ML20215G724

Responds to 860731 Request for Addl Info Re NUREG-0737, Item II.D.1,concerning Valve Operability & Thermal Hydraulic & Structural Analysis of Inlet & Discharge Piping.Complete Responses to Questions 2,3,4 & 10 Provided

ML20215G724
Person / Time
Site:	Davis Besse
Issue date:	10/14/1986
From:	Williams J TOLEDO EDISON CO.
To:	Stolz J Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM 1308, NUDOCS 8610210077
Download: ML20215G724 (8)

Text

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TOLEDO EDISON JOE VVILUAMS. JR.

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Docket No. 50-346 I41912d9 23m (419] P49 5d23 License No. NPF-3 Serial No. 1308 October 14, 1986 Director of Nuclear Reactor Regulation Attention: Mr. John F. Stolz PWR Project Directorate #6 Division of PWR Licensing - B United States Nuclear Regulatory Commission Washington, DC 20555

Dear Mr. Stolz:

This is in response to your letter of July 31, 1986 (Log No. 2044) con-cerning NUREG-0737 II.D.1 - Request for Additional Information (RAI). The questions related to (1) valve operability and (2) thermal hydraulic and structural analysis of the inlet and discharge piping. Toledo Edison has reviewed the submitted questions and will provide a complete response to questions 2, 3, 4 and 10 and a partial response to question 5 at this time. Response to questions 1, 6, 7, 8, 9, 11, 12, 13, 14, 15, 16, 17 and 18 will be provided by December 31, 1986 and to question 5 two months after completion of the testing.

Very truly yours, oe 'il ams , Jr.

gbenior Vice President, Nuclear JW:MGF Attachment cc: DB-1 Resident Inspector A0 THE Tauf DO EOGON COMPANY EDISON FUZA 300 MADGON AVENUE TOLEDn Osru 43650

D cket No. 50-346 Licensa No. NPF-3 Serial No. 1308 October 14, 1986 '

Page 1 Question 2. NUREG-0737, Item II.D.1 requires qualification of the PORV control rircuitry of the plant specific PORVs for design-basis accidents and transients. Response 8 to the request for additional information dated June 7,1985 was insuffi-

.cient in discussing the utilities qualification of the PORV circuitry to NUREG-0737 II.D.I. The NRC Staff has agreed, however, that such qualification does not have to be submitted for this review if it has already been included in the submittal to fulfill the requirements of 10 CFR 50.49. Verify whether the in-plant PORV control circuits have been included in the 10 CFR 50.49 review. If the PORV circuitry has not been reviewed under 10 CFR 50.49, provide the following to demonstrate that the requirements of NUREG-0737 Item II.D.1 concerning the control circuitry have been met:

A. Provide a list of all PORV control circuitry needed to mitigate NUREG-0737 transients such as the following:

1. Switchgear

2. Motor control centers

3. Valve operators and solenoid valves

4. Motors

5. Logic equipnent

6. Cable

7. Connectors

, 8. Sensors (p . sure, pressure differential, temper-ature, flow and level, neutron, and other radiation)

9. Limit Switches

10. Heaters

11. Fans

12. Control boards

13. Instrument racks and panels

14. Electric penetrations i

Docket No. 50-346 Lic:n:2 Ns. NPF-3 Serial No. 1308 October 14, 1986 Page 2

15. Splices

16. Terminal blocks B. For each. item of equipment identified in A, provide the following:

1. Type (Functional designation)

2. Manufacturer

3. Manufacturer's type number and model number

4. Plant ID/ tag number and location C. For each item c.f equipment listed in above, provide the environmental envelope, as a function of time, that includes all extreme parameters, both maximum and minimum values, expected to occur during NUREG-0737 transients, including postaccident conditions.

D. For each item of equipment identified above, state the actual qualification envelope simulated during testing (defining the duration of the environment and the margin in excess of the design requirements). If any method other than type testing was used for qualifi-cation, identify the method and define the equivalent

" qualification envelope" so derived.

E. Provide a summary of test results that demonstrates the adequacy of the qualification program. If any

- analysis is used for qualification, justification of all analysis assumptions must be provided.

F. Identify the qualification documents that contain detailed supporting information, including test data, for items D and E.

