ML17263A506

Discusses Closeout of GL 89-19, Request for Action Related to Resolution of Unresolved Safety Issue A-47 Safety Implication of Control Systems in LWR Nuclear Power Plants Per 10CFR50.54(f)

ML17263A506
Person / Time
Site:	Ginna
Issue date:	12/21/1993
From:	Andrea Johnson Office of Nuclear Reactor Regulation
To:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
References
REF-GTECI-A-47, REF-GTECI-SY, TASK-A-47, TASK-OR GL-89-19, TAC-M74945, NUDOCS 9401030033
Download: ML17263A506 (7)

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Docket No. 50-244 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 December 21, 1993 Or. Robert C. Hecredy Vice President, Nuclear Production Rochester Gas and Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

CLOSEOUT OF GENERIC LETTER (GL) 89-19, "REQUEST FOR ACTION RELATED TO RESOLUTION OF UNRESOLVED SAFETY ISSUE A-47 'SAFETY IHPLICATION OF CONTROL SYSTEHS IN LWR NUCLEAR POWER PLANTS'URSUANT TO 10 CFR 50.54(f)"

(TAC NO. H74945)

At Ginna Station an advanced digital feedwater control system (AOFCS) installation was completed and placed into operation during the 1991 annual refueling outage, upon receipt and issuance of Amendment No.

41 to Facility Operating License No.DPR-18, dated February 28, 1991.

Verification that the design criteria for the ADFCS installation is acceptable, was performed by the Region on Hay 20-24, 1991 (Inspection No. 50-244/91-13),

and November 26-30, 1990 (Inspection No. 50-244/90-29).

The NRC staff concluded that the AOFCS met all of the applicable guidelines and regulations, and that its utilization to replace the existing analog feedwater control

system, discussed in the amendment Safety Evaluation, is acceptable.

Your letter, dated October 27,

1993, responded to an NRC request for additional information (RAI), dated October 18, 1991, regarding the steam generator (SG) overfill protection at Ginna Station and your completed implementation of the recommendations outlined in Generic Letter 89-19.

The NRC understands that at Ginna Station SG overfill protection is initiated on high water level and isolates main feedwater (HFW) flow to the SGs by closing the HFW control and bypass valves (HFW isolation).

However, SG overfill protection circuits share common level instruments with the ADFCS median signal selector (HSS) but are separated by qualified isolation devices, consequently HFW control perturbation problems could be linked to unreliable SG overfill protection.

The RAI requested an update on this matter with regard to your original response letter of Harch 19, 1990, which addressed the recommendations of Generic Letter 89-19, applicable to only the then existing analog feedwater control system.

Additionally, the RAI recommended testing actions on the ADFCS for one cycle of plant operation to evaluate the reliability of the ADFCS, to evaluate any impact on the SG overfill protection circuitry.

Two related events have occurred during this period which could have possibly affected SG overfill protection as follows:

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Robert C. Necredy December 21, 1 993 I'I (1)

On November ll, 1991, spurious signals occurred in the ADFCS r'esulting in feedwater flow oscillations and an increase in SG water level causing a

NFW isolation.

However, operators stabilized SG level and dampened NFW flow on manual control.

A condensate booster pump trip resulted as a

result of the NFW transient and a

IOX turbine runback occurred.

The root cause was determined to be the starting of a diesel driven fire pump where pump cables are run in close proximity to ADFCS cables.

The appropriate corrective action to prevent recurrence using diode suppression techniques was accomplished and verified by the Region on November ll to December 9,

1991 (Inspection No. 50-244/91-27).

The event has not recurred.

(2)

On Hay 18, 1992, the HFW flow overfed SG A while operators attempted to stabilize SG water level in the manual

mode, due to a secondary system (heater drain tank level control) pressure transient.

The NFW control perturbations caused a

SG high level and HFW flow isolation.

The appropriate corrective action was accomplished after the SG level was stabilized and the HFW valves returned to automatic control.

Included in the corrective action plan were requirements for Westinghouse to evaluate and adjust the ADFCS "tuning constants" used in the ADFCS software.'owever, the root cause later identified the HFW flow irregularity as a

hardware problem requiring valve packing adjustments, valve actuator snubber adjustments, and disassembly and repacking during the 1993 annual refueling outage.

Longer term corrective actions were also hardware oriented and considered for an upgrade to the HFW regulating valve internals.

The corrective action plan has been verified, and continues to be monitored by the Region on Hay 18, 1992 (Inspection No. 50-244/92-08).

