses. The LOCA analyses take credit for injection flow regardless of the source. Verification of the response times will assure that the assumptions used for the LOCA and Non-LOCA analyses with respect to operation of the VCT and RWST valves are valid.

BEAVER VALLEY - UNIT 2 B 3/4 3-2 Change No. 2-016

3/4.3 INSTRUMENTATION Provided for Information Only.

BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

The maximum response time for control room isolation on high radiation is based on ensuring that the control room remains habitable following a small line break outside the containment.

From a control room habitability aspect, the worst case accident that does not initiate a Containment Isolation - Phase B signal is the small line break outside the containment. This response time includes radiation monitor processing delays associated with the monitor averaging techniques. Diesel Generator starting and sequence loading delays are not included since these delays occur prior to the control room environment exceeding the high radiation setpoint.

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor, signal processing and actuation logic response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) in place, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

WCAP-14036-P-A Revision 1, "Elimination of Periodic Protection Channel Response Time Tests" and WCAP-15413, Westinghouse 7300A ASIC-Based Replacement Module Licensing Summary Report provide the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time. The allocations for sensor, signal conditioning and actuation logic response times must be verified prior to placing the component in operational service and re-verified following maintenance that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

WCAP-15413 provides bounding response times where 7300 cards have been replaced with ASICs cards.

The Engineered Safety Feature Actuation System interlocks perform the following functions:

P-4 Reactor tripped - Actuates turbine trip, closes main feedwater valves on Tavg below setpoint, prevents the opening of the main feedwater valves which were closed by a safety injection or high steam generator water level signal, allows safety injection block so that components can be reset or tripped.

Reactor not tripped - prevents manual block of safety injection.

P-11 Above the setpoint, P-11 automatically reinstates safety injection actuation on low pressurizer pressure, automatically blocks steamline isolation on high steam pressure rate, and enables safety injection and steamline isolation (with Loop Stop Valve Open) on low steamline pressure. Below the setpoint, P-11 allows the manual block of safety injection actuation on low pressurizer pressure, allows manual block of safety injection and steamline isolation (with Loop Stop Valve Open) on low steamline pressure and enables steamline isolation on high steam pressure rate.

P-12 Above the setpoint, P-12 automatically reinstates an arming signal to the steam dump system. Below the setpoint P-12 blocks steam dump and allows manual bypass of the steam dump block to cooldown condenser dump valves.

BEAVER VALLEY - UNIT 2 B 3/4 3-3 Amendment No. 120 Change No.

Attachment C Beaver Valley Power Station, Unit No. 2 Commitment Summary License Amendment Request No. 184

Commitment List The following table identifies those actions committed to by FirstEnergy Nuclear Operating Company (FENOC) for Beaver Valley Power Station (BVPS) Unit No. 2 in this document. Any other actions discussed in the submittal represent intended or planned actions by FENOC. They are described only as information and are not regulatory commitments. Please notify Mr. Larry R. Freeland, Manager, Regulatory Affairs/Performance Improvement, at Beaver Valley on (724) 682-5284 of any questions regarding this document or associated regulatory commitments.

COMMITMENT DUE DATE

1. Applicable plant procedures will be revised to Prior to implementation stipulate that RTS/ESFAS pressure sensor response of the proposed times must be verified by performance of an amendment appropriate response time test prior to placing a new RTS/ESFAS pressure sensor into operational service and re-verified following maintenance that may adversely affect sensor response time.
2. If BVPS Unit 2 replaces any RTS or ESFAS pressure Prior to application of sensors for which response time verification is WCAP-13632 required in the future with sensors using capillary methodology for the tubes, then BVPS Unit 2 will implement plant associated sensors procedure changes (and/or other appropriate administrative controls) to assure the sensors are response time tested after initial installation and after any maintenance or modification activity that could damage the capillary tubes.
3. If BVPS Unit 2 replaces any RTS or ESFAS pressure Prior to application of transmitters for which response time verification is WCAP-13632 required in the future with pressure transmitters which methodology for the have variable damping capability, then BVPS Unit 2 associated transmitters will implement procedure changes and/or establish appropriate administrative controls to assure the variable damping potentiometer cannot be inadvertently changed.

C-1

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