ML20054L037

Advises That Submittal of Repts Re Qualification/Operability of Pressurizer power-operated Relief Valves & Safety Valves. Provides Response to NUREG-0737,Item II.D.1.A

ML20054L037
Person / Time
Site:	McGuire
Issue date:	06/30/1982
From:	Parker W DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8207070093
Download: ML20054L037 (12)

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DUKE POWEH COM1%NY l'owru Uusunwo 4a2 Sourn Cuuncu STnent, CHANIDTTE, N. C. una4a WI LLI AM O. PA R M E R, J R.

V'C r Pat s.o t % v Tet gewo%c; Ant A 704 sec.- e.couctio June 30, 1982 nr oea Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: McGuire Nuclear Station Docket Nos. 50-369, 50-370

Dear Mr. Denton:

The attached reports address the qualification / operability of pressurizer power operated relief valves (PORVs), safety valves, PORV block valves and associated piping and supports. Specifically, these reports address the applicability of the generic Electric Research Power Institute (EPRI) testing program to McGuire Nuc1 car Station by plant specific analysis and reference to EPRI and Westinghouse reports.

With the submittal of this information, paragraph 2.C.(ll).j of the McGuire Facility Operating License is satisfied.

Please advise if there are questions concerning the attached information.

Ver truly yours, i

N William O. Parker, s

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I GAC/php Attachment cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Rcgion II 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. P. R. Bemis Senior Resident Inspector McGuire Nuclear Station 7070o

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, MCGUIRE NUCLERR STATION SAFETY VALVE OPERABILITY REPORT lVREG 0737, ITEM II.D.I.A.

l This report is in response to the requirements of NUREG 0737, Item II.D.l. A.

to demonstrate Pressurizer Safety Valve operability for McGuire Nuclear Station.

The basis for this report is the generic safety / relief valve test program that was developed on behalf of participating utilities by the Electric Power Research Institute (EPRI). Duke Power has been a participant in the EPRI program, and the results of this program were submitted to the NRC on April 1,1982 on behalf of utilities by David flof fman.

VALVE DESCRIPTION MAN'lFACTURER -

Crosby Valve and Gage Company TYPE -

Spring Loaded Safety Valve MODEL -

IlB-BP-86, 6M6 MANUFACTURERS DRAWING -

DS-C-56925 VALVES PER UNIT -

3 Each safety valve is connected to a separate pressurizer nozzle with the valve inlet piping run to form a water loop seal.

EPPI TEST PROGRAM Duke Power has been a full participant in the EPRI test program, and the McGuire safety valve is represented by the program results submitted to the NRC on April 1, 1982 in the following reports.

1) Valve Selection Justification Report

2) Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse - Designed Plants.

1) Test Condition Justifi' cation Report

4) Safety and Relief Valve Test Report

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McGuire Nuclear Station.is also represented by Westinghouse's evaluation of valve operability presented in WCAP 10105, "Sa fety Valve Operability Report."

WCAP 10105 will be submitted to the NRC on July 1, 1982 by the Westinghouse Owner's Group on behalf of the participating utilities.

APPLICABILITY OF EPRI TEST RESULTS The EPRI test results are applicable to licGuire Nuclear Station as follows:

Test Valve - McGuire safety valves and the test valve are both, Crosby, Model llB-BP-86, size 6M6 nozzle type relief valves with similar designs and specifications.

Fluid Conditions - The tested conditions covered all the design basis transient events for the McGuire pressurizer relief system. In addition, the test backpressures envelope those at McGuire.

Test Facility - The test facility and the McGuire pressurizer relief system in the safety valve region are similar with respect to piping configuration, pipe size, and loop seal arrangement.

VALVE OPERABILITY The Crosby 6M6 valve successfully opened and closed for all design basis test transients. In addition, rated flow was exceeded.

For tests outside of the design basis, highly subcooled water flow, the valve performance was less than optimum. Westinghouse analysis (reference Valve Inlet Condition Report) has shown that for extended high pressure injection events a n.inimum of 20 min to 6 hrs is required to fill the pressurizer solid.

This time period provides adequate time for mitigation of the event by operator action. .

