ML20245E921
| ML20245E921 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/31/1988 |
| From: | Fineman C, Nalezny C, Seaquist C EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20245E924 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-8224, NUDOCS 8902020177 | |
| Download: ML20245E921 (28) | |
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EGG-NTA-8224 ,
TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING DAVIS-BESSE UNIT 1 DOCKET NO. 50-346 C. A. Seacuist C. P. Fineman C. L. Nalezny August 1988 Idaho National Engineering Laboratory EGtG Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the
- U.S. Nuclear Regulatory Comission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6492
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ABSTRACT Light water reactors have experienced a number of occurrences of improper performance of safety and relief valves installed in the primary coolant system. As a result, the authors of NUREG-0578 (TMI-2 Lessons
- Learned Task Force Status Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions. This report documents the review of these programs by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc. Specifically, this report documents the review of the Davis-Besze Nuclear Plant Licensee response to the requirements of NUREG-0578 and NUREG-0737. This review found the Licensee provided an acceptable resconse thereby reconfirming General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met.
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CONTENTS l
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ABSTRACT ............................................................... ii 1.
INTRODUCTION ..................................................... 1 1.1 Background ................................................. I 1.2 General Design Criteria and NUREG Requirements ............. 1 2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM ................ 4 3.
PLANT SPECIFIC SUBMITTAL ......................................... 6 4
l REVIEW AND EVALUATION ............................................ 7 4.1 Valves Tested .............................................. 7 4.2 Test Conditions ............................................ 9 4.3 Operability ................................................ 12 4.4 Piping and Support Evaluation .............................. 18 5.
EVALUATION
SUMMARY
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REFIRENCES....................................................... 24 iii 1
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.,[ TECHNICAL EVALUATION rep 0RT.
TMI-ACT!0N--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING FOR 4
DAVIS-BESSE NUCLEAR POWER PLANT DOCKET NO. 50-346-i
- 1. INTRODUCTION.
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1.1 Background
Light water' reactor experience has included a number of instances of improper performance of relief and safety valves installed in the primary coolant . systems. There were instances of valves opening below set pressure, valves opening above set pressure, and valves failing to open or restat.
From these past instances of improper valve performance, it is not known whether they occurred because of a limited qualification of the valve or because of.a basic unreliability of the valve. design. It is known that the failure of a power-operated relief valve (PORV) to resent was a significant contributor to the Three Mile Island (TMI-2) sequence of events. These facts led the task force which prepared NUREG-0578 (Reference 1) and, subsequently, NUREG-0737 (Reference 2) to recommend that programs be developed and executed which would reexamine the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions. These programs were deemed necessary to reconfirm that the General Design Criteria 14, 15, and.30 of Appendix A to Part 50 of the Code of Federal Regulations, 10 CFR, are indeed satisfied. .
1.2 General Design Criteria and NUREG Requirements General Design Criteria 14, 15, and 30 require that (1) the reactor primary coolant pressure boundary be designed, fabricated, and tested so as to have an extremely low probability of abnormal leakage, (2) the reactor
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coolant system and associated auxiliary, control, and protection systems be 1
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. 1 designed with sufficient margin to assure that the design conditions are not exceeded during normal operation or anticipated transient events, and (3) the components which are part of the reactor coolant pressure boundary shall be constructed to the highest quality standards practical.
l To reconfirm the integrity of overpressure protection systems and thereby assure that the General Design Criteria are met, the NUREG-0578 l
posi, tion was issued as a requirement in a letter dated September 13, 1979, j
i by the Division of Licensing (DL), Office of Nuclear Reactor Regulation (NRR), to ALL OPERATING NUCLEA'R POWER PLANTS. This requirement has since been incorporated as Item II.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements (Reference 2), which was issued for implementation on October 31, 1980. As stated in the NUREG reports, each pressurized water reactor Licensee or Applicant shall:
- 1. Conduct testing to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
- 2. Determine valve expected operatinr, conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Rev. 2.
- 3. Choose the single failures such that the dynamic forces on the safety and relief valves are maximized.
- 4. Use the highest test pressures predicted by conventional safety analysis procedures.
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- 5. Include in the relief and safety valve qualification program the qualification of the associated control circuitry.
- 6. Provide test data for Nuclear Regulatory Commission (WRC) staff review and evaluation, including criteria for success or failure of valves tested.
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Submit a correlation or other evidence to substantiate that the valves tested in a generic test program demonstrate the functionability of as-installed primary relief and safety valves.
This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (FSAR). The effect of as-built relief and safety valve discharge piping on valve operability must be considered.
- 8. Qualify the plant specific safety and relief valve piping and supports by comparing to test data and/or performing appropriate analysis.
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- 2. PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM In response to the NUREG requirements previously listed, a group of utilities with PWRs requested the assistance of the Electric Power Research Institute (EPRI) in developing and implementing a generic test program for pressurizer power operated relief valves, safety valves, PORV block valves, and associated piping systems. Toledo Edison, the owner of the Davis-Besse Nuclear Plant, was one of_the utilities sponsoring the EPRI Valve Test Program. _The results of the program are contained in a group of reports which were transmitted to the NRC by Reference 3. The applicability of these reports is discussed below.
l EPRI developed a plan (Reference 4) for testing PWR safety and relief valves under conditions which baund actual plant operating conditions.
EPRI, through the valve manufacturers, identified the valves used in the overpressure protection systems of the participating utilities.
Representative valves were selected for testing with a sufficient number of the variable characteristics that their testing would adequately demonstrate the performance of the valves used by utilities (Reference 5). EPRI, through the Nuclear Steam Supply System (NSSS) vendors, evaluated the FSARs of the participating utilities and arrived at a test matrix which bounded the plant transients for which overpressure protection would be required (Reference 6).
The utilities participating in the EPRI Safety and Relief Valve Test Program also obtained information regarding the performance of PORV block valves (Reference 7). A list of valves used or intended for use in participating PWR plants was developed. Seven block valves believed to be representative of the block valves utilized in the PWR plants were selected for testing. Additional tests were performed by Westinghouse ~
Electro-Mechanical Division (WEMD) on valve models they manufacture (Reference 8).
EPRI contracted with Babcock and Wilcox Company (B&W) to produce a report on the inlet fluid conditions for pressurizer safety and relief 4
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i valves in Babcock and Wilcox designed plants (Reference 9). Since s
., Davis-Besse was designed by B&W, that report is relevant to this evaluation.
Several test series were sponsored by EPRI. PORVs and bicek valves I
were tested at the Duke Power Company Marshall Steam Station located in L Terrell, North Carolina. Only steam tests were conducted at the Marsh'all Station. Block valves, therefore, were only tested for full flow, full l
. pressure, steam conditions at Marshall. Water flow tests were performed by WEMD on four valve models they manufacture. Conditions ranged from 60 to 600 gpm and 1500 to 2000 psi differential pressure. Additional PORV tests were conducted at the Wyle Laboratories Test Facility located in Norco, l California. Safety valves were tested at the Combustion Engineering -l l
Company, Kressinger Development Laboratory, which is located in Windsor,
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Connecticut. The results for the relief and safety valve tests are reported '
in Reference 10. The results for the block valve tests are reported in References 7 and 8.
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The primary objective of the EPR1/C-E Valve Test Program was to test each of the various types of primary system safety valves used in PWRs for the full range of fluid conditions under which they may be required to operate. The conditions selected for test (based on analysis) were limited to steam, subcooled water, and steam to water transition. Additional
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objectives were to (1) obtain valve capacity data, (2) assess hydraulic and structural effects of associated piping on valve operability, and (3) obtain
, piping response data that could ultimately be used for verifying analytical piping models, i
Transmittal of the test results meets the requirement of Item 6 of Section 1.2 to provide test data to the NRC.
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.- 3. PLANT SPECIFIC SUBMITTAL l-An initial assesse. ant of the adequacy of the overpressure protection-system and prope:;ed plant modification was submitted by Toledo Edison Company on July 1, 1982 (Reference 11). A preliminary evaluation of the discharge piping and supports and safety valve, PORV, and block valve operability was submitted on December 17, 1982 (Reference 12). A detailed evaluation report on the safety valve and PORV piping and safety valve, PORV, and block valve operability followed on February 2, 1983 (Reference 13). A request for additional information was submitted to i To kdo Edison by the NRC on June 7, 1985 (Reference 14). Toledo Edison !
submitted a partial response on July 19, 1985 (Reference 15) with the l
remainder of the requested information submitted October 1, 1985 and December 6, 1985 (References 16 and 17). A second request for additional information was sent to Toledo Edison on July 31, 1986 (Reference 18) to which the Licensee responded on October 14, 1986 (Reference 19) and February 24, 1987 (Reference 20).
The response of the overpressure protection system to Anticipated Transient Without Scram (ATVS) and the operation of the system during feed and bleed decay heat removal are not considered in this review. Neither the Licensee nor the NRC have evaluated the performance of the system for these events.
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- 4. REVIEW AND EVALUATION 4.1 Valves Tested The Davis-Besse Nuclear Power Station Unit 1 is a B&W 177-FA raised loop plant with low head high pressure injection (Hp!) pumps (maximum pressure < 1900 psi). The unit's two safety valves are Crosby HB-BP-B6 4M1 6 self-actuated valves with a design set pressure of 2525 psia and a rated capacity of 376,000 lbm/h. A plant modification was completed which removed the loop seals and mounted the safety valves directly on the pressurizer nozzles. Each safety valve discharges through a tee, fitted with 75 psig rupture discs, attached to the valve outlet flange. At safety valve lift, the rupture discs open directing safety valve discharge to the containment atmosphere. The safety valves were also modified by replacing the loop seal internals with steam internals.
The unit's one power operated relief valve (PORV) is a 2 1/2 x 4 in.
Crosby HPV-SN solenoid actuated pilot operated valve with a 11/2 in.
diameter bore. The PORV has a design set pressure of 2465 psia and a flow capacity of 158,000 lbm/h. A drained loop seal is upstream of the PORV.
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The PORV is not used for low temperature overpressure protection. l The unit's one PORV block valve is a 2 1/2 in. Velan gate valve F9-454-B-13MS with a Limitorque SMB-00-10 electric actuator.
J The Davis-Beste safety valve is a Crosby 4M1 6 valve with steam internals. The plant safety valves have typical plant ring settings of
-275, -18 for valve RC13-1, and -300 -18 for valve RC13-2 as measured from the locked reference position. The plant-specific safety valve was not tested by EPRI. However, Crosby safety valves Model 3K6 and 6M6, with steam and loop seal internals, respectively, were tested in the EPRI test program. The valves were tested on short and long inlet piping l configurations with drained loop seals which envelop the Davis-Besse installation. The Davis-Besse 4M 1 6 safety valve is similar to the tested 6M6 valve with the exception of the orifice size, valve body construction, inlet flange size, and disc holder construction and material. The 4
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,7 difference in orifice size only affects capacity and not operability. The body construction and inlet flange configuration, forged for the 6M6 and cast for the 4M16, and cisc holder construction and material are not significant factors it determining valve operability. The 4M 1 6 is similar to the EPRI 3K6 test valve with the exception of orifice size and inlet flange size and pressure rating. The orifice size difference only affects' capacity, not operability. The inlet flange size and rating are not significant factors in determining valve operability. The results from the EPRI tests can therefore be used to demonstrate operability of the
. Davis-Besse safety valve.
The Davis-Besse PORV is a Crosby model HPV-SN with a 1 1/2 in. diameter nozzle bore and was not tested by EPRI. However, a Crosby HPV-SN 1 3/8 in.
nozzle bore diameter valve was tested by EPRI. The valves are similar functionally but do include design differences other than the bore size.
The differences between the EPRI test PORV and the Davis-Besse PORV include body configuration, pilot valve location and assembly, and main valve guide retention. The design changes from the Davis-Besse PORV design were made to improve ease of manufacturing (Reference 5). These differences do nct affect valve operability. The difference in bore size affects valve relief capacity and not operability. The PORV installation configurations are similar. The Davis-Besse PORV is installed in a vertical run of pipe, and the EPRI test PORV was tested in a vertical run of piping. Therefore, the test PORV is considered an adequate representation of the Davis-Besse in-plant PORV.
The Davis-Besse block valve is a Velan Model F9-454-B-13MS with a Limitorque SMB-00-10 electric motor actuator. The block valve / actuator combination was not tested by EPRI. However, EPRI did test two 3 in. Velan -
Model B10-3054-B-13MS gate valves. One was tested with a Limitorque SB-00-15 operator and the other was tested with a Limitorque SMB-000-10 .
l operator. The differences between the EPRI tested block valves and the 1-Davis-Besse block valve is the end connections (Butt weld vs Flanged), body rating (1500 lb ANSI vs 2500 lb ANSI), and valve size (3 in, vs 21/2 in.).
The gate valves in all other respects are similar and the above noted differences do not affect operability. The 00 actuator differs from L
a 000 actuator in physical size, where the 00 actuator is larger and can accept. larger valve stem diameters. Thus, the Davis-Besse SMB-00-10 and the EPRI.SMB-000-10 actuators are essentially identical with the exception of the physical size of the actuator. The block valve is installed in the Davis-Besse plant in a horizontal piping configuration and the valve tested by EPRI was in a similar horizontal piping configuration. *he valve is designed.for'use in either a horizontal or vertical orientation. Therefore, the EPRI test valves are representative of the Davis-Besse in-plant block valve.
Based on the above, the valves tested are considered to be applicable to the in-plant valves at Davis-Besse and to have fulfilled the criteria of Items 1,and 7 of Section 1.2 regarding applicability of the' test valves.
4.2 Test Conditions The valve inlet fluid conditions that bound the overpressure transients for Babcock-Wilcox designed PWR plants are identified in Reference 9, and discussed in Reference 20. The transients considered below are for FSAR transients resulting in steam discharge only. The low temperature overpressurization event was not included because it does not challenge the safety valves and Davis-Besse does not use the PORV for low temperature overpressure protection.
For FSAR transients resulting in steam discharge, the safety valves experience a peak pressure of 2677 psia and a maximum pressurization rate of 175 psi /s based on a Rod Ejection accident at hot zero power. The backpressure expected is less than or equal to the 75 psi nominal burst pressure of the rupture discs on the safety valve discharge tees I
(Reference 13).
l The Crosby 3K6 safety valve with steam internals was subjected to ,
i 1 14 tests with short inlet piping. Of these tests, four (406, 408, 411, 442)
L are applicable to the Davis-Besse inlet conditions with steam discharge.
The ring settings used in these tests were the manufacturer's recomended ring settings (-55, -14) and are typical PWR plant ring settings (ring 9
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)1 setting reference is relative to the bottom of the disc ring). In these i tests.the 3K6 valve popped open at pressures of 2456 to 2489 psia, the pressurization rates for the applicable tests were 2.5 to 314 psi /s, l
backpressure 602 to 678 psia, and the peak tank pressure was 2683 psia.
The Crosby 6M6 safety valve with loop seal internals was subjected to 17 tests with long inlet piping. Although the Crosby SM6 was not tested on short inlet piping, one drained loop seal test (1411) is applicable to the Davis-Besse inlet conditions with steam discharge. The ring settings in this test were (-77, -18) and are representative of typical PWR plant ring settings as shipped by the manufacturer. The 6M6 popped open at a pressure of 2420 psia with a pressurization rate of 300 psi /s, a backpressure of 245 psia, and a peak tank pressure of 2664 psia.
The test inlet fluid conditions for the steam discharge tests discussed above are representative of the expected conditions for the Davis-Besse FSAR transients resulting in steam discharge for the safety valve.
FSAR transients resulting in liquid discharge through the safety valve are discussed in Reference 20. The Licensee reviewed the Davis-Besse Updated Safety Analysis Report (USAR) and concluded that since the safety valve set point is 2525 psia and the maximum reactor coolent system pressure following the loss of normal feedwater (including a feedwater line break (FWLB)) would be 2512 psia, that no liquid discharge is expected as a result of a design basis accident. Reference 9 indicated the peak pressure in the FWLB was 2515 psia. The 2 psi difference is not significant.
Davis-Besse is a B&W FA-177 raised loop plant with low head high pressure injection (HPI) pumps (maximum pressure < 1900 psi). Because the Hp! pump cutoff head at Davis-Besse is below the safety valve and PORV
- setpoints, the extended HPI events cannot challenge the safety valves or PORV with liquid discharge. .
The analyses in Reference 9 used to determine the FSAR safety valve inlet conditions were performed assuming the PORV was inoperable. The PORV inlet conditions are assumed to be the same as those determined for the safety valves. This conservatively bounds the PORV inlet conditions.
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For FSAR transients resulting in steam discharge, the PORV will open at
, a pressure somewhat above the opening set point of 2465 psia. A maximum PORV pressure of 2677 psia is expected for Davis-Besse based on the rod ejection at hot zero power accident. The maximum predicted backpressure is l 1ess than 550 psia.
I The Crosby HPV-SN valve was subjected to 17 steam tests in the EPRI program. In the steam tests, the maximum pressure at valve opening ranged 3
from 2150 psia to 2505 psia with backpressure ranging from 60 to 560 psia.
Based on Reference 9, the bounding liquid inlet conditions representative of a feedwater and steam line break transient at. Davis-Besse are: ma'ximum pressure, 2515 psia, minimum liquid temperature, 4000F, and maximum liquid temperature, 6420F. To bound these conditions, the test PORV'was. subjected to one steam to water transition test and three water
. discharge tests. In the steam to water transition test the valve opened on 2510 psia. saturated steam and then passed water at 6490F. The backpressure was not available. In the three water tests, the valve opened at pressures ranging from 2502 to 2510 psia with temperatures ranging from 446 to 6340F and backpressure ranging from 155 t'o 315 psia.
1 The test inlet fluid conditions for PORV steam and liquid tests bound all of the Davis-Besse transient and accident conditions.
The PORV block valves are required to operate under the same fluid conditions as the PORVs. The Velan B10-3054B-13MS gate valve and Limitorque i SMB-000-10 and SB-00-15 actuators tested by EPRI were only subjected to full pressure steam tests. The EPRI PORV block valves were subjected to 21 steam l tests for each valve / actuator assembly. Steam pressures upstream of the block valves during the opening cycles varied from 2425 psia to 2515 psia and during the closing cycles varied between 2330 psia and 2425 psia. The f
full pressure steam tests are representative of the Davis-Besse FSAR inlet
. conditions based on the discussion below.
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-In the EPRI tests, the block valve was only tested for full pressure U
y (to 2500. psia) steam conditions. The operability of the block valves under water flow conditions was not directly addressed in the EPRI tests.
However, the Westinghouse' gate valve closing tests-(Reference 8) demonstrated that-the required torque to open or close the valve depended almost entirely on the differential pressure across the valve disk and was insensitive to the momentum load. Therefore, the required force is nearly independent of the type of flow (i.e., water or steam). Furthermore, IL
-according to friction tests done by Westinghouse'on a stellite coated specimen (the Velan valve has stellite coated disk and' seats), the friction u
coefficient between stellite surfaces is approximately the same for steam and water tests. .In some instances, the friction force in water media is lower than~in steam. Thus, the force required to overcome disk friction in steam is essentially equal to the force required in water, and the steam
. tests are adequate to demonstrate the operability of the block valves for
-the expected water conditions.
The presentation above demonstrates that the EPRI test conditions bounded the plant-specific conditions for the safety, PORV, and PORV block
-valves. Items 2 and 4 of Section 1.2 were met, in that conditions for the operational occurrences were determined and the highest predicted pressures were chosen'for the tests. The presentation also verifies that the portion !
of Item 7, which requires showing that the test conditions are aquivalent to those predicted in the FSAR, was met.
4.3 Operability As discussed in the previous section the safety valves are required to operate under full pressure steam conditions, and the PORV is required to operate under full pressure steam, steam to water transition, and sub:ooled -
water conditions. Representative valves were tested under full pressure steam, transition, and water conditions in the EPRI test program. The PORV .
. block valves are required to operate under fluid conditions similar to the PORV. Representative PORV block valves were tested under full pressure steam conditions in the EpRI test program, the results of which also apply to liquid flow.
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Since the Davis-Besse Crosby 4M 1 6 safety valve was not tested by EPRI, operability will be demonstrated by utilizing test data for the Crosby 3K6 and 6M6 safety valves, one smaller and one larger than the 4My6 with typical plant ring settings.
Reference 10 provides the EPRI safety valve and PORV test results.
Full pressure steam tests 406, 408, 411, and 442, performed on the 3K6 valve, had typical plant ring settings which correspond to the typical plant ring settings of the Davis-Besse safety valves. During these tests the valve opened 0.4% to 2.3% below the design setpoint, performed stably with some flutter-on test 406, and closed with blowdowns of 10.1 to 10.9%. Rated flow at 3% accumulation was exceeded for tests 411 and 442, though the valve was in a lift position as low as 96% of rated lift. Flow data for tests 406 and 408 was not available.
One applicable steam test (1411) on the Crosby 6M6 was performed with typical PWR ring settings of (-77, -18). During this test the valve opened 3.2% below the design setpoint, performed stably, and closed with a blowdown of 8.2%. Rated flow was exceeded at 3% accumulation though the valve was at 92% of rated lift.
Steam tests on the 3K6 (411 and 442) and 6M6 (1411) safety valves demonstrated that different size valves with typical plant ring settings and !
inlet conditions behaved in a stable manner with blowdowns between 8.2 and 10.9%.
In addition to the EPRI tests, Davis-Besse contracted Crosby Va.ve and Gage Company through their consultant, Teledyne Engineering Services, to perform steam testing on the 4M 1 6 safety valve. The purpose of the
, testing was to evaluate the operability of the Crosby 4M16 safety valves with blowdowns of 15% or more. It should be noted that the Crosby test facility is not capable of simulating the Davis-Besse inlet conditions (steam at 2500 psia). However, the Davis-Besse supplied 4M1 6 test valve was tested with a spring of lower but proportional-loading to the Davis-Besse spring. This allowed valve testing at pressures of <1320 psia.
The safety valve was installed directly on the test tank at the Crosby 13
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facility, without inlet piping, simulating the Davis-Besse installation.
, Test results of the 4M16 safety valve with the prorated spring, backpressure of 65 to 95 psia, and ring settings corresponding to the in-plant ring settings, demonstrated blowdowns of -9.3% by linear interpolation. Therefore, the Davis-Besse expected blowdown with the ' '
factory recommended ring settings is -9.3% and is within the range of blowdowns measured in the EPRI tests. 4 4
The Crosby safety valves tested by EPRI did not perform well during 1 subcooled water discharge tests. While water discharge through the safety valves at Davis-Besse is not expected, the Licensee recognized the need for inspection and maintenance of the plant Crosby safety valves for continued
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reliable operation should they discharge water. Therefore, the Licensee stated in Reference 20 that the plant operating procedures are being modified to' require inspection and. maintenance following any reactor trip involving a safety valve actuation if, during Post Trip Review, it is ascertained that the valve may have discharged water.
-The two EPRI test valves 3K6 and 6M6 physically bound the 4M1 6 safety valve. As discussed above, the observed test blowdown results ranged from 8.0% to 13% which exceeds the design value of 5%. B&W performed an analysis to determine the maximum allowable blowdown on 177-FA plants (Reference 22). It was concluded that there were no adverse effects on plant safety with blowdowns up to a maximum of 20%. Although the increased blowdowns observed in the EPRI tests are higher than the design values for Davis-Besse, they are deemed satisfactory and acceptable for the Davis-Besse
. 4M1 6 safety valves.
Bending moments ranging from 114,000 to 133,000 in-lb were induced on the discharge flange of the 3K6, and a bending moment of 239,000 in-lb was '
induced on the 6M6 discharge flange during the applicable tests. In all cases, safety valve performance was unaffected. The bending moment on the l Davis-Besse safety valve nozzle to inlet flange weld was calculated and I included the effects of deadweight, seismic loads, and the effects of loads due to the ruptures discs in the valve discharge tee not bursting simultaneously. One disc was assumed to rupture at 75 psi and the other at 14
l 150 psi. The resulting moment on the weld was 94,911 in-lb. This would bound the bending moment on the safety valve. Thus the bending moments applied to the test valves bound the bending moment for the plant valve.
Valve stability at Davis-Besse is demonstrated when the plant-specific iniet piping pressure drop is less than the piping pressure drup of a similar valve in the EPRI test program. Since the Davis-Besse safety valves are mounted directly on the pressurizer nozzles, the inlet piping was minimized, and thus, no inlet piping pressure drop calculations or comparisons to EPRI data were made. The Davis-Besse safety valves should perform in a stable manner similar to that observed in the applicable steam tests.
During each of the steam tests, the Crosby HPV-SN 1-3/8 in. bore PORV opened on demand, with total actuation times ranging from 0.29 to 0.37 s, passed 151,200 lbm/h to 168,000 lbm/h steam and closed on demand, with actuation times ranging from 0.14 to 0.16 s. The rated flow capacity of the EPRI test PORV was 120,000 lbm/h of saturated steam at 2300 psig and 3%
overpressure. Considering the revised set pressure of 2450 psig, the corre'ted c rated flow capacity is 127,250 lbm/h. The test PORV flow capacity was observed to be as high as 132% of rated capacity.
One steam to water transition test and three water tests were performed simulating the feedwater line break and extended high pressure injection events. During the transition test, the PORV opened on demand in 0.16 s, passed 316,800 lbm/h of water after the steam / water transition occurred, and closed on demant in 0.25 s. For the three water tests, the PORV opened on demand, with actuation times r'anging from 0.05 to 0.1 s, passed water flow rates ranging from 385,200 to 792,000 lbm/h and closed on demand, with actuation times of 0.10 to 0.20 s. These flows are in excess of the 198,000 lbm/h insurge rate predicted to occur with inlet coolant temperatures of 4000F, and the 385,200 lbm/h flow rate is close to the maximum insurge rate (420,300 lbm/h) predicted to occur with inlet coolant temperatures of 6400F.
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- A bending moment of 31,600 in-lb was induced on the discharge flange of the PORV during one of the EPRI ', team tests. PORV performance was unaffected by the application of this bending moment. The largest predicted l bending moment on the Davis-besse PORV was calculated to be 13,810 in-lb.
Therefore, the bending loads expected on the Davis-Besse PORV will not '
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4 As discussed in Reference 10, pre-evaluation tests on the Crosby PORV e
at Marshall indicated failure to open on demand for several actuations and failure to seat properly. Disassembly and inspection revealed a fractured f
bellows weld and improperly machined bellows flange. A second Crosby PORV j used in.the Wyle Phase II and III testing was disassembled and inspected prior to testing. Inspection revealed an improperly machined bellows flange. Replacement of the damaged bellows and remachining of the bellows L
flange restored the PORV to operation and operability was demonstrated by successful EPRI tests.
Reference 15 Response 5 outlined the maintenance and test history of
.the Davis-Besse PORV. On one occasion the PORV failed open during a transient requiring its operation. Disassembly and inspection of the PORV revealed a stuck open pilot valve. Upon replacement of the pilot valve stem and repair, the valve was cycled. On the sixth cycle the PORV failed open.
Disassembly and inspection indicated a correction to the pilot stem to nozzle guide clearance was required. After machining and reassembly the valve was successfully tested several times. On June 9, 1985, the PORV failed to resent on the third lift during a transient requiring its operation. The PORV was isolated by closing the upstream PORV block valve. l The PORV closed -2 min later on its own. Cause of the PORV failure is not -
known, and according to the NRC review of the transient (Reference 21), the cause may never be known. Toledo Edison stated that if the cause of the
- PORV sticking open is not determined, a new valve may be procured and tested prior to installation (Reference 21). -
As discussed above, the PORV testing demonstrated operability of the I Davis-Besse PORV under the expected inlet fluid conditions provided the i 1 1
valve is in good working order and all parts properly machined. It snould i
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l .. i be noted that the EPRI tests did not verify PORV reliability, that is, the- 1 L'e capability of the PORV to operate innediately after a preceding open and close cycle. This would occur.as a result of repeated cycling during a transient as observed in the Davis-Besse incident on June 9, 1985.
i NUREG-0737 II.D.1 required qualification of the associated control i circuitry'as part of the safety and relief valve qualification task. i Meeting the licensing requirements of 10 CFR 50.49 for this electrical equipment is considered satisfactory and specific testing per the NUREG-0737 requirement is not required. In Reference 19, the Licensee stated that there is no design-basis accident listed in Chapter 15 of the Davis-Besse {
Updated Final Safety Analysis Report for which the PORV is required.
Therefore, the PORV control circuitry was not environmentally qualified underIdCFR50.49.
f 1
However, the fact that plant emergency proceduras do not specifically prohibit PORV use during accident conditions such as those listed in Regulatory Guide 1.70, Rev 2,'was demonstrated by the June 9, 1985 loss of main and emergency feedwater event, during which operation of the PORV
-contributed to plant recovery. While the PORV control ;ircuitry is not qualified, it is noted that the PORV block valve (Reference 19) and the flow monitor downstream of the PORV (Reference 23) are environmentally qualified under 10 CFR 50.49. This configuration ensures the capability to detect and
. isolate a stuck open PORV. Therefore, it is concluded the Davis-Besse PORV control circuitry meets the qualification requirements of NUREG-0737, Item II.D.I. ,
The PORV block valve must be capable of closing over a range of steam and water conditions. As described in Section 4.2, high pressure steam
. tests are adequate to bound operation over the full range of inlet i
conditions, and as described in Section 4.1, the tests with the 3 in. Velan valve and SMB-000-10 operator conservatively demonstrate the operability of
'the plant valve. The test valve was cycled successfully at full steam pressure with full flow. It was shown to open and close successfully with torques as low as 82 ft-lb (Reference 9). The plant valve operator is set to produce a torque of 94.7 ft-lb (Reference 19), and therefore, the tests are considered to adequately demonstrate acceptable valve operation.
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1-The presentation above, demonstrating that the safety valves, PORV, and block valves operated satisfactorily, verifies that the portion of Item 1 of Section 1.2.that requires conducting tests to qualify these valves, and that part of Item 7 of Section 1.2 requiring that the effect of discharge piping on operability be considered were met. Item 5 of-Section 1.2 requiring .
qualification of the PORV control circuitry was met. J l
4.4 piping and Support Evaluation In the piping and support evaluation, the safety valve piping between the pressurizer nozzles and valves and the PORV piping between the pressurizer nozzle and the pressurizer relief quench tank were analyzed.
The requ'irements of the ASME Boiler and Pressure Vessel Code,Section III, 1971 edition with addenda to Sumer 1973 were used as the governing code for the Class 1 and Class 3 piping. The basic allowables listed in the AISC Code (7th Edition) were ased to define limits for the piping supports. The Licensee stated that no multiplying factors as allowed for seismic loads were applied to these lirr.its. Lead combinations esed in the analyses were consistent with those in the EPRI guidelines. The piping was analyzed f'or thermal expansion, pressure, weight, earthquake, plant operational thermal and pressure transients, and safety valve and PORV discharges.
Because the safety valves were moved to the pressurizer nozzles and the discharge piping removed, the thermal-hydraulic analysis for Davis-Besse only considered the PORV piping. The transient conditions analyzed were based.on Reference 9 and included discharge of saturated steam with and without hot loop seals, and subcooled water at 4000F at the assumed PORV opening pressure of 2465 psia.' For the saturated steam analyses a pressurization rate of 175 psi /s was assumed in ramping the pressurizer pressure from the PORV set point of 2465 psia to the maximum pressure of 2677 psia. The forces generated from these conditions bound those from all other conditions expected at the plant.
The thermal-hydraulic analysis was performed with the program RELAPS-FORCE. RELAPS-FORCE is a University Computing Company (UCC), Dallas, Texas-version of RELAP5/ MOD 1, Cycle 14, which was modified to include the 18
_ _ _ _ _ . _ _ _ _ _ . _ , _ . _ _ - - - - - - - - - - - - - - - " - ^ ' ~'"
.6
- , l capability to computc the hydraulic force on each pipe segment.
. Verification showed RELAPS-FORCE preserves the results from RELAP5.
Furthermore, the ability of RELAPS-FORCE to calculate pipe segment forces was verified through simulations of EPRI/CE SRV tests, and Edward's and Hanson's blowdown experiments. Repeatability of the RELAP5 results with RELAP5-FORCE indicates the basic sode was preserved and the RELAP5 verification work in Reference ;4 applies to RELAPS-FORCE to show it a suitable tool for calculation of valve discharge transients. Comparison of the RELAPS-FORCE calculated pipe segmcnt forces to EPRI/CE data and the other experiments presented in Reference 25 shows good agreement, verifying the capability of RELAPS-FORCE to calculate pipe forces due to valve discharge.
A RELAPS model for the PORV piping from the valve discharge to the
. quench tank was developed. The safety valve was not considered since no discharge piping to the quench tank is used. In the piping model, the key parameters of node size, time step size, choked flow locations, and valve opening times were. reviewed. Choking was used only at the junction representing the PORV. This is appropriate for this type of analysis. The PORV opening time was modeled as 50 ms and is representative of the opening time measured by EPRI. The maximum time step of 2.0 x 10-4 s was sufficiently small to accurately calculate the piping forces for the node size used in the RELAPS model. Nodes sizes were approximately 0.75 ft long. To account for uncertainties in valve flow rates, the valve flow area and, therefore, the flow rate in the piping analysis was conservatively adjusted. A conservative factor of 1.35 was included in the maximum rated valve mass flow rate for the PORV. The conservative flow rate used in the analysis acceptably accounts fdr 10% ASME derating and potential error in l
the flow rate.
The thermal-hydraulic analysis is considered adequate for predicting the PORV discharge loads.
]
The structural analyt,is for the safety valve and PORV piping was performed ud ag the Teledyne Engineering Services (TES) computer code TMRPIPE. TMRPIPE is an integral system of computer programs for the 19 L ___ _ _ _ - - -
- y. ,
/
4 complete linear elastic analysis and evaluation of nuclear power piping i systems to the requirements of Section III of the ASME Code. This system of codes features independent subsets which can function as independent codes.
The integral package includes capabilities for temperature distribution l analysis, static and dynamic piping analysis, tabulation of output for inclusion in reports, and generation of model plots. TMRSAP is the subset of TMRpIPE used in piping analysis. TMRSAP is a modified version of SAPIV, I a stfuctural analysis program for linear elastic systems. SAPIV was prepared by the Earthquake Engineering Research Center of the University of California and widely used for piping system response analysis and other structural applications. The TMRSAp computer code was verified by the Teledyne Engineering Services and the verification results were presented in a Teledyne report which is part of the February 24, 1987 submittal (Reference 20). The portion of the report relating to piping discharge analysis was reviewed and the program was found to be acceptable for the piping dynamics application. Also, TMRSAP was included in an NRC Vendor Inspection Branch inspection of TES conducted in 1985. This inspection found TMRSAP to verified in an acceptable manner.
For the analysis of the valve discharge conditions, the input forcing functions were obtained from the thermal-hydraulic analysis using the RELAp5-FORCE program. The force time histories representing an unbalanced fluid force in each pipe segment were applied to the structural model at appropriate node points. The modeling technique and the key input parameters used in the analysis, such as lumped mass spacing, structural damping value, cutoff frequencies, etc., were reviewed and considered to be acceptable.
The piping analysis was performed in accordance with the requirements of Section !!! of the 1971 Edition of the ASME Code. The piping support analysis was performed using stress limits defined by the 7th Edition of the AISC Code. The load combination equations and stress limits used for the
- evaluation of the piping stresses upstream and downstream of the PORV for ,
Davis-Besse are consistent with the EPRI guidelines.
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, The piping stresses calculated by Teledyne Engineering Services, a consultant of the Licensee, were reviewed. The piping stresses evaluated on the basis of the load combinations defined above are found to be acceptable.
The safety valves at Davis-Besse are mounted directly on the pressurizer nozzles. Toledo Edison evaluated the stresses on the safety valve nozzle and the weld attaching the valve flange to the nozzle. The stress at the valve flange was 21390 psi compared to an allowable of 25050 psi. The safety valve nozzle stresses were determined to be negligible, with the largest stress being 2306 psi. Nozzle loads (bending moments) were also evaluated and found to be within acceptable limits.
The results of the PORV piping support analysis identified three piping supports with loads exceeding the design loads. Davis-Besse modified the piping supports to meet the new loading conditions and ASME code requirements (Reference 15 Response 12).
The selection of a bounding case for the piping evaluation and the piping and support stress analysis demonstrate that the requirements of Items 3 and 8 of Section 1.2 of this report were met, j
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- 5. EVALUATION SutHARY a
The Licensee for. Davis-Besse provided an acceptable response to the requirements of NUREG-737, thereby reconfirming the General Design )
Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met with regard to the safety valves and PORVs. The rationale for this conclusion is given f
j below. I The Licensee developed an acceptable relief and safety valve test program to qualify the operability of the prototypical valves and to demonstrate that their operation would not invalidate the integrity of the associated equipment and piping. The subsequent tests were successfully completed under operating conditions which by analysis bound the most probable maximum forces expected from anticipated design basis events. The i test results showed that the valves tested functioned correctly and safely for all steam and water discharge events specified in the test program that are applicable to Davis-Besse. The pressure boundary component design criteria were not exceeded. Analysis and review of both the test results and the Licensee justifications indicated the direct applicability of prototypical valves and piping performance to the in-plant valves and piping intended to be covered by the test program. The plant-specific piping also was shown by analysis to be acceptable.
In addition, while water discharge through the safety valves is not 1
expected at Davis-Besse, the Licensee recognized the need for inspection and maintenance of the plant Crosby safety valves should they discharge water.
Therefore, the Licensee committed to implementing procedures for inspection and maintenance of the valve if, during Post Trip Review, it is ascertained I the valve may have discharged water.
Thus, the requirements of Item II.D.1 of NUREG-0737 (Items 1-8 of Section 1.2) were met thereby ensuring that the reactor primary coolant -
pressure boundary will have a low probability of abnormai leakage (General Design Criteria No. 14). In addition, the reactor primary coolant pressure l
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'.e
..- boundary and its associated mechanical components were designed with
s
~ sufficient margin so that design conditions are not exceeded during o
relief / safety valve events (General Design Criteria No. 15). Further, the prototypical teste and the successful performance of the valves and associated mechanical components demonstrate that this equipment was constructed in accordance with high quality standards, meeting General
-Design Criteria No. 30.
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I e 6. REFERENCES s'- 1. TMI-Lessons Learned Task Force Status Report and Short Term Recommendations, NUREG-0578, July 1979.
- 2. Clarification of TMI Action Plan Requirements, NUREG-0737, November 1980.
- 3. Letter, D. P. Hoffman, Consumers Power Co. to H. Denton, NRC, Transmittal of PWR Safety and Relief Valve Test Program Reports, September 30, 1982.
- 5. EPRI PWR Safety and Relief Valve Test Program Valve Selection / Justification Report, EPRI-NP-2292, December 1982.
- 6. EPR'1 PWR Safety and Relief Valve Test Program Test Condition Justification Report, EPRI NP-2460, December 1982.
- 7. EPRI/ Marshall Electric Motor Operated Block Valve, EPRI NP-2514-LD, July 1982.
B. EPRI Summary Report: Westinghouse Gate Valve Closure Testing Program, Engineering Memorandum 5683, Revision 1, March 31, 1982.
- 9. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves for B&W 177-FA and 205-FA Plants, EPRI NP-2352, December 1982.
- 10. EPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Report, EPRI NP-2628-SR, December 1982.
- 11. J. F. Stolz, NRC, from R. P. Crouse, Toledo Edison, SN-834, July 1, 1982.
- 12. J. F. Stolz, NRC, from R. P. Crouse, Toledo Edison, SN-886, December 17, 1982.
- 13. J. F. Stolz, NRC, from R. P. Crouse, Toledo Edison, SN-905, February 2, 1983.
- 14. R. P. Crouse, Toledo Edison, from J. F. Stolz, NRC, June 7, 1985.
- 15. J. F. Stolz, NRC, from J. Williams Jr., Toledo Edison, Serial No. 1171, July 19, 1985.
- 16. J. F. Stolz, NRC, from J. Williams, Jr., Toledo Edison, Serial .
No. 1191, October 1, 1985.
- 37. J. F. Stolz, NRC, from J. Williams, Jr., Toledo Edison, Serial i No. 1221, December 6, 1985.
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. 18. J. Williams, Jr., Toledo Edison, from J. F. Stolz, NRC, July 31, 1986.
s I
. 19. J. F. Stolz, NRC, from J. Williams, Jr., Toledo Edison, Serial !
No. 1308, October 14, 1986.
- 20. J. F. Stolz, NRC, from J. Williams, Jr., Toledo Edison, Serial ;
No. 1338, February 24, 1987. l 1
- 21. Loss of Main and Auxiliary Feedwater Event at the Davis-Besse Plant on June 9, 1985, NUREG-1154.
- 22. Pressurizer Safety Valve Maximum Allowable Blowdown, B&W Report 77-113-5671-00, August 1982.
- 23. -J. F. