ML19261D803

Forwards Responses to Questions Raised by NRC During 790608 Site Visit Re Procedural Gaps,Performance Effect of Flow on Power Operated Relief Valves,Valve Error Correction & Const Cleanliness

ML19261D803
Person / Time
Site:	Davis Besse
Issue date:	06/18/1979
From:	Roe L TOLEDO EDISON CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
NUDOCS 7906260431
Download: ML19261D803 (4)

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Docket No. 50-346 -

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License No. NPF-3 Serial No. 316 June 18, 1979 Director of Nuclear Reactor Regulation Attention: Mr. Robert N. Reid, Chief Operating Reactors Branch No. 4 Division of Operating Reactors United States Nuclear Regulatory Commission 5.'a s h in g t on , D. C. 20555

Dear Mr. Reid:

During the June 8, 1979 NRC site visit to the Davis-Besse Nuclear Power Station, Unit 1, several questions were raised by your staf f. Attached are the requested responses.

Yours very truly, LER:TJM cc:

Mr. Guy Vissing Operating Reactors Branch No. 4 Division of Operating Reactors U. S. Nuclear Regulatory Commission Vashington, D. C. 20555 Attachment}) jk}

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,e - Docket No. 50-346 Attachment 1 License No. NPF-3 Serial No. 516 June 18, 1979 Response to Question Raised at NRC Site Visit June 3, 1979 at Davis-Besse Nuclear Power Station Unit 1 Question 1 In relation to the recent start up problem at Arkansas Nuclear One - Unit 1, how do Davis-Besse operators handle procedural gaps?

Response

All procedural steps are followed as described in the appropriate test pro-cedure. If the operator nceds to do additional steps not provided by his procedure he must request a temporary modification as required by Administra-tive Procedure AD 1805. Section 7.3.4 of AD 1805 describes the form, require-ments and review of a Temporary Modification request. Any such modification to a procedure must receive interim approval for implementation f rom two members of the station management staff. One of these staf f members must hold a Senior Reactor Operator's License. Review and final approval by the Station Review Board is required within 14 days. Question 2 Describe the performance effect of flow on the PORV as a result of the September 24, 1977 event.

Response

The actions taken to restore the PORV to service are described on page 41 of Davis-Besse Nuclear Power Station, Unit 1 Licensee Event Report Supplement NP 7 7- 16 dated November 14, 1977. For flow ef fects, the valve only received seat 5 disc lapping. Question 3 What is being done to preclude similar valving errors to those made in January 1979?

Response

3bnual valves important to plant safety and operations are now controlled under the locked valve verification procedure described in item 5 of Toledo Edison's letter of April 11, 19 79 (Serial No. 1-56). Motorized valves with controls located remote from the control room are being designed to have locked covers installed over their remote c en t rols . Padlock control over the manual handwheels associated with these motorize 6 valves have been installed and are cont rolled under the station's locked valve procedure as above. f 44

Docket No. 50-346 Attachment 1 License No. NP F- 3 (continued) Serial No. 516 J une 18, 1979 Ouestion 4 5l hat is being done to insure that construction activity cleanliness does not effect safety related valves similar to that of Licensee Event Report NP-33-79-03 of January 2, 1979?

Response

Changes are being proposed to Administrative Procedure AD 1844.05 to enhance station superivison of maintenance areas to identify potential cleanliness problems. Question 5

 '.ih a t is being done to preclude similar problems exhibited when personnel traffic caused a misalignment of a panel switch to an intermediate position?

Response

All non-positive detent switches at the Davis-Besse facility have been identi-fled and are being scheduled for replacement with positive control spring detented switches. Additionally, for those panels located in high traffic areas, personnel protection barriers are being installed in the interin to pre-clude inadvertant misalignment. Question 6 Document the tines when auxiliary feedwater system was controlled in manual.

Response

Attachment 2 lists ten events since January, 1978 in which the operater manually controlled steam generator level below its automatic setpoint. None of the events involved an SFAS actuation. 2311 345

Docket No. 50-346 Attachment 2 License No.NPF-3 Serial No. 516 June 18, 1979 AFU Manual Control Events Date of Event Brief Description January 6, 1978 SFRCS trip due to MFPT speed control trouble, reactor power manually reduced to 17 full power, see LER NP-33-78-06. January 21, 1978 Manual reactor trip following tripping of both MEPs and SFRCS trip. January 31, 1973 MSIVs inadvertantly closed, reactor tripped on high RCS pressure. SFRCS trin due to loss of MFPT. March 1, 1978 SFRCS trip on high main feedwater - steam generator differential pressure. Reactor trip on RCS high pressure. October 10, 1978 SFRCS trip due to MFPT speed control trouble. Reactor trip on low RCS pressure. November 3, 1978 TPS00.04 Natural Circulation Test. January 12, 1979 Reactor to p on high flux / delta flux / flow trip of RPS. SFRCS trip due to steam generator low level. See LER NP-33-79-13. January 14, 1979 TPS00.25 Shutdown From Outside The Control Room Test. January 15, 1979 TP800.26 Loss of Offsite Power Test. Februa ry 22, 1979 Manual reactor trip following EHC system backup speed control circuit f ailu re . SFRCS trip on main steam low pressure. 2311 346

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