For McGuire, the plant loop seal inlet piping is shorter than that used in the EPPI testing. As stated in the Test Condition Justification Report, this will make the McGuire valve operation more stable than the test valve. The McGuire

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water loop seal volume is also less than the test facility. This shortens the time period of the oscillatory motion experienced on a loop seal lift, which shortens the valve opening delay time.

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Based on the test data and the above discussion, operability of the McGuire safety valve has been demonstrated for expected operating and accident conditions as required by flVREG 0737. Item II.D.1. A.

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-POWER OPERATED RELIEF VALVE QUALIfICAT10ft REPORT flUREG 0737, ITEM II. D.I.A.

This report is in response to the requirements of NtlREG 0737, Item II.D.I. A. to demonstrate Power Operated Relief Valve (PORV) operability for McGuire Nuclear Station. The basis for this report is the generic safety / relief valve test program that was developed on behalf of participating utilities by the Electric Power Research Institute (EPRI) combined with in-house and in-plant testing per-formed by Duke Power. Duke Power has been a participant in the EPRI program, and the results of this program were submitted to the NRC on April 1,1982 on behalf of the utilities by David floffman of Consumers Power.

POPV SYSTEM DESCRIPT10fl The pressurizer is equipped with three power operated relief valves which limit system pressure for a large power mismatch condition. The relief valves are operated at a high setpoint either automatically or by remote manual control.

The operation of these valves also limits the undesirable opening of the spring-loaded safety valves.

The power-operated relief valves are also used for overpressure protection during periods of water solid operation. Protection against such over-pressuriza-tion events is provided through the addition of low pressure setpoints to two PORV's. Since this protection is required only during low temperature water solid operation, the low pressure setpo, int, 400 psig, is enabled by the operator at reactor coolant loop temperatures below 320 F. The low pressure setpoint is interlocked with reactor coolant loop temperature to minimize the possibility of inadvertent actuation.

VALVE DESCRIPTI0fl Manufacturer - Control Components International Type - 3 inch Air Operated Globe Valve Model - Self Drag Element

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118760 Opera tor -

Air Piston (air to open; spring / air to close)

The CCI PORV uses an air operated piston type actuator. Normal operation uses

• air to open and close the valve. A closure spring is also provided in the actuator to automatically close the valve in the event of loss of air pressure.

BACKGROUND As a result of PORV problems in the initial McGuire hot functional test, Duke performed full scale in-house PORV testing at Marshall Steam Station. An identical PORV from Catawba Nuclear Station was used as a test valve. In-conjunction with the valve manufacturer, a series of modifications and proof tests were performed to improve valve performance. The modifications included:

1) revising bonnet design from pressure seal to bolted bonnet

2) increasing operator size to provide additional spring closure force in event of loss of air supply.

3) continuously draining upstream water loop seal to reduce thermal transient loading on valve disc stack

4) revising air supply system to decrease valve operator stroke time The above problem and modifications were documented with the NRC under the requirements of 10CFR50.55e via significant Deficiency peport No. 50369-370/79-01 dated April 13, 1979 and supplemented on August 15, 1979.

After completion of all modifications, a series of proof tests were performed and the valve met all operating requirements. The tests were documented in a Duke report, " Report of Pressurizer PORV Operability Test", dated February 5 1. 6, 1980. ,

The above modifications were then made to all six (6) McGuire valves. During a subsequent hot functional test, the three Unit i valves were operated at full pressure and temperature and met all operating requirements including valve stroke time.

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EPRI TEST PROGRAM Duke Power has been a full participant in the EPRI test program, and,as such, the McGuire PORV testing is represented in the following EPRI reports submitted to the NRC on April 1, 1982.

1) Value Selection Justification Report

2) Valve Inlet Fluid Conditions for Pressurizer Safety and Relief.

Valves in Westinghouse Designed Plants

3) Test Condition Justification Report

4) Safety and Relief Valve Test Report Items of a plant specific nature not covered in the above reports are addressed below.

TEST VALVE APPLICABILITY The EPRI test valve is the same valve Duke used for in-house testing and is functionally identical to the McGuire valves. Therefore, the EPRI test results are directly applicable to McGuire.

