DCL-03-016, License Amendment Request 03-02, Response Time Testing Elimination & Revision to TS 3.3.1, Reactor Trip System Instrumentation

License Amendment Request 03-02, Response Time Testing Elimination & Revision to TS 3.3.1, Reactor Trip System Instrumentation

ML030660616
Person / Time
Site:	Diablo Canyon
Issue date:	02/28/2003
From:	Oatley D Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-03-016
Download: ML030660616 (50)

Text

PacificGas and ElectricCompany David H. Oatley Diablo Canyon Power Plant Vice President and PO Box 56 General Manager Avila Beach, CA 93424 February 28, 2003 8055454350 Fax 805 545 4234 PG&E Letter DCL-03-016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units I and 2 License Amendment Request 03-02 Response Time Testing Elimination and Revision to Technical Specification 3.3.1, "Reactor Trip System (RTS) Instrumentation"

Dear Commissioners and Staff:

In accordance with 10 CFR 50.90, enclosed is an application for amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant respectively. This License Amendment Request (LAR) revises Technical Specification (TS) 3.3.1, "Reactor Trip System (RTS) Instrumentation," to add Surveillance Requirement (SR) 3.3.1.16 to function 3.a, Power Range Neutron Flux Rate - High Positive Rate Trip in Table 3.3.1-1. Westinghouse recently identified that the Power Range Neutron Flux Rate - High Positive Rate Trip function is credited to provide protection against Reactor Coolant System over pressurization during a rod withdrawal at power event.

This LAR additionally proposes to eliminate periodic pressure sensor response time testing (RTT) in accordance with WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensing Response Time Testing Requirements," and to eliminate periodic protection channel RTT in accordance with WCAP-1 4036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests."

In 1999, Pacific Gas and Electric Company (PG&E) implemented Diablo Canyon License Amendment (LA) 135 /135 incorporating changes to the TS that allow response times to be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor 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.

A member of the STARS (Strategic Teaming and Resource Sharing) Altiance Callaway

• Comanche Peak

• Diablo Canyon

• Palo Verde

• South Texas Project

• Wolf Creek

Document Control Desk PG&E Letter DCL 03-016 February 28, 2003 Page 2 The definitions contained in TS section 1.1, "Definitions" for Engineered Safety Features Response Time and RTS Response Time, require NRC review and approval of any methodology used to verify the response times in lieu of measuring them. The application of the methodology to verify the response times for selected components does not require changes to the TS.

The Bases for SR 3.3.1.16 and SR 3.3.2.10 are being revised, consistent with NRC approved traveler TSTF-1 11, Revision 6, to 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, using the methodology in WCAP-14036-P-A, Revision 1. The NRC approved WCAP-14036-P-A, Revision 1, by letter dated October 6, 1998, Thomas H. Essig (NRC) to Lou Liberatori, Westinghouse Owners Group (WOG).

The Bases for SR 3.3.1.16 and SR 3.3.2.10 were previously revised, consistent with TSTF-1 11, Revision 4, through LA 135/135 to provide the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific pressure and differential pressure sensors identified in WCAP-1 3632-P-A, Revision 2. By letter dated September 5, 1995, Bruce A. Boger (NRC) to Roger A. Newton (WOG), the NRC approved WCAP-1 3632-P-A, Revision 2. Specific approval for use of WCAP-1 3632-P-A, Revision 2 was not previously obtained.

PG&E has verified that the specific components for which response time testing elimination is proposed are the same manufacturer and model number as those components evaluated in WCAP-1 3632-P-A, Revision 2, and WCAP-1 4036-P-A, Revision 1.

PG&E has reviewed the proposed amendment in accordance with 10 CFR 50.92 and determined that it does not involve a significant hazards consideration. In addition, Diablo Canyon has determined that the proposed amendment satisfies the criteria of 10 CFR 51.22(c)(9) for categorical exclusion from the requirement for an environmental assessment.

Enclosure 1 provides a description of the proposed change, supporting technical analyses, no significant hazards consideration determination, and environmental evaluation. The marked-up TS pages, retyped TS pages, marked-up TS Bases pages (for information only), Updated Final Safety Analysis Report changes (for information only), and the allocated response time tables are contained in enclosures 2 through 6, respectively.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway 9 Comanche Peak

• Diablo Canyon

• Palo Verde e South Texas Project a Wolf Creek

Document Control Desk PG&E Letter DCL 03-016 February 28, 2003 Page 3 The proposed changes in this LAR are not required to address an immediate safety concern. However, in order to facilitate scheduling and avoid preparatory costs associated with the twelfth refueling outage for Unit 1 currently scheduled for March 2004, PG&E requests that this amendment be approved no later than November 30, 2003. PG&E requests the LAR be made effective upon NRC issuance, to be implemented within 60 days from the date of issuance.

Sincerely, David H. Oatley Vice Presidentand GeneralManager- Diablo Canyon mjr/4557 Enclosures cc: Edgar Bailey, DHS Ellis W. Merschoff David L. Proulx Diablo Distribution cc/enc: Girija S. Shukla A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak a Diablo Canyon e Palo Verde

• South Texas Project

• Wolf Creek

PG&E Letter DCL-03-016 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

) Docket No. 50-275 In the Matter of ) Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY) No. DPR-80

))

Diablo Canyon Power Plant Docket No. 50-323 Units 1 and 2 ) Facility Operating License No. DPR-82 AFFIDAVIT David H. Oatley, of lawful age, first being duly sworn upon oath states that he is Vice President and General Manager - Diablo Canyon of Pacific Gas and Electric Company; that he is familiar with the content thereof; that he has executed LAR 03-02 on behalf of said company with full power and authority to do so; and that the facts stated therein are true and correct to the best of his knowledge, information, and belief.

David H. Oatley Vice Presidentand GeneralManager- Diablo Canyon Subscribed and sworn to before me this 28th day of February 2003.

SANDRA L. RECTOR Commission# 1339380 z Notary Public z Notary Public - Calrfornia _

County of San Luis Obispo z

• San Luis Obispo County State of California My Comme E0esJan 122 00~6

Enclosure 1 PG&E Letter DCL-03-016 EVALUATION

1.0 DESCRIPTION

This letter is a request to amend Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP),

respectively.

The proposed change revises Technical Specification (TS) 3.3.1, "Reactor Trip System (RTS) Instrumentation," to add Surveillance Requirement (SR) 3.3.1.16 to function 3.a, Power Range Neutron Flux Rate - High Positive Rate Trip (hereafter referred to as positive flux rate trip (PFRT)) in Table 3.3.1-1. Westinghouse recently identified that the PFRT function is credited to provide protection against Reactor Coolant System (RCS) over pressurization during a rod withdrawal at power (RWAP) event.

This letter additionally requests NRC review and approval of use of the methodology of WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," dated October 1998, (Reference 2) and WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," dated January 1996, (Reference 1) to eliminate periodic protection channel response time testing (RTT) and periodic pressure sensor R'-, respectively at DCPP.

The definitions contained in TS Section 1.1, "Definitions," for Engineered Safety Features (ESF) Response Time and RTS Response Time, require prior NRC review and approval of any methodology used to verify the response times for selected components in lieu of measuring them.

The application of the methodology to verify response times for selected components does not require changes to the TS. The TS Bases will be revised consistent with NRC-approved TSTF-1 11, Revision 6.

2.0 PROPOSED CHANGE

2.1 TS Table 3.3.1-1 This amendment application would add SR 3.3.1.16 to function 3.a of TS table 3.3.1-1. SR 3.3.1.16 requires that RTS response times be verified to be within limits every 24 months on a staggered test. Function 3.a is the PFRT function.

Enclosures 2 and 3 provide the TS markup and the retyped TS for the Table 3.3.1-1 revision. Enclosures 4 and 5 provide an information-only 1

Enclosure 1 PG&E Letter DCL-03-016 copy of the associated TS Bases changes and Updated Final Safety Analysis Report (UFSAR) changes, respectively.

2.2 Response Time Testing Elimination The current DCPP TS 3.3.1 and TS 3.3.2 require measurement of response times of reactor protection and engineered safety features instrumentation channels. The proposed change would allow verification in lieu of measurement for sensors, the process protection system, the nuclear instrumentation system, and the logic system. The verification involves use of allocated values for the response times. These allocated values will be added to the measured times for the actuated devices and compared to the overall analysis limits. The TS requirements for response time verification will continue to be implemented by SRs 3.3.1.16 and 3.3.2.10. The implementation of the methodology in WCAP-14036-P-A, Revision 1, requires changes to the TS Bases B 3.3.1 and B 3.3.2. The TS Bases changes are consistent with TSTF-1 11, Revision 6.

