ML20054L324

Forwards Response to 810929 Request for Info Re Performance Testing of PWR Relief & Safety Valves Per NUREG-0737,Item II.D.1

ML20054L324
Person / Time
Site:	Sequoyah
Issue date:	06/30/1982
From:	Mills L TENNESSEE VALLEY AUTHORITY
To:	Adensam E Office of Nuclear Reactor Regulation
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM NUDOCS 8207070397
Download: ML20054L324 (11)

Text

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' TENNESSEE VALLEY AUTHORITY CH ATTANOOG A. TENNESSEE 37401 400 Chestnut Street Tower II June 30, 1982 Director of Nuclear Reactor Regulation Attention: Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC 20555

Dear Ms. Adensam:

In the Matter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 The Sequoyah Nuclear Plant unit 1 and unit 2 operating license conditions 2.C(23).j(2) and 2.C(16).1, respectively, require TVA to " conform to the EPRI test program" and to " provide documentation for qualifying (a) reactor coolant system relief and safety valves, (b) piping and supports, and (c) block valves in accordance with the review schedule in SECY 81-491. . . ."

Also, NUREG-0737 item II.D.1, as revised by D. G. Eisenhut's September 29, 1981 letter, requires documentation to be suomitted for items (a), (b), and (c) above. A response for item (a) was submitted by letter from me to you on April 1, 1982. Enclosed is our response for item (b), piping and supports, and item (c) block valves, which satisfies the requirements of the license conditions ar.d NUREG-0737.

If you have any questions concerning this matter, please get in touch with J. E. Wills at FTS 858-2683 Very truly yours, TENNESSEE VALLEY AUTHORITY L. M. Mills, Manager Nuclear Licensing Sworn to dsups ibed before me this_ Y day of.6 % X_ _J982

/ 2<t/{ M b h ) . W.

Notary /Public g My Commission Expires . . Jo i t Enclosure cc: See page 2 8207070397 820630 htyk PDR ADOCK 05000327 P PDR An Equal Opportunity Employer

I e .. s Director of Nuclear Reactor Regulation June 30, 1982 00: U.S. Nuclear Regulatory Countission Region II Attn: Mr. James P. O'Reilly, Regional Administrator 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303

ENCLOSURE Performance Testing of Pressurized Water Reactor Relief and Safety Valves (NUREG-0737, Item II.D.1)

Sequoyah Nuclear Plant Units 1 and 2 Operating License Conditions Unit 1 - 2.C(23).j(2); Unit 2 - 2.C(16).1

Table of Contents 1.0 Introduction 2.0 Summary of Valve Evaluations 2.1 Block Valve Performance 2.2 Relier Valve Performance 23 Safety Valve Performance 30 Summary of Piping / Support Evaluations

1.0 Introduction As required by NUREG-0737, Item II.D.1, TVA, as a participating utility in the Electric Power Research Institute (EPRI) Safety and Relier Valve Testing Program, has completed a full-scale test program to demonstrate the functional performance capabilities of the relief and safety valves utilized in the Sequoyah Nuclear Plant's reactor coolant system. The results of the many tests performed by EPRI have been forwarded to the Nuclear Regulatory Commission and form an integral part of this submittal.

The following reports prepared by EPRI are applicable to Sequoyah Nuclear Plant.

a. Valve Selection / Justification Report

b. Test Condition Justification Report

c. Westinghouse Plant Condition Justification Report

d. Safety and Relief Valve Test Report

e. EPRI/ Marshall Motor Operated Valve (Block Valve) Interim Test Data Report Documents a, b, c, and d were transmitted to NRC by David Hoffman of Consumers Power Company by letter on April 1, 1982, and document e was tranmitted to NRC by R. C. Youngdahl of Consumers Power Company by letter on June 1, 1982.

TVA's submittal to the NRC dated April 1, 1982, indicated our preliminary assessment based on the preliminary test data available at that time.

Since then, additional program information has been received. This submittal is based on evaluation of applicable test information received to date. Final test data from EPRI is scheduled to be released on July 1, 1982.

Pertinent design information on the Sequoyah block, relief, and safety valves is shown on Tables 1.0, 2.0, and 3.0, respectively. Additional information is available in the valve selection / justification report. The EPRI inlet and discharge piping is similar to the physical piping geometry at Sequoyah.