Response The PORV is not required for mitigation of a design basis accident at Davis-Besse. There is no design basis accident as listed in Chapter 15 of the Davis-Besse USAR for which the PORV is required. The safety function of the PORV is to maintain the Reactor Coolant System boundary. The PORV is operated by a'125 VDC, Channel B non-essential power supply.

The block valve is environmentally qualified in accordance with 10 CFR 50.49 (E.Q. Rule) and is powered by Class IE Channel 1 versus the PORV which is Channel B (non-essential).

Dockat Na. 50-346 Licanza No. NPF-3 Serial No. 1308 October 14, 1986 Page 3 Channel 1 and Channel B are not bundled together when located in the same enclosure or routed in same tray or conduit. Thus independence or separation of the two circuits is provided which meets the single failure criteria.

On February 6,1986, Toledo Edison reported to the ACRS on the operability qualification of the PORV solenoid for conditions expected during extended feed and bleed opera-tions (not a design basis accident). Those test conditions included:

• Testing environment of 220 F and 15 psig.

• Initial 20 minutes on/off cyclic operation.

• 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> continuously energized.

• Solenoid mounting bracket connected to a thermal source at 650*F.

• Solenoid plunger loaded to represent value operation.

The purpose was specifically for qualification of the solenoid and to assure continuous open condition during feed and bleed with an extended period of time when the containment begins to get hot at a higher pressure.

Question 3. As stated in response 5 to the request for additional information dated June 7, 1985, Davis-Besse identified two occasions where the PORV failed to operate on demand. What actions are being considered by Davis-Besse to ensure reliable operation of the PORV? Provide a discussion on these actions such as disassembly and inspection of the PORV after a challenge and lift. In addition, describe what will be_done to ensure the valve is in good working order and all parts properly machined so that problems, such as those encountered during EPRI testing and past problems at the plant, will not prevent the PORV from functioning properly.

Response After the valve actuation on June 9, 1985, an extensive program.of investigation was performed. A visual inspec-tion of the PORV and associated linkage vis performed in order to check for broken or missing parts, boric acid buildup, or other abnormalities. This inspection was performed by Davis-Besse Maintenance and Crosby Field Service personnel and no abnormalities were observed which could have any effect on the valve's operability.

D:ckst No. 50-346 Licznsa N2. NPF-3 Serial No. 1308 October 14, 1986 Page 4 The valve was then disassembled aad a complete visual, functional and dimensional inspection of the internal parts were performed using a checklist prepared by Crosby. The results of this inspection are as follows:

A. Minor steam cutting had occurred on the pilot seat and disc.

B. Minor wear marks were observed on the guide lands for the main disc.

C. A brown substance (possibly boric acid) was observed on the valve body in the vicinity of the pilot valve.

D. A sliver of metal from a flexitallic gasket and a small gouge were found on the outside edge of the bellows housing gasket surface.

E. All other inspection results indicated normal condition.

An inspection of the PORV actuation circuitry was performed using a detailed checklist and no abnormalities were observed which would have any effect on the operability of the PORV.

Detailed inspection results indicate that the PORV and its associated circuitry are normal. It is possible that the

' failure to reclose during its third actuation on June 9,

. 1.985, was a result of interference between internal parts

caused by the presence of a foreign substance. If this happened, this substance was subsequently dislodged by the fluid hydraulic forces thereby allowing.the valve to later reclose. The existence of close tolerance is an inherent feature of a valve of this type. The failure rate experi-enced to date with the Davis-Besse PORV is consistent with the industry-wide failure rate of 0.02 per challenge (NUREG 0560, Pg. 3-14). During the plant startup (est.

October 1986) confirmation of operability will be accom-i plished by actuating the PORV at reduced and full pressure.

During future plant operation, the PORV will be exercised at reduced pressure during plant shutdowns at the following frequency:

• For operating intervals of three (3) months or longer, exercise during each shutdown.

• For intervals less than three (3) months, exercising is not required unless three (3) months have passed since last shutdown exercise.

j' _ Docket No. 50-346

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Licance No. NPF-3 Serial No. 1308 October 14, 1986 Page 5

• If the valve was actuated in the course of plant operation at a frequency which satisfies the above 1 requirements, additional exercising during cold shutdown is not required.

• The PORV. disc movement will lue confirmed by exercising the valve while observing the PORV solenoid position 4

and flow indicators.

Crosby PORVs that were used during the EPRI testing were

. procured and refurbished to use for operability and flow testing in October and November of 1985.