VALVE FLUID CONDITIONS The cold overpressure conditions in the Westinghouse fluid Inlet Condition Peport do not specifically address McGuire, since the valves and controls wern purchased by Duke Power. However, the McGuire system is based on the Westinghouse design.

The PORV high setpoint is 2335 psig, with liquid discharge temperatures ranging from 320 F to 650 F. The low setpoint is 400 psig with liquid discharge tempera-tures ranging from 100 F to 450 F.

Liquid tests representative of the above conditions were performed. Tests number 38 and 43 of table 3-13 in the Test Condition Justification Report are For the reduced pressure representative of the high set point liquid conditions.

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condition, tests 39 and-40 are representative. The valve opened and closed as required for each test.

Based on the above tests, the fluid conditions for which the CCI valve was tested represent the range of conditions fnr McGuire Cold Overpressurization events. ,

VALVE OPERABILITY The criterial for successful operation is that the valve opens and closes for all fluid conditions and passes the specified flow. The CCI valve successfully met this criteria for every test.

For the abnormal operating mode of unassisted spring closure on loss of all air, several delays in valve closing were observed in the Wylie EPRI testing that were not observed in the Marshall (EPRI) steam testing. Five (5) steam ' tests were run at Wylie with closure on spring only. The valve had delays of 1 to 3 seconds on 3 of the tests. After the Marshall testing, the valve was disassembled, cleaned, reassembled, and repacked. Delays in the Wylie testing were attributed to the tightness to which the valve was assembled prior to testing. Valve opera-tion after reassembly was not as smooth as in the Marshall testing.

Six (6) water tests were also run with spring only closure. The valve closed with no delay on four tests and had delays of 20 and 40 seconds on the other two.

The last two water tests were high pressure water at saturated temperature conditions (?650 F). Valve closure did occur when inlet pressure decreased to 2035 psia. Evaluations with the valve manufacturer determined that high tempera-ture water flashing across the drag element creates higher pressure in the valve throat than expected. Under spring only closure, this throat pressure is sufficient to prevent valve closure until inlet pressure falls to approximately 2035 psia.

This condition occurs only for high pressure saturated water in the range of 650 F. This was verified by a 640 F high pressure water test, in which no valve delay occurred on spring only closure.

Duke considers the above valve performance as acceptable for the following reasons:

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1) the occurrence of 650 F water at the valve inlet is a degraded i

event of low probability.

2) the valve has been sh an inlet pressure of gwn to close on spring force alone by test at i 2035 psia for this case. This closure pressure is above reactor trip and safety injection setpoints.

3) the delay will only occur on the improbable set of circumstances that the valve has opened with air, had a transistion to high temperature water, and then had a complete loss of air prior to receiving a close signal. Any air remainTng in the air lines will assist in normal closure.

CONCLUSION Based on the above testing, operability of the CCI power operated relief valve has been demonstrated for expected operating and accident conditions as required by NUREG 0737 Item II.D.I.A.

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, MCGUIRE NUCLEAR STATION PORV-BLOCK VALVE QUALIFICATION REPORT NUREG 0737 ITEM II. D.1.B.

This report is in response to the requirements of NUREG 0737, Item II. D.I.B pertaining to qualifications of Pzr PORV block valves for McGuire Nuclear Station.

Duke Power has met this requi,ement through a combination of full scale 'in-house testing, in-plant' testing, and participation in the Electric Power Research Institute PORV Block Valve Test Program.

Valve Description Manufacturer -

Nuclear Valve Division of Borg Warner Type -

3" Gate 1523 lb pressure class Model -

79294 Manufacturer Drawing -

74380-1 Opera tor -

Rotork 16NA)1 Valves Per Unit -

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Background

Duke first performed full scale testing of the PORV Block Valve in conjunction with in-house testing of the PORV at Marshall Steam Station. The test valve used was obtained from Catawba Nuclear Station, and was identical to the PORV block valves installed at McGuire. In initial tests, the valve failed to fully close against flow and Duke notified the NRC under the requirements of 10CFR50.55e via Significant Deficiency Report No. 369-370/79-14 on January 16, 1980. Duke evaluated the closing problem with the valve vandor, and severa1 modifications to the valve were made to improve performance. These modifications were:

1) increased operator size from Rotork 14NAX1 to 16NAX1 ,

2) installed disc guide rails in the valve body

3) contoured valve disc to reduce friction

4) strengthened stem to disc connection 4

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The modified valve was'successfully retested in a 21 cycle evaluation test at full pressure and temperature. Successful completion of this test was documented in an in-house report, " Report of EMO Valve Operability Test",

dated January 30, 1980. The. modified valve-operator combination was shown by test to have an excess torque available margin of '75%.

