ML20206G006

Forwards Response to NRC 870309 Request for Addl Info Re NUREG-0737,Item II.D.1, Performance Testing of BWR & PWR BWR & PWR Relief & Safety Valves

ML20206G006
Person / Time
Site:	Wolf Creek
Issue date:	04/08/1987
From:	Withers B WOLF CREEK NUCLEAR OPERATING CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM WM-87-0112, WM-87-112, NUDOCS 8704140420
Download: ML20206G006 (5)

Text

.

NUCLEAR OPERATING CORPORATION April 8, 1987 U. S. Nuclear Regulatory Comission ATTN: Document Control Desk Washington, D. C. 20555 Letter: WM 87-0112 Re: Docket No. 50-482 Ref: Eetter dated 3/9/87, from PNO'Connor, NRC to BDWithers, NCNOC SubJ: NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves Centlemen:

The Reference requested that additional information be provided in support of the NRC staff review of NUREG-0737, Item II.D.1, " Performance Testing of Boiling-Water Reactor and Pressurized-Water Reactor Relief and Safety Valves", for Wolf Creek Generating Station.

Attached is Wolf Creek Nuclear Operating Corporation's response to questions 1, 2, 3 and 5 transmitted by the Reference. As discussed with the Staff, a response to question 4 requires additional time to prepare. It is anticipated that this response will be provided by May 29, 1987.

If you have any questions concerning this matter, please contact me or Mr.

O. L. Maynard of my staff.

Very truly yours, l

Bart D. Withers President and Chief Executive Officer BDW:jad Attachment cc: PWO'Connor (2)

RDMartin JCumins 8704140420 870408 ,Oh PDR ADOCK 05000482 P PDR (

PO. Box 411 1 Burbngton. KS 66839 i Phone. (316) 364-8831 l An Eoual opro tundy Employer M F HC. VET l

t _ _ _

Attachment to WM 87-0112 Page 1 of 4 April 8,1987 WOLF CREEK NUCLEAR OPERATION CORPORATION RESPONSE 10 NRC REQUEST POR ADDITIONAL INFORMATION CONCERNING NUREG-9737, ITIM II.D.1

" PERFORMANCE TESTING OF BOILING-MATER REAC10R APO PRESSURIZED-WhTER RFACTOR RELIEF AND SAFETY VALVES" Question 1 Results from the series of EPRI tests performed on the Crosby 6M6 safety valve indicated the need for an inspection and maintenance program to assure continual operability of the valve. Four of the loop seal tests in the series were terminated due to chatter on closure. After each such test the valve was found to have galled guiding surfaces and damaged parts which required refurbishment or replacement in the tests imediately following repair of the valve, the valve performed acceptably but then experienced the same chatter problem on a subsequent test. Thus, the licensee is requested to develop a formal procedure requiring inspection of the safety valves after each lift involving a loop seal or liquid discharge. This procedure should be incorporated- into plant operating procedures or licensing documents such as the plant technical specifications.

Response

By applying the EPRI applicability criteria (ring settings, backpressure, inlet pressure drop, and inlet fluid conditions), five loop seal tests of the Crosby 6M6 safety valve were determined to be applicable to Wolf Creek.

These are tests 929, 931, 1406, 1415, and 1419. In all these tests except 1415, the valves exhibited the typical loop seal discharge characteristics of flutter or chatter at partial opening during loop seal liquid passage and

, then stable performance on steam discharge. Test 1415 dmonstrated stable valve performance throughout the test, including during loop seal passage. I Of these tests, only one (1419) was terminated because of valve chatter on  !

closure.

1 The effects of pressure oscillations resulting from loop seal passage were j addressed in the response to NRC question 15 in SLNRC 86-07, dated 6/30/86. l Also, safety valve stability for the SNUPPS plants (Callaway and Wolf Creek) was discussed in the response to NRC Question 5 in SLNPC 86-07. The information provided supports the conclusion that Wolf Creek safety valve performance should be more stable than the tested valves. Based on the responses to these questions, the loop seal passage and steam discharge events are not expected to result in damage to the safety valve from either pressure oscillations or instability at closure. Thus, inspection of safety valves after each loop seal discharge is not warrented.

Potential liquid discharge through the safety valves following a postulated feedwater line break (FLB) or extended high pressure injection (HPI) event was discussed in the response to NRC question 8 in SLNRC 86-09, dated

9/26/86. In that response, it was concluded that the Wolf Creek safety valves are expected to perform acceptably for the water relief conditions 1

Attachment to NM 87-0112 Page 2 of 4 April 8,1987 predicted in the FLB analysis. It was also concluded that water relief is not predicted in the analysis of the the exte'ided HPI event. Therefore, a comnitment to inspect safety valves after each lift involving a liquid discharge is not warranted.

It should be noted that the absence of a formal conunitment to inspect safety valves does not imply that no inspections would be performed. Wolf Creek Nuclear Operating Corporation evaluates incidents and transients that occur at Wolf Creek and initiates appropriate actions based on these evaluations.

Should an evaluation of a safety valve lift event determine that an inspection were required to assure valve operability, the inspection would be performed.