TS Bases B 3.3.1 requires a change to section SR 3.3.1.16. TS Bases B 3.3.2 requires a change to section SR 3.3.2.10.

The following paragraph will be added to the above sections:

"[WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," 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 in operational service and reverified following maintenance work 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."

In each case, this paragraph will replace the existing paragraph that states:

"The allocations for sensor response times must be verified prior to placing the component in initial operational service and re-verified following maintenance that may adversely affect response time. In 2

Enclosure 1 PG&E Letter DCL-03-016 general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value.

One example where response time could be affected is replacing the sensing assembly of a transmitter."

Additionally, the following reference will be added to the reference sections of TS Bases B 3.3.1 and B 3.3.2:

"WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.

These changes are noted on the marked-up TS Bases, Enclosure 4.

3.0 BACKGROUND

3.1 TS Table 3.3.1-1 SR 3.3.1.16 requires a verification that RTS response times are within their limits every 24 months on a staggered test basis, as defined in TS 3.3.1. As discussed in the SR 3.3.1.16 Bases, the acceptance criteria time limits for the response time tests are included in an Equipment Control Guideline (ECG). ECG's provide administrative control over plant equipment. ECG 38.1, table 38.1-1 lists the time limit acceptance criteria for the response time tests for RTS functions. These limits are less than or equal to the maximum values assumed in the accident analyses.

DCPP license amendment (LA) 135/135 relocated the response time acceptance criteria for individual functions requiring response time verification from TSs to ECGs. The associated response time requirement for PFRT has always been listed as "N.A." in both the initial licensing basis TS Table 3.3-2 and currently in ECG 38.1, table 38.1-1. As a result, the current DCPP Response Time Testing Program does not verify the response time for the PFRT function.

When Westinghouse performed the original analysis for the licensing of DCPP, the PFRT function was not credited in any UFSAR Chapter 15 analysis. Therefore, it was appropriate to use "N.A." for the response time of the PFRT function.

However, Pacific Gas and Electric Company (PG&E) recently determined that Westinghouse had performed a generic evaluation, which credited the PFRT to provide protection against certain rod withdrawal at power accidents that are analyzed with assumptions intended to maximize the 3

Enclosure 1 PG&E Letter DCL-03-016 primary system pressure response (Reference 10). With these assumptions, it was determined that the RWAP event was not limiting, relative to RCS overpressure. The limiting event for RCS overpressure was determined to be the "Loss of Load / Turbine Trip," whose analysis is reflected in chapter 15 of the UFSAR.

After PG&E determined the Westinghouse generic evaluation which credits the PFRT function is applicable to DCPP, response time testing of all PFRT channels for Units 1 and 2 was performed; testing was completed on April 6, 2002.

Westinghouse did not identify the PFRT function as a DCPP analysis assumption requirement based on a conclusion that the generic evaluation which credited the PFRT function was conservatively bounding and did not require response time testing. Westinghouse assumed a three-second response time for the PFRT in the generic evaluation. This is significantly greater than the maximum delay time of 0.5 seconds typically assumed for the Nuclear Instrumentation System (NIS) trip functions in the safety analyses. Westinghouse concluded that normal functional testing of the PFRT instrumentation should assure that the three-second response time assumption would be satisfied since no conceivable failure could extend the response time past three seconds without rendering the channel inoperable. This conclusion is consistent with the basic premise for allocating a response time to a trip function (Reference 2). However, since PG&E has determined that the generic PFRT response time assumption is part of the basis for preventing the RWAP event from resulting in the limiting the RCS overpressure condition, this application proposes that SR 3.3.1.16 be applied to the PFRT function.

As discussed in Section 4.1 below, the response time limit that will be specified in plant procedures for the PFRT function will be three seconds or less.

3.2 Response Time Testing Elimination In 1975, RTT requirements were included in the Westinghouse Standard Technical Specifications and were required for all plants after that date.

IEEE Standard 338-11977, "Criteria for the Periodic Testing of Class 1E Power and Protection Systems," defines a basis for eliminating periodic response time testing (Reference 8). Section 6.3.4 of the Standard states in part:

4

Enclosure 1 PG&E Letter DCL-03-016 "Response time testing of all safety-related equipment, per se, is not required if, in lieu of response time testing, the response time of the system equipment is verified by functional testing calibration checks, or other tests, or both. This is acceptable if it can be demonstrated that changes in response time beyond acceptable limits are accompanied by changes in performance characteristics which are detectable during routine periodic tests."

The NRC stated in Regulatory Guide 1.118, "Periodic Testing of Electric Power and Protection Systems, Revision 2," that the requirements and recommendations contained in IEEE Standard 338-1977 are considered acceptable methods for the periodic testing of electric power and protection systems.

The DCPP TS contain definitions for both RTS and ESF response times.

The response time definitions are:

"The RTS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RTS trip setpoint at the channel sensor until loss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and methodology for verification have been previously reviewed and approved by the NRC."

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

WCAP-1 3632-P-A, Revision 2 provides the basis for elimination of response time testing requirements on pressure sensors identified in the WCAP. By letter dated September 5, 1995, from Bruce A. Boger (NRC) to Roger A. Newton, Westinghouse Owners Group (WOG), the NRC 5

Enclosure 1 PG&E Letter DCL-03-016 approved the technical basis and methodology of WCAP-1 3632-P-A, Revision 2.

DCPP LA 135/135 incorporated the following wording into the Bases section of SR 3.3.1.16 and SR 3.3.2.10 supporting use of WCAP-13632-P-A, Revision 2:

"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 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) inplace, 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" (Reference 8) 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."

Although the TS and TS Bases changes required to support use of WCAP-1 3632-P-A, Revision 2, have been incorporated into the DCPP TS, PG&E has not previously requested specific NRC approval for application of WCAP-13632-P-A, Revision 2 methodology to DCPP. Therefore, PG&E requests NRC approval to apply the methodology of WCAP-13632-P-A, Revision 2 to DCPP.

WCAP-14036-P-A, Revision 1 provides the basis for elimination of protection system channel response time testing. By letter dated October 6, 1998, Thomas H. Essig (NRC) to Lou Liberatori, (WOG), the NRC approved the technical basis and methodology of WCAP-14036-P-A, Revision 1.

Due to the complexity of testing an entire instrument channel from the sensor to the final device, plant surveillance procedures currently conduct RTT in multiple tests, summing the response times to obtain the total channel response time for the sensors, NIS, Eagle 21 Protection System, and Solid State Protection System (SSPS) relays as applicable.

PG&E requests NRC approval to apply the methodology of the WCAP-14036-P-A, Revision 1 to DCPP.

6

Enclosure I PG&E Letter DCL-03-016 4.0 TECHNICAL ANALYSES 4.1 TS Table 3.3.1-1 The discussion in DCPP UFSAR Section 15.2.2 and the assumptions listed in FSAR Table 15.1-1, are based on analyzing the Uncontrolled Rod Cluster Control Assembly Bank Withdrawal at Power event for minimum departure from nucleate boiling ratio (DNBR).

The overtemperature delta-T (OTAT) trip function and the PFRT function are the primary reactor trips credited in UFSAR Section 15.2.2 to protect against minimum DNBR conditions. Since the DNBR becomes more limiting with a lower RCS pressure, the DCPP UFSAR analysis of the RWAP event assumes that the pressurizer pressure control system functions to minimize RCS pressure. The RWAP event with a malfunctioning pressurizer pressure control system was not originally evaluated and was not considered limiting for RCS overpressure concerns at the time the DCPP UFSAR Section 15.2 analyses were last updated.

However, the generic Westinghouse evaluation discussed above in Section 3.1 notes that a low power RWAP could result in the RCS pressure exceeding the 110 percent design limit (2750 psia) if only the typically credited UFSAR Chapter 15 trip functions (high pressurizer pressure, overtemperature delta-T, and power range neutron flux - high) are credited and assuming that the pressurizer pressure control system malfunctions. The heat generated by the nuclear fuel in response to the positive reactivity addition resulting from the postulated rod withdrawal would cause an increase in RCS pressure. The potential for RCS overpressure increases as the time between the reactivity insertion and a reactor trip increases due to the time lag associated with transfer of the increased heat generated in the core through the fuel cladding and into the reactor coolant.