2.0 Summary of Evaluations Based on our evaluations to date, TVA has concluded the following:

2.1 Block Valve Performance The block valves at Sequoyah are the Velan Engineering Company's model No. B10-3054B-13MS, 3-inch motor-operated, bolted bonnet gate valve with the Limitorque SMB-00-15 operator for unit 1 and SMB-00-25 for unit 2 (see table 1. 0). The same valve was tested at the Marshall facility. These tests demonstrated the capability of full opening and closing on demand with preloads at pressure up to 2500 psig for all tests. Leakage testing indicated zero leakage at pressures up to 2500 psig. Additional information has been provided to NRC by R. C.

Youngdahl by letter dated June 1, 1982. From the test data presented, TVA concludes that operability of the block valves, as defined in NUREG-0737, item II.D.1, has been satisfactorily demonstrated.

Page 2 2.2 Relier Valve Performance i

The Sequoyah unit 1 PORV is a 2-inch Masoneilan International 20,000 i

, series, air-operated globe valve (see Table 2.0). TVA provided this valve model to EPRI for the testing program. These tests demonstrated the capability of the valve to be fully opened on demand and fully closed on demand under operating conditions similar to that expected at Sequoyah.

The Sequoyah unit 2 PORV is a 3-inch Copes-Vulcan Model No. D-100- ,

160, air-operated globe valve (see table 2.0). This valve was also tested by EPRI. The valve also fully opened on demand and fully closed on demand. These valves were inspected during and after j

testing with no damage observed.

With regard to valve performance for cold overpressure protection conditions, NRC has conditioned each of the Sequoyah operating licenses to require plant modifications for protection against cold overpressurization events prior to startup after the first refueling outages. TVA will ensure that the cold overpressure protection system  !

will meet the Westinghouse performance requirements at the time of installation. .

5 2 3 Safety Valve Perfordhoce The Sequoyah units 1 and 2 safety valves are Crosby HB-BP-86 6M6 l valves with loop seal internals (see Table 3.0). TVA's evaluation of i the EPRI tests is summarized below.

It was shown during testing that the Crosby 6M6 valve exhibits fluttering and/or chattering with liquid loop seal discharge.

Fluttering caused a momentary delay in valve lifting. The results indicate that the delay did not cause an increase in overpressure to i an unacceptable limit. Also, the test valve fluttering during loop

! seal discharge did not result in damage to the valve that would prevent the valve from further operation. The valve remains in stable operation with steam discharge.

I The EPRI test results indicate initial factory recommended ring i settings may not necessarily provide expected valve performance, as I specified by the ASME Code, i.e. , Slowdowns for the 6M6 valve were in j excess of 5 percent (range of 5 to approximately 14 percent).

l Tentatively, Westinghouse has indicated that blowdowns of the magnitude experienced in the 6M6 testing are adequate (considering their transient analyses) and will be addressed more fully in their

, " Safety Valve Operability Report" to be submitted to Westinghouse f

Owners Group utilities. However, the EPRI tests for the 6M6 valve

indicates full rated flow was achieved.

Page 3 TVA has also reviewed the effect of high ramp rates (280 to 375 psi /seo pressurization rates) used during the EPHI tests. These high ramp rates caused the valve to " pop" open (after loop seal discharge) ,

at higher pressures. TVA has reviewed the applicable transients and finds the maximum expected ramp rate to be 144 psi /sec. Extrapolating from the EPRI tests would indicate valve opening pressures of no more than 2600 psia for a ramp rate of 144 psi /sec. This would indicate a maxirsum orer u ' the bottom of the reactor vessel of 2660 psia (consi ing ... piping pressure drop and static head). This figure is below the permissible 110 percent maximum pressure (2750 psia).

Liquid discharge is predicted for only one FSAR transient, the feedline break accident. The temperature of the reactor coolant for Sequoyah is expected to be not less than 650 F for this transient.

The Crosby valve, when tested at this condition, showed acceptable performance.

On extended operation of the safety injection system, subcooled water discharge may occur after the pressurizer becomes water solid. As j reflected in the EPRI test data, the Crosby valve showed undesirable performance at subcooled conditions. However, under these conditions, a " water solid" Reactor Coolant System is not expected to occur for at least 20 minutes. Given this timeframe, successful mitigative action by the reactor operators is expected. Therefore, it is our opinion

! that the Crosby 6M6 valve will perform satisfactorily.

3.0 Summary of Piping / Support Evaluations Qualification of the SQNP-1 and 2 safety and relief valve (S/RV) system piping and supports has been accomplished in part through the benchmarking of the original design stock loads analysis methodology (RELAP4/ MODS) against applicable EPRI S/RV test data. However, the preliminary results '

of this benchmarking evaluation have identified unanticipated physical phenomena occurring at the inlet and discharge of the safety valves. The nature of these new phenomina and their potential impact on SQN design

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pipe stresses and support Joads is discussed below.

Tests of the Crosby Model H3-BP-86 (6M6) Safety Valve have shown oscilliatory behavior of the valve stem during the period when the valve is discharging subcooled loop seal water. These stem oscillations produced high frequency, high magnitude water hammer pressure spikes at the valve inlet. It is our opinion that the oscillatory nature of these pressure spikes (frequencies ranged from 100-250 Hz) and their short duration places them outside the range of applicability of the SQN design basis for piping stress (ANSI B31.1). The coce allowable, for the integrity of the pressure boundary, is based on quasi-statically applied pressure throughout the piping, not on localized. press 1re pulses. These pressure oscillations therefore do not invalidate the original qualification of this portion of the S/RV piping system. In addition, preliminary results indicate that no permanent strains result from these spikes. The pressure spikes therefore pose no threat to the inte;"ity of the SQN unit 1 and unit 2 S/RV piping system.

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I . .

Page 4 Results of the RELAP 4/ MODS benchmarking evaluation indicate that the original SQN design shock loads analysis conservatively predicted piping stresses and support loads for the majority of the S/RV discharge piping.

However, higher than anticipated support loads were measured immediately downstream of the safety valves in applicable EPRI tests. These high loads, which resulted from the discharge of a high density slug of subcooled water through the piping system, were not predicted in the original SQN design analysis. This analysis must therefore be augmented to include consideration of these higher discharge piping loads. Our final evaluation of the magnitude of these water slug loads at SQN unit 1 and unit 2 and their impact upon the existing S/RV piping support designs is currently underway and will be completed by October 15, 1982. Should the results of this study indicate that design modifications are required to ensure complete compliance with the current SQN unit 1 and unit 2 design basis, a schedule for implementation of those modifications will be provided at that time.

1 J

TABLE 1.0 1.0 Block Valves Unit 1 Unit 2 1.1 No. of valves 2 2 1.2 Manufacturer Velan Velan 13 Type model# B10-3054B- B10-3054B-13MS 13MS Bolted bonnet Bolted bonnet motor-motor-operated flex operated flex wedge gate valve wedge gate valves 1.4 Limitorque model number SMB-00-15 SMB-00-25 1.5 Size (inlet, outlet) (3" BW,3"BW) (3" BW,3" BW) 1.6 Design temperature and 680 F, 2485 psig 688 F, 2485 psig pressure Open Close Open Close 1.7 Torque switch setting 2 2 2 2 (specified @ delta P=2750 0 0 psi) 1.8 Attaching inlet pipe (3", 160, 316 SS) (3", 160, 316 SS)

(size, schedule, material) 1.9 Attaching discharge pipe (3", 160, 316 SS) (3", 160", 316 SS)

J TABLE 2.0

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2.0 Relier Valves Unit 1 Unit 2

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2.1 No. of Valves 2 2 1 2.2 Manufacturer Masoneilan Copes-Vulcan 2.3 Type 20,000 series, Air-operated 2" NPS, air-operated globe D-100-160, i globe, standard quick 17-4PH cage, 3" NPS, change trim 316 w/ stellite plug 2.4 Size (inlet, outlet) (3" butt welding end, (3" BW, 3" BW) l 3" butt welding ends) 2.5 Steam flow capacity 210,000 lb/hr 210,000 lb/hr (rated) 2.6 Design temp. & 2485 psig, 688 F 2485 psig, 688 F pressure

2.7 Air supply (spec max) 55 psig 100 psig 2.8 Attaching inlet pipe 3", 160, 316 SS 3", 160, 316 SS

! (size, schedule, material)

2.9 Attaching discharge pipe (3",40, 304 SS) (3", 40, 304 SS)

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TABLE 3.0 3.0 Safety Valves (Unita 1 & 2) 3.1 Number of valves 3 3.2 Manufacturer crosby 33 Type Spring-loaded safety, back pressure compensating valve, HB-BP-86, 6M6 3.4 Size (inlet, outlet, orifice) (6" NPS, 6" NPS, 3.6 12) 3.5 Steam flow capacity (rated) 420,000 lb/hr 3.6 Design temperature and 688 F, 2485 psig pressure 3.7 Inlet flange rating ANSI 1500 lb, 6" 3.8 Discharge flange rating ANSI 600 lb, 6" 3.9 set pressure 2485 psig 3.10 Inlet piping pressure drop 30 psi 3.11 Discharging piping pressure drop 610 psi max based on 3 sarety and 2 relier valves discharging simultaneously 3.12 Attaching inlet pipe (6", 160, 316 SS)

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