ML20058N507

Addresses TS Mod Recommendations of NRC Generic Ltr 89-19, Request for Action Re Resolution of USI A-47, 'Safety Implication of Control Sys in LWR Nuclear Power Plants.'

ML20058N507
Person / Time
Site:	Davis Besse
Issue date:	12/16/1993
From:	Storz L CENTERIOR ENERGY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR 2194, GL-89-19, NUDOCS 9312220085
Download: ML20058N507 (4)

Text

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l% CENTEnson EfWEHtCiY i 300 Madison Avenue Louis F. Storr  ;

Toledo, OH 43652-0001 Vice President-Nuclear .

419 249-2300 Davis-Besse i

b Docket Number 50-346 License Number NPF-3  ;

Serial Number 2194 December 16, 1993

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United States Nuclear Regulatory Commission ,

Document Control Desk l Vashington, D. C. 20555 -

Subj ect: Response to NRC Generic Letter Number 89-19,' Request for Action Related to Resolution of Unresolved Safety 1ssue A " Safety implication of Control: Systems in LVR Nuclear Power l Plants" Gentlemen:

By this letter Toledo Edison addresses the Technical Specification-  ;

modification recommendations of NRC Generic Letter Number 89-19, ,

" Request for Action Related to Resolution of Unresolved Safety Issue A-47" (Log Number 3091 dated September 120, 1989) as applicable to the Davis-Besse Nuclear Power Station Unit 1 (DBNPS). Generic Letter' Number 89-19 discussed that as a result of the NRC staff's technical .

resolution of Unresolved Safety Issue (USI) A-47, certain control  !

system failures need protection, and that selected emergency procedures ,

should assure that plant transients resulting from control system i failures do not compromise public safety. As a result of this technical resolution, the'NRC staff concluded that Pressurized Vater  ;

Reactor (PVR) plants should provide automatic steam generator overfill  ;

protection, and that plant procedures and Technical Specifications  !

(TSs) should include provisions to periodically verify the operability  :

of the overfill protection and to assure that automatic overfill protection is available to mitigate potential Main Feedwater (MPV) overfeed events during reactor power operation.  !

Toledo Edison's response letter, dated March 20, 1990, (Serial Number 1783) stated that the DBNPS has in place a Steam and Feedvater Rupture Control System (SFRCS) trip on high Steam Generator (SG) level that closes the MFV isolation valves, the MTV control valves, the Main Steam Isolation Valves (MSIVs), and initiates Auxiliary Feedvater (AFV).

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Docket Number 50-346 License Number NPF-3  :

Serial Number 2194 l Page 2 This is a safety-grade trip using two actuation channels with a 2-out-of-2 per actuation channel initiating logic. The closure.of the HSIVs removes the main steam supply to the Main Feedvater Pump Turbines (HFPTs) and results in their shutdown. The HFPTs are not tripped separately. The design for the overfill-protection system is '

sufficiently separate from the MFV control-system to ensure that HFV isolation occurs on a steam generator high-water-level signal when required, even if a loss of power, a loss of ventilation, or a fire in '

the control portion of the MFV control system should occur. The

> response further stated that the DBNPS Operating License Appendix A Technical Specifications (TSs) do not include the Steam and Feedvater Rupture Control System SG high-level trip.

Toledo Fdison's letter dated August 12, 1993 (Serial Letter 2164) -

discussed its plans to submit a License Amendment Application that i vould add SFRCS SG high-level trip surveillance testing to the Technical Specifications. This was initially proposed in_ order to credit the trip in the USAR safety analysis. However,-during the course of preparing the license amendment application and assessing the safety impact, it was determined that crediting this trip in the USAR 7 was unnecessary because the USAR safety analysis already provided adequate protection for those accidents or transients where the SFRCS high level trip actuates. In USAR Safety Analysis Section 15.2.10, .

Excess Heat Removal Due to Feedvater System Halfunction, the SFRCS SG high-level trip is referenced, however, the USAR specifically does not take credit for it.

Toledo Edison has performed a review of the Probabilistic Safety Assessment (PSA) associated with Generic Letter 89-19. This review, when utilizing recent Babcock and Vilcox Owners Group and Combustion Engineering Owners Group information, showed that steam generator overfill protection is not a significant contributor to public health and safety. Accident sequences commonly found to dominate risk do not appear in the steam generator overfill sequences and are not of prime  ;

importance in limiting the likelihood or severity of the accident.