The magnitude of the RCS pressure increase resulting from the RWAP is a function of the reactivity insertion rate, the initial power level, and the amount of reactivity feedback. For small positive reactivity insertion rates, nuclear power and RCS temperature increase relatively slowly and in equilibrium such that the thermal lag effect on RCS overpressure is not a concern. For large reactivity insertion rates at the end of core life conditions, there is a large reactivity feedback effect such that the nuclear power and RCS temperature still increase in relative equilibrium and RCS overpressure is not a concern. In the case of large reactivity insertion rates at the beginning of core life with the corresponding minimal reactivity 7

Enclosure 1 PG&E Letter DCL-03-016 feedback effects, nuclear power increases much faster than the rate at which the energy can be transferred into the RCS. For fast RWAP transients that occur at a relatively high initial power levels, the high flux trip is reached before the RCS heat-up rate has increased significantly, thus RCS overpressure is still not a concern.

However, for a RWAP event at low initial power levels, more time is available to transfer the heat generated as a result of the positive reactivity addition to the reactor coolant before any other reactor trip is actuated.

This leads to a potential for significant energy transfer to the RCS, which then results in a substantial increase in RCS pressure.

For the RWAP event, the factors that result in a potential overpressure condition in the RCS are only present at low reactor power levels.

Westinghouse identified that the limiting RWAP overpressure case occurred at an initial power level of 10 percent rated thermal power (RTP).

This is the lowest power level at which the power range neutron high flux low trip function could be blocked by permissive P-10. Westinghouse determined that by the time a reactor trip on high pressurizer pressure, OTAT, or power range neutron flux - high would occur, the RCS heatup and volumetric expansion rate could exceed the relief capacity of the pressurizer safety valves. However, Westinghouse concluded in their generic evaluation that crediting the PFRT function with a setpoint of 9 percent RTP and a time constant on the NIS rate circuit of 2 seconds, as well as a 3-second delay time on the rest of the PFRT function circuitry, would mitigate this event before the RCS pressure limit was exceeded.

Note that in this submittal the term time constant is referring to the NIS rate circuit in each of the four NIS power range channels which is surveilled under the SR 3.3.1.11 channel calibration.

The 3-second response time assumed for the PFRT in the generic evaluation was significantly greater than the maximum delay time of 0.5 seconds typically assumed for the NIS trip functions in the safety analyses. Westinghouse concluded that the assumptions related to the PFRT function were conservatively bounding and did not require response time testing and the PFRT function was not identified to licensees as a safety analysis assumption requirement.

Although not currently documented in the UFSAR, the Westinghouse generic RWAP evaluation for the RCS overpressure case represents part of the DCPP design basis for establishing that the Chapter 15 events do not result in exceeding the 110 percent RCS design pressure limit.

8

Enclosure 1 PG&E Letter DCL-03-016 To ensure the validity of the assumptions made in the Westinghouse generic evaluation, TS Table 3.3.1-1 should list SR 3.3.1.16 for the PFRT function. The PFRT safety analysis limit of 9 percent reactor power with a time constant greater than 2 seconds is bounded by both the DCPP Nominal Trip Setpoint of 5 percent RTP with a time constant of greater than or equal to 2 seconds, as well as the Allowable Value of less than or equal to 5.6 percent RTP with a time constant of greater than or equal to 2 seconds, as reflected in TS Table 3.3.1-1.

Probabilistic Risk Assessment (PRA) Evaluation There is no adverse impact on the DCPP PRA since this change adds an additional surveillance requirement to the PFRT function to provide assurance that it will satisfy its credited function. This change will not modify the physical plant. Any additional required testing will not increase the risk.

Summary and Conclusion Adding SR 3.3.1.16 to Function 3.a of TS Table 3.3.1-1 will assure that accident analysis assumptions regarding the response time of equipment credited in accident mitigation are verified on a periodic basis. The discussions presented above assess the potential impact of this change on the safety analysis that credits the PFRT trip function. These assessments demonstrate that the change will not adversely affect the design basis, safety analyses, or the safe operation of the plant.

4.2 Response Time Testing Elimination Basis for Proposed Change for Sensors WCAP-1 3632-P-A, Revision 2, utilizes the sensor failure modes effects analyses (FMEA) contained in Electric Power Research Institute (EPRI)

Report NP-7243, "Investigation of Response Time Testing Requirements,"

dated May 1991, and EPRI Report NP-7243, Revision 1, dated March 1994. The information presented in WCAP-13632-P-A, Revision 2 shows that, in general, failure modes associated with the pressure sensors analyzed by EPRI and the WOG would not affect sensor response time without an accompanying effect on sensor output.

Therefore, the failure modes that have the potential to affect sensor response times would be detected during the performance of other TS surveillance requirements, principally sensor calibration.

9

Enclosure 1 PG&E Letter DCL-03-016 PG&E has reviewed the plant data for Diablo Canyon Units 1 and 2. The sensors installed in both units are those that are bounded by the generic analysis contained in WCAP-13632-P-A, Revision 2.

They are:

"* Rosemount Model 1153

"* Rosemount Model 1154

"° Barton Model 763 , Allocated Response Time Tables, provides the list of equipment installed at DCPP. Method 3 in section 9 of WCAP-13632-P-A, Revision 2, specifies the use of vendor engineering specifications for response time allocations. Rosemount vendor manuals

1. 00809-0100-4388, #4514, and #00809-0100-4631 provide vendor engineering specifications for response times. Response times for Barton 763 transmitters were provided in table 9-1 of WCAP-13632-P-A, Revision 2. The Westinghouse E Spec time allocation, which is the most conservative time provided in table 9-1, was used for Barton 763 transmitters.

The NRC safety evaluation report (SER) for WCAP-13632-P-A, Revision 2, requires confirmation by the licensee that the generic analysis in the WCAP is applicable to their plant and that the 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 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 10

Enclosure 1 PG&E Letter DCL-03-016 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 to make sure that technicians monitor for sensor response time degradation during 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.

The PG&E responses to the conditions of the NRC SER contained in WCAP-13632-P-A, Revision 2 are as follows:

Response to Item 1 Applicable plant procedures include stipulations that pressure sensor response times must be verified by performance of an appropriate response time test (hydraulic, noise, or power interrupt tests) prior to placing a new transmitter/switch into operational service and reverified following maintenance that may adversely affect sensor response time.

Response to Item 2 Plant procedures stipulate that pressure sensors (transmitters and switches) utilizing capillary tubes, e.g., containment pressure, must be subjected to RTT after initial installation and following any maintenance or modification activity which could damage the transmitter capillary tubes.

Response to Item 3 DCPP has no pressure transmitters with variable damping installed in any RTS or Engineered Safety Features Actuation System application for which RTT is required; therefore, no DCPP procedure changes or enhanced administrative controls are required. Ifany transmitters are 11

Enclosure 1 PG&E Letter DCL-03-016 replaced in the future with variable damping capability, then either a hydraulic RTT of the sensor will be performed following each calibration, or procedure changes will be implemented and/or appropriate administrative controls will be established to assure the variable damping potentiometer can not be inadvertently changed.

Examples of such administrative controls may include use of pressure transmitters that are factory set and hermetically sealed to prohibit tampering or in situ application of a tamper seal (or sealant) on the potentiometer to secure it and to give a visual indication of the potentiometer position.

Response to Item 4 DCPP has Rosemount model 1153 Series B, 1153 Series D and 1154 transmitters for which RTT elimination is proposed. For Rosemount transmitters that do not meet the criteria for exclusion from the enhanced monitoring program as indicated in NRC Bulletin 90-01 Supplement 1, DCPP will continue to perform enhanced monitoring using the guidelines of Rosemount Technical Bulletin No. 4 and the requirements of NRC Bulletin 90-01 Supplement 1, until these transmitters are replaced.

Basis for Proposed Change for Protection Channels WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," 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 reverified following maintenance that may adversely affect the response time.

The NRC SER for WCAP-14036-P-A requires confirmation by the licensee that the analysis performed by the WOG is applicable to the equipment installed in the plant and that the analysis is valid for the versions of the boards used in the protection system.

The equipment installed in the plant, as listed below, is of the same manufacturer and model identified in WCAP-14036-P-A, Revision 1:

NIS

"* Detector Current Monitor Circuits

"* Summing and Level Amplifier 12

Enclosure 1 PG&E Letter DCL-03-016

"* Level Trip Bistables

"* Isolation Amplifiers EAGLE 21 (E21)

"* ERI - RTD Input Board

"* EAI -Analog Input Board

"* DFP - Digital Filter Processor

"* LCP - Loop Calculation Processor

"* DDC - Digital-Digital Converter

"* EPT - Partial Trip Output Board SSPS Input and Master Relays

"* G.P. Clare GP1 Series

"* Midtex 156

"* Potter and Brumfield KH Series SSPS Slave Relays

• Potter and Brumfield MDR The bounding response times in table 8-1 of WCAP-14036-P-A, Revision 1, are used for the above equipment. The allocated times are specified in Enclosure 6, "Allocated Response Time Tables."