TEST - OCTOBER 18, 1985 The October 18, 1985 PORV test was divided into two parts.

The first part consisted of 50 valve cycles with the test

facility configured for high PORV downstream backpressure l (approximately 425 psig). The second part consisted of 50 i valve cycles with the test facility configured for low PORV

, downstream backpressure (approximately 160 psig). -The 3 valve actuated successfully for all 100 cycles.

TEST - NOVEMBER 15, 1985 1

The November 15, 1985 PORV Test consisted of 20 valve cycles. The test facility was configured for high PORV i downstream backpressure (approximately 500 psig). The purpose of the second test was to verify a desired higher flow rate. The valve actuated successfully for all 20 cycles.

! A Facility Change Request (FCR 85-0160) was implemented to l' install a drain line in the PORV inlet piping. The purpouc of this drain is to eliminate the loop seal, since there

. was some difficulty with maintaining the minimum required j loop seal temperature. Analysis of the system without a

! loop seal shows that the system meets its design basis requirements without additional hardware changes to the 1 system. The valve manufacturer has confirmed that the valve will operate and perform its intended function satisfactorily without a loop seal. The EPRI report NP-2628-SR, confirms that operation without a loop seal has no adverse impact on valve overability. A drained loop seal (i.e. no loop seal present) decreases the temperature dif ferential af ter valve actuation and would be expected to enhance valve reliability.

In addition to all the above, the valve was rebuilt with a complete set of new internals and a new solenoid. <

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Dockst No. 50-346 Licenze No. NPF-3 Serial No. 1308 0ctober 14,-1986 Page 6 Question 4. The submittal did not provide the expected backpressure during PORV discharge. Provide the expected PORV backpressure.

Response Calculations show the backpressure to be.less than 550 psia for both steam and water conditions with valve inlet conditions of 2500 psia. The calculations show the flow to be choked during PORV actuation.

Question 5. EPRI tests on the Velan gate valve with the SMB-000-10 actuator demonstrated successful operation with torques as low as 82 ft-lbs. For purposes of comparison, provide the present Davis-Besse block valve torque switch settings and the corresponding torques produced. If the torque produced by the plant block valve operator is less than 82 ft-lbs, it is the staff position that it is not adequate to con-clude proper operation based solely on manufacturer's calculations. The problems encountered with the Westing-house gate valve on closing, which was traced to the calculations used to size the operator torque requirements, indicate the need to experimentally verify the adequacy of the block valve / operator combination. Toledo Edison should provide test data to demonstrate the SMB-00-10 operators at Davis-Besse are capable of providing adequate torque.to close the block valves.

Response Due to the large number of options available between Limitorque operator and valve combination, it is not possible to directly compare valve and actuator combina-tions unless gear ratios, stem dimensions, spring packs and valve designs are taken into account.

Davis-Besse has adopted a comprehensive motor operated valve test program which includes the calculation of the required thrust to close a valve against its design differ-ential pressure. For the 2 inch Velan gate valve with an SMB-00-10 operator installed as the PORV block valve at Davis-Besse, 6,390 lbs. of thrust (calculated) is required to close against a differential pressure of 2,525 psig.

l The thrust applied in closing the valve at the point of I torque switch trip has been measured at 7,795 lbs. These j thrust values can be converted to a torque value by use of the valve " stem factor". This conversion for the valve installed at Davis-Besse results in the calculated thrust i of 6,390 lbs. being equal to 77.6 ft.-lbs. and the measured

. thrust of 7,795 lbs. being equal to 94.7 ft.-lbc.

! As documented on page 3-10 in NUREG-1154, entitled " Loss of

! Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985", the Davis-Besse block valve was success-fully closed and re-opened under actual seevice conditions.

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Dockst No. 50-346 4- Licrasa No. NPF-3 Serial No. 1308 ,

October 14, 1986 Page 7 Adequacy of the block valve to close against full rated flow and pressure will be demonstrated in the pre-startup testing. Test results will be submitted two months after completion of the. testing.

Question 10. Nozzle stresses on page 148 of TES report TR-5639-2 did not include the effect of unequal rupture disc bursting pres-sure. Evaluate the stresses when this effect is considered.

Response. TES report TR-5639-2 (Ref #13 of NRC submittal dated July 31, 1986)' contains the requested calculation on page 145. Resulting stress levels were within long term USAR allowables.

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