The above modifications were then made to all six (6) valves installed at McGuire. The three Unit 1 valves were then closed successfully against flow at full pressure and temperature during subsequent hot functional testing.

EPPI Test Program Duke has been a full participant in the EPRI safety / relief valve test program, including the Block Valve Program. Duke's modified test valve was also used in the EPRI Block Valve Test Program. The valve was cycled in additional 21 times against flow at full pressure and temperature and operated success fully for each cycle.

The results of this program has been documented in a ~ report entitled "EPRI PWR Sa fety and Relief Valve Test Program PORV Block Valve Information Package."

This report will be transmitted to the NRC by Mr. David Hoffman on behalf of Duke Power and the other participating utilities on or before July 1,1982.

CONCLUSIONS

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l The McGuire PORV Block Valve, after modification, has been cycled 21 times at full pressure and temperature under a Duke test program and 21 times at full pressure and temperature under the EPRI program. In addition, the three McGuire Unit I valves have been cycled against flow at full reactor coolant pressure and temperature.

For each test, the valve successfully closed and , reopened.

These tests confirmed McGuire PORV Block Valves operability for all fluid condi-tions and meet the requirements of NUREG 0737. Item II. D.I.B.

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SUBMITTAL FOR NUREG 0737

'S/RV PIPING AND SUPPORTS This report presents the preliminary evaluation of the Pressurizer Safety and Relief (S/RV) System piping of McGuire Nuclear Station, Unit 1, as required by NUREG 0737. An evaluation is underway to substantiate the adequacy of the Pressurizer S/RV piping and supports in compliance with NB-3600 of Section III of the ASME Code and Duke Support / Restraint Design Procedures. This evaluation is based on preliminary applicable results of test data obtained from the EPRI Safety and Relief Valve Test Program (Ref.1). The final report on the test results is schedules for completion on July 1,1982.

Qualification of the S/RV piping and supports on McGuire 1 was originally accomplished through comparison of calculated loads (steady and transient) and pipe stresses resulting from design basis events with applicable design allow-ables. As part of our effort to comply with the requirements of NUREG 0737, the original analysis has been benchmarked against applicable EPRI S/RV test data. The results of this benchmarking evaluation indicate that the original McGuire design analysis conservatively predicted piping stresses and support ~

loads for the majority of the S/RV piping. This analysis should be augmented to consider unanticipated physical phenomena at the discharge segment of the SV piping (Ref. 2). Additionally, unanticipated oscillatory behavior of the Crosby 6M6 Safety Valve resulted in high frequency, high magnitude pressure oscillation imediately upstream of the safety valves.

A preliminary review of the piping upstream of the safety valves subjected to the pressure oscillation after valve discharge indicates.that pipe stresses are within ASME Code allowables. A similar review of the safety valve discharge piping (Duke Class E) ar.d the upstream and downstream FORV piping indicates the piping is adequate.

A review of S/RV discharge piping support load: resulting from the unanticipated physical phenomena (water surge load) on valve opening is underway. Based on anticipated receipt of the Final Piping Data from the EPRI Test Program in July,1982, Duke Power Company will submit a final evaluation of the piping associated with the Relief and Safety Valves on November 1, 1982. This final evaluation will provide a schedule for implementation of any modifications that may be required.

References:

1) " Runs 903, 908, 917, Crosby 6M6 Safety Valve Loop Seal Configur,ation" Preliminary Piping Data Package, EDS Design Input Document No. '0092-325-DIN, EDS Nuclear, Inc., Dec. 1981.

2) " Evaluation of EDS Nuclear Pressurizer Safety and Relief Valve Analysis Methods", EDS Report No. 01-0092-1210, Rev. O, June, 1982.

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