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Pacific Gas and Electric (PG&E) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1) The proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Overall protection system performance will remain within the bounds of the previously performed accident analyses since there are no hardware changes.

The design of the Reactor Trip System (RTS) instrumentation, specifically the positive flux rate trip (PFRT) function, will be unaffected. The reactor protection system will continue to function in a manner consistent with the plant design basis. All design, 13

Enclosure 1 PG&E Letter DCL-03-016 material, and construction standards that were applicable prior to the request are maintained.

The proposed change imposes additional surveillance requirements to assure safety-related structures, systems, and components are verified to be consistent with the safety analysis and licensing basis. Inthis specific case, a response time verification requirement will be added to the PFRT function.

The Technical Specification Bases changes do not result in a condition where the design, material, or construction standards that were applicable prior to change are altered. The same RTS and engineered safety features actuation system 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 change any system interface and could not increase the likelihood of an accident since these events are independent of this change.

The proposed change will not affect the probability of any event initiators. There will be no degradation in the performance of, or an increase in the number of challenges imposed on safety-related equipment assumed to function during an accident situation. There will be no change to normal plant operating parameters or accident mitigation performance.

The proposed activity will not change, degrade or prevent 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) The proposed change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

There are no hardware changes nor are there any changes in the method by which any safety-related plant system performs its safety function. This change will not affect the normal method of plant operation or change any operating barameters. No performance requirements will be affected; however, the proposed change does impose additional surveillance requirements for the 14

Enclosure 1 PG&E Letter DCL-03-016 PFRT function. These additional requirements are consistent with assumptions made in the safety analysis and licensing basis.

This change does not alter the performance of the process protection racks, nuclear instrumentation, and logic systems used in the plant protection systems. These systems will still have their response time verified by test before being placed in operational service. Changing the method of verifying instrument response for these systems (assuring equipment operability) from time response testing to channel and calibration checks will not create any new accidents initiators or scenarios. Periodic surveillance of these systems will continue and may be used to detect degradation that could cause the response time characteristic to exceed the total allowance. The total response time allowance for each function bounds all degradation that cannot be detected by periodic surveillance.

No new accident scenarios, transient precursors, failure mechanisms, or limiting single failures are introduced as a result of this change. There will be no adverse effects or challenges imposed on any safety-related system as a result of this change.

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

3) The proposed change does not involve a significant reduction in margin of safety.

There will be no effect on the manner in which safety limits or limiting safety system settings are determined nor will there be any effect on those plant systems necessary to assure the accomplishment of protection functions. There will be no impact on the overpower limit, departure from nucleate boiling ratio limits, heat flux hot channel factor, nuclear enthalpy rise hot channel factor, loss of coolant accident peak cladding temperature, peak local power density, or any other margin of safety. The radiological dose consequence acceptance criteria listed in the Standard Review Plan will continue to be met.

The safety analysis limits assumed in the transient and accident analyses are unchanged. None of the acceptance criteria for any accident analysis are changed. The imposition of additional surveillance requirements maintains the margin of safety by 15

Enclosure 1 PG&E Letter DCL-03-016 assuring that the affected safety analysis assumptions on equipment response time are verified on a periodic frequency.

This change does not affect the total system response time assumed in the safety analysis. The periodic system response time verification method for the process protection racks, nuclear instrumentation, and logic systems are modified to allow use of engineering data. The method of verification still provides assurance that the total system response is within that defined in the safety analysis, since calibration tests will continue to be performed and may be used to detect any degradation which might cause the response time to exceed the total allowance. The total response time allowance for each function bounds all degradation that cannot be detected by periodic surveillance. Based on the above, it is concluded that the proposed change does not result in a significant reduction in margin with respect to plant safety.

Based on the above evaluation, it is concluded that the proposed change presents no significant hazards consideration under the standards set forth in 10 CFR 50.92, and accordingly, a finding of "no significant hazards consideration" is justified.

5.2 Applicable Regulatory Requirements I Criteria The regulatory bases and guidance documents associated with the systems discussed in this amendment application include:

General design criteria (GDC)-13 requires that instrumentation shall 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 assure adequate safety, including those variables and systems that can affect the fission process, the integrity of the reactor core, the reactor coolant pressure boundary, and the containment and its associated systems.

GDC-20 requires that the protection system(s) shall be designed (1) to initiate automatically the operation of appropriate systems including the reactivity control systems, to assure that specified acceptable fuel design limits are not exceeded as a result of anticipated operational occurrences and (2) to sense accident conditions and to initiate the operation of systems and components important to safety.

GDG-21 requires that the protection system(s) shall be designed for high functional reliability and testability.

16

Enclosure 1 PG&E Letter DCL-03-016 GDC-22 through GDC-25 and GDC-29 require various design attributes for the protection system(s), including independence, safe failure modes, separation from control systems, requirements for reactivity control malfunctions, and protection against anticipated operational occurrences.

Regulatory Guide 1.22 discusses an acceptable method of satisfying GDC-20 and GDC-21 regarding the periodic testing of protection system actuation functions. These periodic tests should duplicate, as closely as practicable, the performance that is required of the actuation devices in the event of an accident.

10 CFR 50.55a(h) requires that the DCPP protection systems, including RTS Function 3.a, meet IEEE 279-1971. Sections 4.9-4.11 of IEEE 279-1971 discuss testing provisions for protection systems.

Regulatory Guide 1.118, Revision 2, discusses acceptable methods for testing protection systems, including Section 6.3.4 of IEEE 338-1977 for response time testing.

To meet the guidance of Regulatory Guide 1.118, Revision 2, and IEEE 338-1977, Section 6.3.4, R'- is needed unless it has been shown that changes in the response time of a component requiring test will be accompanied by performance characteristics that are detectable during routine periodic tests.

The sensor analysis results contained in EPRI Report NP-7243, Revision 1, concluded that, in general, RI- is redundant to other periodic surveillance tests, such as channel checks and calibrations, because these other surveillance tests will detect sensor component failures that cause response time degradation. These tests are performed more frequently than current response time tests.

The FMEA in WCAP-14036-P-A, Revision 1, 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 checks and calibrations.

There will be no changes to the RTS instrumentation design such that there would be any adverse impact on the regulatory requirements and guidance documents above. This amendment application imposes additional surveillance requirements on RTS Function 3.a consistent with the above requirements.

17

Enclosure 1 PG&E Letter DCL-03-016 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 CONSIDERATION

PG&E has evaluated the proposed changes and determined the changes do not involve: (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criteria 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 amendments.

18

Enclosure 1 PG&E Letter DCL-03-016

7.0 REFERENCES

1. WCAP 13632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," dated January 1996, Proprietary Class 2C

2. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," dated October 1998, Proprietary Class 2C

3. NRC letter for License Amendment (LA) 30 to Facility Operating License NPF-76 and LA 119 to Facility Operating License NPF-80 for the South Texas Project, Units I and 2, respectively, "South Texas Project, Units 1 and 2 - Issuance of Amendments on Elimination of Response Time Testing (TAC NOS. MB1412 and MB 1420)," dated August 21, 2001

4. NRC Letter for PG&E License Amendments 135/135, "Conversion to Improved Technical Specifications for Diablo Canyon Power Plant, Units 1 and 2 -Amendment No. 135 to Facility Operating License Nos. DPR-80 and DPR-82," dated May 28, 1999

5. TSTF-111, Revision 6, "Revise Bases for SRs 3.3.1.15 and 3.3.2.10 to eliminate pressure sensor response time testing," dated April 28, 1999

6. Electric Power Research Institute Report NP-7243 "Investigation of Response Time Testing Requirements," dated May 1991

7. Electric Power Research Institute Report NP-7243, Revision 1 "Investigation of Response Time Testing Requirements," dated March 18, 1994

8. IEEE Standard 338-1977, "Criteria for the Periodic Testing of Class 1E Power and Protection Systems"

9. Regulatory Guide 1.118, Revision 2, "Periodic Testing of Electric Power and Protection Systems," dated June 1978

10. Nuclear Safety Advisory Letter, "Reactor Protection System Response Time Requirements," dated July 29, 2002 19

Enclosure 1 PG&E Letter DCL-03-016 7.1 PRECEDENCE A similar submittal was made by the South Texas Project (STP) Nuclear Operating Company for STP Units I and 2 in letter NOC-AE-01001020, "License Amendment Request - Proposed Modification to Technical Specifications Requirements Associated With Response Time Testing of Selected Pressure Sensors and Selected Protection Channels," dated February 28, 2001. The submittal requested changes to the plant TS and approval for application of the methodology of WCAP-1 3632-P-A, Revision 2 and WCAP-14036-P-A, Revision 1, to eliminate response time testing of select RTS and ESF components. The NRC approved the submittal by LA 30 to Facility Operating License NPF-76 and LA 119 to Facility Operating License NPF-80 for the STP, Units 1 and 2, respectively, in NRC letter "South Texas Project, Units 1 and 2 - Issuance of Amendments on Elimination of Response Time Testing (TAC NOS.