Plant-specific PSA risk insights also show that steam generator overfill sequences are not significant within the context of the DBNPS Individual Plant Examination.

The SFRCS SG high-level trip is presently subjected to surveillance testing similar to the safety-grade SFRCS SG low-level trip and -

includes a channel check, channel functional testing and channel calibration (including setpoint verification). The high-level trip is tested during the performance of the technical specification surveillance testing for the low-level trip bistables by surveillance test procedures DB-HI-3245 through 3246 and DB-MI-3237 through 3244.

Based on a review of this past testing, the SFRCS SG high-level trip has proven to be highly reliable. However, should the high level trip  :

be determined to be inoperable during the testing of the SFRCS SG high-level trip bistables, it would be promptly brought to the ,

attention of the DBNPS senior plant management and corrected. )

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Docket Number 50-346 '

License Number NPF ;

Serial Number 2194 [

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The criteria of the NRC's Final Policy Statement on TS Improvements (58FR39132, dated July 22, 1993) were also evaluated by Toledo Edison. ,

The Policy Statement delineates four criteria that establish.the ,

constraints on design and operation of nuclear power plants. appropriate- ,

for inclusion in the TS required by 10CFR50.36. These criteria vere- -

evaluated by Toledo Edison with the.following results:

Criterion 1: Installed instrumentation that is used'to detect, and. .

indicate in the control room, a significant abnormal degradation.of the j reactor coolant pressure boundary.  ;

Toledo Edison Evaluation of Criterion 1: The SFRCS SG'high-level trip is not used to detect a significant abnormal degradation of the reactor coolant pressure boundary.

Criterion 2: A process variable, design feature, or operating' ,

restriction that is an initial condition of a Design Basis Accident (DBA) or transient analysis that either assumes the. failure of or f presents a challenge to the integrity of a fission product barrier.

Toledo Edison Evaluation of Criterion 2: The SFRCS SG high level. trip is not a process variable, design feature, or operating restriction ,

that is an initial condition of a DBA or transient analysis that.either assumes the failure of or presents a challenge to the integrity of a-  !

fission product barrier.

Criterion 3: A structure, system, or component that is part of'the  !

primary success path and which functions or actuates to mitigate a DBA ,

or transient that either assumes the failure of or_ presents a challenge to the integrity of a fission barrier.

Toledo Edison Evaluation of Criterion 3: The SFRCS SG high-level trip is not part of the primary success path which functions or actuates to mitigate a DBA or transient that either assumes the' failure of.or presents a challenge to the integrity of a fission product barrier.

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

Toledo Edison Evaluation of Criterion 4: As discussed earlier, the SFRCS SG high-level trip has not been shown to be significant to public health and safety.

Accordingly, Toledo Edison has concluded.that it is not appropriate to incorporate its plant-specific SFRCS SG high-level trip into the Technical Specifications.

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. Docket Number 50-346 License Number NPF-3

'S'erial Number 2194 Page 4 In order to ensure the concerns in GL 89-19 on maintaining adequate provisions to periodically verify the operability of overfill protection are addressed, Toledo Edison vill continue to perform-testing of the SFRCS high-level trip in accordance with surveillance-test procedures DB-MI-3245 through 3246 and DB-MI-3237 through 3244 as-previously discussed. This testing. vill be tracked as a commitment to  !

the NRC on the Toledo Edison' Regulatory Management System (TERMS). .

Commitments tracked by. TERMS cannot be changed without prior Toledo l Edison management approval. l The NRC letter dated April 18, 1990, to'the Toledo Edison Company (Log Number 3220) Enclosure 2 states the resolution of USI A-47, " Safety -l '

Implication of Control Systems in LVR Nuclear Power Plants," in all respects except for the high steam generator: level' trip technical specifications that are avaiting resolution with'the BWOG. With the '

submittal of this letter and commitment to surveillance testing for the' SFRCS SG high-level trip as previously discussed, Toledo Edison has addressed this remaining item for the DBNPS from Generic Letter 89-19. l Should you have any questions or require additional information, please contact Mr. Villiam T. O'Connor, Manager - Regulatory Af fairs at- 1 (419) 249-2366.

! Very truly yours, FVK/amb -

cc: J. B. Hopkins, NRC Senior Project Manager .

J. B. Martin, Regional Administrator, NRC Region III  !

S. Stasek, DB-1 NRC Senior Resident Inspector j Utility Radiological Safety Board 1

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