The staff concludes that during these two events the SG overfill protection circuitry performed its intended function.

Your response letter to the RAI on October 27, 1993, togetherwith a low number of operational events during the first cycle of plant op'eration after installation of the ADFCS, confirms that your pl'ant provides a satisfactory design for SG overfill protection.

Your respons'e 1'etter which addresses SG overfill and ADFCS power sources, cabinet loc'ation, 'cable routing, and Technical Specification Limiting Conditions for Operation/Surveillance Testing, provides an adequate basis to consider NRR s'eview of USI A-47

'omplete.

TAC No. N74945 is hereby closed.

t incerejy

.'TgTna,1, signed by:

Allen R. Johnson," Project Hanager Project Directorate, I-3 Division of Reactor Projects, I/II, "

Office of Nuclear 'Reoacto'i '-Regulation I

cc:

See next page

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Robert C. Mecredy December 21, 1993 (1)

On November ll, 1991, spurious signals occurred in the ADFCS resulting in feedwater flow oscillations and an increase in SG water level causing a

HFW isolation.

However, operators stabilized SG level and dampened HFW flow on manual control.

A condensate booster pump trip resulted as a

result of the MFW transient and a lOX turbine runback occurred.

The root cause was determined to be the starting of a diesel driven fire pump where pump cables are run in close proximity to ADFCS cables.

The appropriate corrective action to prevent recurrence using diode suppression techniques was accomplished and verified by the Region on November 11 to December 9,

1991 (Inspection No. 50-244/91-27).

The event has not since recurred.

(2)

On Hay 18, 1992, the HFW flow overfed SG A while operators attempted to stabilize SG water level in the manual

mode, due to a secondary system (heater drain tank level control) pressure transient.

The HFW control perturbations caused a

SG high level and HFW flow isolation.

The appropriate corrective action was accomplished after the SG level was stabilized and the HFW valves returned to automatic control.

Included in the corrective action plan were requirements for Westinghouse to evaluate and adjust the ADFCS "tuning constants" used in the ADFCS software.

However, the root'ause later identified the MFW flow irregularity as a.

hardware problem requiring valve packing adjustments, valve actuator snubber adjustments, and disassembly and repacking during the 1993 annual refueling outage.

Longer term corrective actions were also hardware oriented and considered for an upgrade to the HFW regulating valve internals.

The corrective action plan has been verified, and continues to be monitored by the Region on Hay 18, 1992 (Inspection No. 50-244/92-08).

The staff concludes that during these two events the SG overfill protection circuitry performed its intended function.

Your response letter to the RAI on October 27, 1993, together with a low number of operational events during the first cycle of plant operation after installation of the ADFCS, confirms that your plant provides a satisfactory design for SG overfill protection.

Your response letter which addresses SG overfill and ADFCS power sources, cabinet location, cable routing, and Technical Specification Limiting Conditions for 'Operation/Surveillance

Testing, provides an adequate basis to consider NRR's review of USI A-47 complete.

TAC No. M74945 is hereby closed.

Sincerely, cc:

See next page Al en R. Johnso

, Project Manager Pr ject Direct rate I-3 Div'on o eactor Projects I/II Office o Nuclear Reactor Regulation

Dr. Robert C. Hecredy R.E.

Ginna Nuclear Power Plant CC:

Thomas A. Hoslak, Senior Resident Inspector R.E.

Ginna Plant U.S. Nuclear Regulatory Commission 1503 Lake Road

Ontario, New York 14519 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, Pennsylvania 19406 Hs.

Donna Ross Division of Policy Analysis 8 Planning New York State Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Charlie Donaldson, Esq.

Assistant Attorney General New York Department of Law 120 Broadway New York, New York 10271 Nicholas S.

Reynolds Winston 5 Strawn 1400 L St.

N.W.

Washington, DC 20005-3502 Hs. Thelma Wideman

Director, Wayne County Emergency Hanagement Office Wayne County Emergency Operations Center 7370 Route 31

Lyons, New York 14489 Hs. Nary Louise Heisenzahl Administrator, Honroe County Office of Emergency Preparedness 111 West Fall

Road, Room ll Rochester, New York 14620

DISTRIBUTION:

Docket File NRC 5 Local PDRs PDI-3 Reading SVarga JCalvo WButler SLittle AJohnson KCotton OGC ACRS (10)

JLinville, RI JWermiel KJabbour,LPH LJCallan