. Question 2 The licensee has previously informed the NRC of a malfunction of a Garrett PORV in a preoperational test conducted during hot functional testing at the Wolf Creek plant. The valve failed to close on steam after operating for 32 seconds in a cold loop seal discharge test. The problem was caused by differential thermal expansion between the valve cage and valve body bore.

Thus, the cages of the plant PORVs were machined to reduce their outside diameters. Kansas Gas and Electric then received NRC approval to conduct a repeat test after fuel loading but prior to initial criticality. The licensee is requested to provide evidenced that the valve functioned

. properly in the repeat test.

1 ,

Response ,

Condition 2.C. (1), Attachment 1, Item 1.C. to Facility Operating License  ;

NPF-32 required that prior to achieving initial criticality Wolf Creek Generating Station establish by additional testing, the capability to close

the pressurizer power operated relief valves following a pressure blowdown l of approximately 200 psig starting at normal operating pressure. This t

license condition was established to track resolution of NRC Inspection Report, 50-482/84-55, open item 482/8455-04. The Wolf Creek Garrett PORVs were successfully restested prior to initial criticality. The license

condition and Inspection Report open item were closed in NRC Inspection Report 50-482/85-23, based on NRC Inspector observation of testing and review of applicable documentation, i Question 3 The licensee has provided information on how the thermal hydraulic analysis was performed but did not indicate what flow rates through the safety valves or PORVs were calculated in the analysis. Provide the flow rates determined

! for the safety valves and PORVs and compare them to the rated flows measured

during the EPRI/CE tests. The calculated flows should be enough greater j than rated flow to account for ASME derating of the valves arrl uncertainties i in the flow rates.

4 i

1 t

- - , . - , , - - ,,_,._,_,,w--. ..- ,.- , .._, .-, .n - . . , - - - - - , , - - , - _ , , , - - , . , , - - . , - -

Attachment to WM 87-0112 Page 3 of 4 April 8, 1987

Response

The purpose of the thermal hydraulic analysis, sulmitted by SLNRC 83-002, dated 1/7/83, and further discussed in SLNRC 86-09 dated 9/26/86, was to demonstrate the integrity of the inlet and outlet piping and supports for the pressurizer safety and relief valves. Westinghouse developed analytical methods to simulate the piping system response to static and dynamic loading conditions including safety and relief valve discharge events. The analytical methods were verified by modelling the EPRI safety valve test piping system, applying the analytical methods and comparing the results with the maasurements taken during the tests.

The Wolf Creek safety and relief valve piping was analyzed by applying the Westinghouse methodology to a model of the plant-specific piping system.

The safety and relief valve flow rates were conservatively treated in the analysis as shown in the following table.

EPRI Wolf Creek Rated Wolf Creek Analysis Flow (lb/hr) Flow (lb/hr) Flow (lb/hr)

(EPRI piping) (Plant piping)

Crosby 467,000 420,000

• 6M6 @ 2575 psia 0 2575 psia Garrett 378,000** 213,000

• PORV @ 2760 psia 0 2350 psia

• For the analysis, a conservative factor of over 1.20 was included in the maximum rated valve mass flow rate.

• • Wyle test, no data available at 2350 psia The ERPI tests of the safety valve demonstrated the relationship between the rated flow and full capacity flow since, by definition, ASME rated flow is 90% of full capacity. Care should be taken when comparing EPRI test flow rates to Wolf Creek analyzed flow rate because the piping configurations are different. Nevertheless, the factor of over 1.20 applied in the analysis adequately addresses flow uncertainties.

Question 4 The submittals from the licensee provide only scant information on the pipe supports. Thus, further information is requested as follows:

A. The load combinations in Tables 2-1 and 2-2 of the January 7,1983 subnittal are said to govern both the piping and supports. The stress limits identified in the tables, however, are applicable to

. _ _ _ _ . . . . . = . . _ . . . _ _ - . . . _ - . __ ._ _ . _ _ ___ _ _ ___ __-_..__ _ -__

! Attachment to NM 87-0112 Page 4 of 4 j April 8, 1987 pipe stresses but are not typical of stress limits for pipe supports. Thus, clarify the stress limits used in the load i combinations for the supports and identify the governing standards i or codes. Also, verify that the load ccmbinations listed in Tables 2-1 and 2-2 were indeed used in the support evaluations.

l B. Present a table comparing the maximum loads (or stresses) in

! several representative supports with appropriate allowable loads (or stresses). Provide conparisons for normal, upset, emergency i and faulted conditions. Indicate in the Table the support number j and support type, and supply sketches to show locations of the j supports listed in the table. ,

Response

A response to this question requires additional time to prepare. It is j anticipated that this response will be provided by May 29, 1987.

I j Question 5 j Provide the pipe size for Class 1 piping upstream of the valves. Is the 3

piping 6 inch Schedule 160?

l j Response i'

The piping upstream of the safety valves is 6 inch Schedule 160. Refer to Table 1 of SLNRC 86-09, dated 9/26/86 or to Section 6.2.1 of the report subnitted by SLNRC 83-002, dated 1/7/83. ,

k j

i 1

4 t

1 i

__ _ _ .._, . __,., _ . . . . _ , , . . _ , , _ - . . . , _ _ . _ - - - _ . __ . . _ ~ . , . , - - - _ . . . . - _ _ . .

_ - _ _ - - - . , _ , - - . . . _ . , .