MB1412 and MB 1420)," dated August 21, 2001.

20

Enclosure 2 PG&E Letter DCL 03-016 MARKED UP TECHNICAL SPECIFICATION

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 7)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL(a)

SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual 1,2 2 B SR 3.3.1.14 NA NA Reactor Trip 3 (b), 4 (b), 5 (b) 2 C SR 3.3.1.14 NA NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 S110.2% 109% RTP SR 3.3.1.2 RTP SR 3.3.1.7 SR 3 3.1.11 SR 3 3.1.16

b. Low 1(c),2 4 E SR 3.3.1.1 S26.2% 25% RTP SR 3.3.1.8 RTP SR 3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux Rate

a. High 1,2 4 E SR 3.3.1.7 _<5.6% RTP 5% RTP Positive SR 3.3.1.11 with time with time Rate constant constant SR2 sec > 2 sec

b. High 1,2 4 E SR 3 3.1.7 -<5.6% RTP 5% RTP Negative SR3.3.1.11 with time with time Rate SR 3.3.1.16 constant constant

Ž 2 sec >2 sec 25%

4. Intermediate 1(c), 2 (d) 2 F,G SR 3 3.1.1

• 30.6%

Range SR 3.3.1.8 RTP RTP Neutron Flux SR3.3.1.11 (continued)

(a) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(c) Below the P-10 (Power Range Neutron Flux) interlocks.

(d) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

DIABLO CANYON - UNITS 1 & 2 3.3-1 Unit 1 - Amendment No. 435 442 TAB 3.3 - R2 12 Unit 2 - Amendment No. 4-35 442

Enclosure 3 PG&E Letter DCL 03-016 REVISED TECHNICAL SPECIFICATION Remove Page Insert Page 3.3-12 3.3-12

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 7) I Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINALma)

SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual 1,2 2 B SR 3.3.1.14 NA NA Reactor Trip 3 (b) 4 (b), 5 (b) 2 C SR 3.3.1.14 NA NA

2. Power Range Neutron Flux

a. High 1,2 4 D SR 3.3.1.1 <1102% 109% RTP SR 3.3.1.2 RTP SR 3 3.1.7 SR 3.3.1 11 SR 3.3.1.16

b. Low 4 E SR 3.3.1.1 _<26.2% 25% RTP SR 3.3.1.8 RTP SR3.3.1.11 SR 3.3.1.16

3. Power Range Neutron Flux Rate

a. High 1,2 4 E SR 3.3.1.7

• 5.6% RTP 5% RTP Positive SR 3.3.1.11 with time with time Rate SR 3.3.1.16 constant constant

Ž 2 sec > 2 sec

b. High 1,2 4 E SR 3.3.1.7 < 5.6% RTP 5% RTP Negative SR 3.3.1.11 with time with time Rate SR 3.3.1.16 constant constant

> 2 sec 2 sec 1 (c), 2 (d) 25%

4. Intermediate 2 F,G SR 3.3.1.1

• 30.6%

Range SR 3.3.1.8 RTP RTP Neutron Flux SR 3.3.1.11 (continued)

(a) A channel is OPERABLE with an actual Trip Setpoint value outside its calibration tolerance band provided the Trip Setpoint value is conservative with respect to its associated Allowable Value and the channel is re-adjusted to within the established calibration tolerance band of the Nominal Trip Setpoint. A Trip Setpoint may be set more conservative than the Nominal Trip Setpoint as necessary in response to plant conditions.

(b) With Rod Control System capable of rod withdrawal or one or more rods not fully inserted.

(c) Below the P-1 0 (Power Range Neutron Flux) interlocks (d) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

DIABLO CANYON - UNITS 1 & 2 3.3-1 Unit 1 - Amendment No. 435 442 TAB 3.3 - R2 12 Unit 2 - Amendment No. 435 442

Enclosure 4 PG&E Letter DCL 03-016 TECHNICAL SPECIFICATION BASES CHANGES (for information only)

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.16 (Continued)

REQUIREMENTS For channels that include dynamic transfer Functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer Function set to one, with the resulting measured response time compared to the appropriate FSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value, provided the required response time is analytically calculated assuming the time constants are set at their nominal values.

The response time may be measured by a series of overlapping tests such that the entire response time is measured.

The response time testing for the SG water level low-low does not include trip time delays. Response times include the transmitters, Eagle-21 process protection cabinets, solid state protection system cabinets, and actuation devices only. This reflects the response times necessary for THERMAL POWER in excess of 50 percent RTP. For those functions without a specified response time, SR 3.3.1.16 is not applicable.

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 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) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-1 3632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" (Ref. 8) 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.

The aIl0oations for censor response times m-ust be verified prier to Splacing the component ininitial operational rer.'ice and re-verifie-d INSET AiI following mnaintenance that may adyersely affect repos tie. in general, electrical repair work does not impc resone timne provide the padts used for repair are- of the sgame type and value. One example where response time could be affected is replacing thesein assembly of a transmitter:.

As appropriate, each channel's response time must be verified every 24 months on a STAGGERED TEST BASIS. Each verification shall (continued)

DIABLO CANYON - UNITS 1 &2 B 3.3-59 Revision I dc103016.doc- R1A 59

RTS Instrumentation B 3.3.1 BASES REFERENCES 8. WCAP 13632 - PA-1, Rev. 2 "Elimination of Pressure Sensor (Continued) Response Time Testing Requirements."

9. FSAR, Chapter 9.2.7 & 9.2.2.

10. FSAR, Chapter 10.3 & 10.4

11. FSAR, Chapter 8.3.

12. DCM S-38A, "Plant Protection System"

13. WCAP-1 3878, "Reliability of Potter & Brumfield MDR Relays", June 1994.

14. WCAP-13900, "Extension of Slave Relay Surveillance Test intervals", April 1994.

15. WCAP-14117, "Reliability Assessment of Potter and Brumfield MDR Series Relays."

16. WCAP-9226, "Reactor Core Response to Excessive Secondary Steam Releases," Revision 1, January 1978.

17. WCAP-1 1082, Rev. 5, "Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 and 2, 24 Month Fuel Cycle Evaluation," January 1997.

18. NSP-1-20-13F Unit I "Turbine Auto Stop Low Oil Pressure."

19. NSP-2-20-13F Unit 2 "Turbine Auto Stop Low Oil Pressure."

20. J-1 10 "24 Month Fuel Cycle Allowable Value Determination /

Documentation and ITDP Uncertainty Sensitivity."

21. IEEE Std. 338-1977.

22. License Amendment 61/60, May 23, 1991.

INS7ERT =

DIABLO CANYON - UNITS 1 & 2 B 3.3-59 Revision I dc103016.doc- RIA 59

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.2.10 (Continued)

REQUIREMENTS Response time Testing requirements," dated January 1996, provides the basis and the methodology of 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.

[.The for alocations.ensor response times m.ust be erified prior. to INSERT A] -,,;,

placin the component

, in initial operational service and re; erified following maintenance that may adversely affect response time. in general, eec~trical repair work does not impac~t response tie proevided the pants used for repair arc of the samne typo and value. One example where response time could be affected is replacing thesein assemnbly of a transmitter.

ESF RESPONSE TIME tests are conducted on a 24 month STAGGERED TEST BASIS.

Each verification shall include at least one train such that both trains are verified at least once per 48 months and one channel per function such that all channels are tested at least once every N times 24 months where N is the total number of redundant channels in a specific ESFAS function. Testing of the final actuation devices, which make up the bulk of the response time, is included in the testing of each train. Therefore, staggered testing results in response time verification of one train of devices every 24 months. The 24 month Frequency is consistent with the typical refueling cycle and is based on unit operating experience, which shows that random failures of instrumentation components causing serious response time degradation, but not channel failure, are infrequent occurrences.

This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching 650 psig in the SGs.

SR 3.3.2.11 SR 3.3.2.11 is the performance of a TADOT as described in SR 3.3.2.8, except that it is performed for the P-4 Reactor Trip Interlock. The 24 month Frequency is based on operating experience.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint.

(continued)

DIABLO CANYON - UNITS 1 & 2 B 3.3-104 Revision I dc103016.doc - RIA 104

RTS Instrumentation B 3.3.1 BASES REFERENCES 9. WCAP-13878, "Reliability of Potter & Brumfield MDR Relays", June (Continued) 1994.

10. WCAP-14117, "Reliability Assessment of Potter and Brumfield MDR Series Relays."

11. WCAP-1 3632-P-A, Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements," January 1996.

12. WCAP-1 1082, Revision 5, 'Westinghouse Setpoint Methodology for Protection Systems, Diablo Canyon Units 1 and 2, 24 Month Fuel Cycle Evaluation," January 1997.

13. Calculation J-54, "Nominal Setpoint Calculation for Selected PLS Setpoints."

14. J-110, "24 Month Fuel Cycle Allowable Value Determination/

Documentation and ITDP Uncertainty Sensitivity."

15. License Amendment 61/60, May 23,1991.

INSRTC ] >. I DIABLO CANYON - UNITS 1 & 2 B 3.3-106 Revision 1 dc103016.doc- R1A 106

Enclosure 4 PG&E Letter DCL 03-016 INSERT A

"[WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," 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 in operational service and reverified following maintenance work 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."

INSERT B

23. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.

INSERT C

16. WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," October 1998.

f

Enclosure 5 PG&E Letter DCL 03-016 UFSAR CHANGES (for information only)

Enclosure 5 PG&E Letter DCL 03-016 7.2.1.1.1.1 Nuclear Overpower Trips - Added clarification that item (4) PFRT also credited for low worth RCCA withdrawal at power events.

7.2.1.1.1.3 Reactor Coolant System Pressurizer Pressure and Water Level Trips Added clarification that item (3) PWLHT is credited for preventing pressurizer overfill for low worth RCCA withdrawal at power events.

Table'7.2-3 Trip Correlation - Added RCCA withdrawal at power event to the list of credited analyses for PFRT function.

15.2.2.1 Uncontrolled RCCA Withdrawal at Power - Added PFRT to list of credited mitigating RTS functions. Added detailed text discussion on the generic evaluations (new Ref.

14), which credit the PFRT and PWLHT functions to prevent RCS overpressure and pressurizer overfill.

15.2.16 References - Added new Ref. 14, Westinghouse Letter PGE-02-72 summarizing generic evaluation results for RWAP that credit PFRT and PWLHT functions.

Table 15.1 Assumed Trip Setpoint and Time Delays - Added PFRT setpoint and delay time, added PWLHT setpoint with note that assumed delay time is insensitive and not necessary.

DCPP UNITS 1 & 2 FSAR UPDATE 7.2 REACTOR TRIP SYSTEM 7.

2.1 DESCRIPTION

This section provides a system description and the design bases for the reactor trip system (RTS).

7.2.1.1 System Description The RTS uses sensors that feed the process circuitry consisting of two to four redundant channels, which monitor various plant parameters. The RTS also contains the logic circuitry necessary to automatically open the reactor trip breakers. The logic circuitry consists of two redundant logic trains that receive input from the protection channels.

Each of the two trains, A and B, is capable of opening a separate and independent reactor trip breaker (52/RTA and 52/RTB). The two trip breakers in series connect three-phase ac power from the rod drive motor generator sets to the rod drive power bus, as shown in Figure 7.2-1, Sheet 2. For reactor trip, a loss of dc voltage to the undervoltage coil releases the trip plunger and trips open the breaker. Additionally, an undervoltage trip auxiliary relay provides a trip signal to the shunt trip coil that trips open the breaker in the unlikely event of an undervoltage coil malfunction. When either of the trip breakers opens, power is interrupted to the rod drive power supply, and the control rods fall by gravity into the core. The rods cannot be withdrawn until an operator resets the trip breakers. The trip breakers cannot be reset until the bistable, which initiated the trip, reenergizes. Bypass breakers BYA and BYB are provided to permit testing of the trip breakers, as discussed below.

7.2.1.1.1 Reactor Trips The various reactoi trip circuits automatically open the reactor trip breakers whenever a condition monitored by the RTS reaches a preset level. In addition to redundant channels and trains, the design approach provides an RTS that monitors numerous system variables, thereby providing RTS functional diversity. The extent of this diversity has been evaluated for a wide variety of postulated accidents and is detailed in Reference 1.

Table 7.2-1 provides a list of reactor trips that are described below.

7.2.1.1.1.1 Nuclear Overpower Trips The specific trip functions generated are:

(1) Power Range High NuclearPower Trip - The power range high nuclear power trip circuit trips the reactor when two of the four power range channels exceed the trip setpoint.

7.2-1 Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE There are two independent bistables each with its own trip setting (a high and a low setting). The high trip setting provides protection during normal power operation and is always active. The low trip setting, which provides protection during startup, can be manually blocked when two of the four power range channels read above approximately 10 percent power (P-10). Three of the four channels sensing below 10 percent power automatically reinstate the trip function. Refer to Table 7.2-2 for a listing of all protection system interlocks.

(2) IntermediateRange High Neutron Flux Trip - The intermediate range high neutron flux trip circuit trips the reactor when one of the two intermediate range channels exceeds the trip setpoint. This trip, which provides protection during reactor startup, can be manually blocked if two of the four power range channels are above approximately 10 percent power (P-10). Three of the four power range channels below this value automatically reinstate the intermediate range high neutron flux trip. The intermediate range channels (including detectors) are separate from the power range channels. The intermediate range channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing during plant shutdown or prior to startup. This bypass action is annunciated on the control board.

(3) Source Range High Neutron Flux Tjip - The source range high neutron flux trip circuit trips the reactor when one of the two source range channels exceeds the trip setpoint. This trip, which provides protection during reactor startup and plant shutdown, can be manually blocked when one of the two intermediate range channels reads above the P-6 setpoint value and is automatically reinstated when both intermediate range channels decrease below the P-6 value. This trip is also automatically bypassed by two-out-of-four logic from the power range interlock (P-10). This trip function can also be reinstated below P-10 by an administrative action requiring manual actuation of two control board-mounted switches. Each switch will reinstate the trip function in one of the two protection logic trains. The source range trip point is set between the P-6 setpoint (source range cutoff flux level) and the maximum source range flux level. The channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing during plant shutdown or prior to startup. This bypass action is annunciated on the control board.

(4) PowerRange High PositiveNuclearPowerRate Trip - This circuit trips the reactor when an abnormal rate of increase in nuclear power occurs in two of the four power range channels. This trip provides protection against rod ejection and rod withdrawal accidents of low worth from middle to low power conditions and is always active.

(5) PowerRange High Negative NuclearPowerRate Trip - This circuit trips the reactor when an abnormal rate of decrease in nuclear power occurs in two of the four power range channels. This trip provides protection against two or more 7.2-2 Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE

(*) Refer to Technical Specifications for current values.

The source of temperature and flux information is identical to that of the overtemperature AT trip and the resultant AT setpoint is compared to the same AT. Figure 7.2-1, Sheet 5, shows the logic for this trip function.

7.2.1.1.1.3 Reactor Coolant System Pressurizer Pressure and Water Level Trips The specific trip functions generated are:

(1) PressurizerLow-Pressure Trip - The purpose of this trip is to protect against low pressure that could lead to departure from nucleate boiling (DNB), and to limit the necessary range of protection afforded by the overtemperature AT trip. The parameter being sensed is reactor coolant pressure as measured in the pressurizer.

Above P-7, the reactor is tripped when the dynamically compensated pressurizer pressure measurements fall below preset limits. This trip is blocked below P-7 to permit startup. The trip logic and interlocks are provided in Table 7.2-1.

The trip logic is shown in Figure 7.2-1, Sheet 6.

(2) PressurizerHig'h-PressureTrip - The purpose of this trip is to protect the reactor coolant system (RCS) against system overpressure.

The same sensors and transmitters used for the pressurizer low-pressure trip are used for the high-pressure trip except that separate comparators are used for the trip. These comparators trip when nondynamically compensated pressurizer pressure signals exceed preset limits on coincidence, as listed in Table 7.2-1.

There are no interlocks or permissives associated with this trip function.

The logic for this trip is shown in Figure 7.2-1, Sheet 6.

(3) PressurizerHigh Water Level Trip - This trip is provided as a backup to the pressurizer high-pressure trip and .... es-,toprevents water rclief thrug_* the pressurizer safety va1.ves from becoming water solid during low worth and low power rod withdrawal accidents. This trip is blocked below P-7 to permit startup.

The coincidence logic and interlocks of the pressurizer high water level signals are provided in Table 7.2-1.

The trip logic for this function is shown in Figure 7.2-1, Sheet 6.

7.2.1.1.1.4 Reactor Coolant System Low-Flow Trips These trips protect the core from DNB in the event of a loss of coolant flow situation. The means of sensing the loss of coolant are:

7.2-7 Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 7.2-3 Sheet 2 of 4 Trip Accident Tech Spec Section 15.2.13 (5) Accidental depressurization -

of the main system Section 15.4.6 (6) Rod ejection Table 3.3.1-1 Section 15.2.2

5. Power range high positive (1) Uncontrolled RCCA bank Table 3.3.1-1 nuclear power rate withdrawal at power Section 15.4.6 (1-)-_(2Rod Ejection I Section 15.2.3

6. Power range high (1) RCCA misoperation Table 3.3.1-1 negative nuclear power rate Section 15.2.2

7. Overpower AT (1) Uncontrolled RCCA bank Table 3.3.1-1 withdrawal at power Section 15.2.10 (2) Excessive heat due to feedwater system malfunction Section 15.2.11 (3) Excessive load increases Section 15.2.13 (4) Accidental depressurization of the main steam system Section 15.2.2

8. Overtemperature AT (1) Uncontrolled RCCA bank Table 3.3.1-1 withdrawal at power Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE 15.2.1.3 Results Figures 15.2.1-1 through 15.2.1-3 show the transient behavior for the indicated reactivity insertion rate with the accident terminated by reactor trip at 35 percent nominal power. This insertion rate is greater than that for the two highest worth control banks, both assumed to be in their highest incremental worth region.

Figure 15.2.1-1 shows the neutron flux transient. The neutron flux overshoots the full power nominal value but this occurs for only a very short time period. Hence, the energy release and the fuel temperature increase are relatively small. The thermal flux response, of interest for departure from nucleate boiling (DNB) considerations, is shown in Figure 15.2.1-2. The beneficial effect on the inherent thermal lag in the fuel is evidenced by a peak heat flux less than the full power nominal value. There is a large margin to DNB during the transient since the rod surface heat flux remains below the design value and there is extensive subcooling at all times in the core. The minimum DNBR at all times remains above the limiting value.

Figure 15.2.1-3 shows the response of the average fuel, cladding, and coolant temperatures.

The average fuel temperature increases to a value lower than the nominal full power value.

15.2.1.4 Conclusions In the event of an RCCA withdrawal accident from the subcritical condition, the core and the RCS are not adversely affected since the combination of thermal power and the coolant temperature result in a departure from nucleate boiling ratio (DNBR) well above the limiting value.

15.2.2 UNCONTROLLED ROD CLUSTER CONTROL ASSEMBLY BANK WITHDRAWAL AT POWER 15.2.2.1 Identification of Causes and Accident Description Uncontrolled RCCA bank withdrawal at power results in an increase in the core heat flux.

Since the heat extraction from the steam generator lags behind the core power generation until the steam generator pressure reaches the relief or safety valve setpoinf, there is a net increase in the reactor coolant temperature. Unless terminated by manual or automatic action, the power mismatch and resultant coolant temperature rise would eventually result in DNB, an RCS overpressure condition, or fill the pressurizer with liquid. Therefore, in ...... r-.... -t damage to the cladding', the reactor protection system is designed to terminate any such transient before the DNBR falls below the safety analysis limit values, the RCS pressure exceeds 110% of the design value, or the pressurizer becomes filled with liquid.

The automatic features of the reactor protection system that prevent core- damage.. ensure these limits are not exceeded following the postulated accident include the following:

15.2-5 Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE (1) The power range neutron flux instrumentation actuates a reactor trip if two-out-of-four channels exceed a high flux or a positive flux rate high setpoint. I (2) The reactor trip is actuated if any two-out-of-four AT channels exceed an overtemperature AT setpoint. This setpoint is automatically varied with axial power imbalance, coolant temperature, and pressure to protect against DNB.

(3) The reactor trip is actuated if any two-out-of-four AT channels exceed an overpower AT setpoint.

(4) A high pressurizer pressure reactor trip actuated from any two-out-of-four pressure channels that are set at a fixed point. This set pressure is less than the set pressure for the pressurizer safety valves.

(5) A high pressurizer water level reactor trip actuated from any two-out-of-three level channels that are set at a fixed point.

In addition to the above listed reactor trips, there are the following RCCA withdrawal blocks:

(1) High neutron flux (one-out-of-four)

(2) Overpower AT (two-out-of-four)

(3) Overtemperature AT (two-out-of-four)

Reference 14 documents that generic and conservatively bounding evaluations have been performed to ensure that the RCS overpressure and the pressurizer overfill conditions are not a concern for this event. One evaluation demonstrates that the positive flux rate trip provides adequate protection to ensure that the most limiting RCCA withdrawl event with respect to RCS pressure, does not result in the peak RCS pressure exceeding 110% of the design limit.

The positive flux rate trip setpoint and response time that are credited for this evaluation are listed in Table 15.1-2. Another evaluation demonstrates that the pressurizer water level high trip prevents a pressurizer overfill condition for those RCCA withdrawal events which are very slow and do not generate any other automatic protection signal. The pressurizer water level high trip setpoint credited in the evaluation is listed in Table 15.1-2. The pressurizer water level high trip response time is listed as N/A with the note indicating that the evaluation results are extremely insensitive to the assumed response time.

The generic evaluations of Reference 14 establish that the RCS overpressure and pressurizer overfill criteria are much less limiting, and only the minimum DNBR analysis is described in detail within this section.

The manner in which the combination of overpower and overtemperature AT trips provide fuel cladding protection over the full range of RCS conditions is described in Chapter 7. This includes a plot (also shown as Figure 15.1-1) presenting allowable reactor coolant loop average 15.2-6 Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE 15.2.16 REFERENCES

1. W. C. Gangloff, An Evaluation of Anticipated Operational Transients in Westinghouse Pressurized Water Reactors, WCAP-7486, May 1971.

2. D. H. Risher, Jr. and R. F. Barry, TWINKLE-A Multi-Dimensional Neutron Kinetics Computer Code, WCAP-7979-P-A (Proprietary) and WCAP-8028-A (Non Proprietary), January 1975.

3. H. G. Gargrove, FACTRAN, A Fortran IV Code for Thermal Transients in UO2 Fuel Rod, WCAP-7908-A, December 1989.

4. T. W. T. Burnett, et al., LOFTRAN Code Description, WCAP-7907-A, April 1984.

5. H. Chelemer, et al., Improved Thermal Design Procedure, WCAP-8567-P-A (Proprietary) and WCAP-8568-A (Non-Proprietary), February 1989.

6. Technical Specifications, Diablo Canyon Power Plant Units 1 and 2, Appendix A to License Nos. DPR-80 and DPR-82, as amended.

7. H. Chelemer, et al., Subchannel Thermal Analysis of Rod Bundle Cores, WCAP-7015, Revision 1, January 1969.

8. M. A. Mangan, Overpressure Protection for Westinghouse Pressurized Water Reactor, WCAP-7769, October 1971.

9. J. S. Shefcheck, Application of the THINC Program to PWR Design, WCAP-7359-L, August 1969 (Proprietary), and WCAP-7838, January 1972.

10. T. Morita, et al., Dropped Rod Methodology for Negative Flux Rate Trip Plants, WCAP-10297-P-A (Proprietary) and WCAP-10298-A (Non-Proprietary), June 1983.

11. Westinghouse letter PGE-96-584, Diablo Canyon Units 1 & 2 Spurious Safety Injection Calculation Note, June 1996.

12. Westinghouse letter PGE-96-565, Diablo Canyon Units 1 & 2 Spurious Safety Injection/Pressurizer Safety Valve Water Relief Final Results, May 31, 1996.

13. PG&E Calculation STA-035, "LOFTRAN 10.01 Simulation of Inadvertent SF with one PORV available," February 29, 1996.

14. Westinghouse letter PGE-02-072, Diablo Canyon Units 1 & 2 Evaluation of Reactor Trip Functions for Uncontrolled RCCA Withdrawal at Power,, December 13, 2002.

15.2-45 Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-2 TRIP POINTS AND TIME DELAYS TO TRIP ASSUMED IN ACCIDENT ANALYSES Trip Limiting TripPoint Assumed Time Delay, Function In Analyses sec Power range high neutron flux, high setting 118% 0.5 Power range high neutron flux, low setting 35% 0.5 Power range high positive nuclear power 9% / 2 sec 3.0 rate Overtemperature AT Variable, see 7(a)

Figure 15.1-1 Overpower AT Variable, see 7(a)

Figure 15.1-1 High pressurizer pressure 2445 psig 2 Low pressurizer pressure 1845 psig 2 High pressurizer water level 100%

Low reactor coolant flow (from loop flow 87% loop flow(d) 1 detectors)

Undervoltage trip (b) 1.5 Turbine trip Not applicable 1 Low-low steam generator level 0% of narrow 2(c) range level span High steam generator level trip of the 100% of narrow 2 feedwater pumps and closure of feedwater range level span(e) system valves and turbine trips (a) Total time delay (including RTD time response and trip circuit channel electronics delay) from the time the temperature difference in the coolant loops exceeds the trip setpoint until the rods are free to fall.

(b) A specific undervoltage setpoint was not assumed in the safety analysis.

(c) When below 50% power, a variable trip time delay is utilized as discussed in Section 7.2.1. 1.1.5.

(d) Westinghouse letter PGE-96-582, Diablo Canyon Units 1 & 2 Evaluation of Revised Low Reactor Coolant Flow Reactor Trip Setpoint, June 27, 1996, concludes that a safety analysis setpoint of 85% loop flow, which results in an additional 0.2 second delay, is acceptable.

(e) Westinghouse SECL 92-151, Increased Steam Generator High Level Turbine Trip Setpoint, 7/17/92, states that the analysis assumed 100% narrow range level span for conservatism. It also states that the plant setpoint analytical limit is 82% narrow range level span for type 51 steam generators due to void effects.

I Revision 14 November 2001

DCPP UNITS 1 & 2 FSAR UPDATE TABLE 15.1-2 (f) Westinghouse letter PGE-02-072, Diablo Canyon Units I & 2 Evaluation of Reactor Trip Functions for Uncontrolled RCCA Withdrawal at Power, 12/13/02, documents that a specific response time is not assumed since it is not a sensitive parameter for the generic evaluation results.

Revision 14 November 2001

Enclosure 6 PG&E Letter DCL 03-016 ALLOCATED RESPONSE TIME TABLES

Table I Enclosure #6 Diablo Canyon Units I and 2 PGandE Letter DCL 03-XX Process Channel and Actuation Logic Response Time Allocations FUNCTION SENSOR TIME EAGLEINIS TIME TIME (sec) STRING (sec) (sec)

Power Range Neutron Flux - EXEMPT (Note 3) NIS 0.065 Input + ssps logic 0.020 High Power Range Neutron Flux - EXEMPT (Note 3) NIS 0.065 Input + ssps logic 0.020 Low Power Range Neutron Flux Rate - High Positive Rate EXEMPT (Note 3) NIS 0.2 Input + ssps logic 0.020 (Note 2) --

Power Range Neutron Flux EXEMPT (Note 3) NIS 0.2 Input + ssps logic 0.020 Rate - High Negative Rate Source Range Neutron Flux EXEMPT (Note 3) NIS (Notel) Input + ssps logic 0.020 Overtemperature AT RTDs not in scope (Note 1) E21 0.409 Input + ssps logic 0.020 Overpower AT RTDs not in scope (Note 1) E21 0.409 Input + ssps logic 0.020 Pressurizer Pressure - Low ROSEMOUNT 0.2 E21 0.409 Input + ssps logic 0.020 11 54SH9RC0.2 Pressurizer Pressure - High ROSEMOUNT 0.2 E21 0.409 Input + ssps logic 0.020 I IS4SH9RC0.2 ROSEMOUNT Pressurizer Water Level - 1153HD5RC High (Note 2) ROSEMOUNT 0.2 E21 0.409 Input + ssps logic 0.020 1153HD5RA

Table 1 Enclosure #6 Diablo Canyon Units 1 and 2 PGandE Letter DCL 03-XX Process Channel and Actuation Logic Response Time Allocations FUNCTION SENSOR TIME (sc EAGLEINIS TIG TIME (e) SSPS RELAYS TIME (sc (see) STRING _ (sec) (sec)

Reactor Coolant Flow - Low ROSEMOUNT 0.2 E21 0.409 Input + ssps logic 0.020 Ci53HD5RC Undervoltage RCPs Relays not in scope (Note 1) N/A N/A Input + ssps logic 0.020 Underfrequency RCPs Relays not in scope (Note 1) N/A N/A Input + ssps logic 0.020 Steam Generator (SG)

Waterm LevelratLowow ROSEMOUNT (SG) R M NT 0.5 E21 0.409 Input + ssps logic Water Level - Low Low 1154DP4RCN0033 0.020 (Note 1) Allocation times not used for these variables. These components will continue to be tested as required.

(Note 2) These variables are currently not required to be response time tested.

(Note 3) Neutron detectors are exempt from RTT [TS SR 3.3.1.16]

(Note 4) This function is triggered by an SI actuation. It is not driven by a specific sensor. It is currently not required to be response time tested. See Note 2.

(Note 5) Additional interposing slave relays are Potter and Brumfield MDR type relays Allocated sensor times are derived from method (3) section (9) WCAP-13632 rev. 2 (Vendor Engineering Specifications).

Barton times were provided in table 9-1. Rosemount times are from Rosemount manuals 00809-0100-4388, 4514, and 00809-0100-4631.

Table I Enclosure 6 Diablo Canyon Units 1 and 2 PGandE Letter DCL 03-XX Process Channel and Actuation Logic Response Time Allocations TIME EAGLEINIS TIME TIME SENSOR TIEse) STRING TIE SSPS RELAYS FUNCTION STRING (sec) (sec)

Safety Injection - Actuation Input + SSPS Logic +

of Motor Driven Auxiliary N/A (Note 4) N/A E21 0.409 Master + Slave Relays 0.088 Feedwater (Note 2)

Safety Injection - ROSEMOUNT Input + SSPS Logic +

Containment Pressure - 1154P6RC 0.2 Master + Slave Relays 0.088 High Safety Injection - Pressurizer ROSEMOUNT 0.2 E21 0.409 Input + SSPS Logic + 0.088 Pressure - Low 1154SH9RC Master + Slave Relays ROSEMOUNT Safety Injection - Steam Line 1154SH9RC & 0.2 E21 0.409 Input + SSPS Logic +

Pressure - Low BARTON 763 Master + Slave Relays 0.088 Containment Spray - ROSEMOUNT Input + SSPS Logic +

Containment Pressure - 11 54DP6RC 0.2Master + Slave Relays 0.088 High High Steam Line Isolation Containment HihHih154DP6RC Pressure - ROSEMOUNT 0.2 E21 0.409 Input + +SSPS Master SlaveLogic Relays+ 0.088 High High ROSEMOUNT Steam Line Isolation - Steam RO54SH9RC & 0.2 E21 0.409 Input + SSPS Logic +

Line Pressure - Low BARTON 763 Master + Slave Relays 0.088 Steam Line Isolation - Steam ROSEMOUNT Line Pressure - Negative 1154SH9RC & 0.2 E21 0.409 Input + SSPS Logic + 0.088 RateMaster + Slave Relays 0.

Turbine Trip and Feedwater ROSEMOUNT Isolation - SG HighDP4High33Master water Level - 1154DP4RCN0033 0.5 E21 0.409 Input ++SSPS SlaveLogic Relays+ 0.088 0.8 High High Auxiliary Feedwater - SG ROSEMOUNT Input + SSPS Logic +

Water Level-Low Low 1154DP4RCN0033 0.5 E21 0.409 IP Master logc

+ Slave Relays 0.8 Auxiliary Feedwater - Under- Input + SSPS Logic +

voltage Reactor Coolant (note 1) n/a *Relay Contact n/a Master + Slave Relays 0.124 Pump (note 5)

Table 1 Enclosure 6 Diablo Canyon Units 1 and 2 PGandE Letter DCL 03-XX Process Channel and Actuation Logic Response Time Allocations (Note 1) Allocation times not used for these variables. These components will continue to be tested as required.

(Note 2) These variables are currently not required to be response time tested.

(Note 3) Neutron detectors are exempt from RTT [TS SR 3.3.1.16]

(Note 4) This function is triggered by an SI actuation. Itis not driven by a specific sensor. It is currently not required to be response time tested. See Note 2.

(Note 5) Additional interposing slave relays are Potter and Brumfield MDR type relays Allocated sensor times are derived from method (3) section (9) WCAP-1 3632 rev. 2 (Vendor Engineering Specifications).

Barton times were provided in table 9-1. Rosemount times are from Rosemount manuals 00809-0100-4388, 4514, and 00809-0100-4631.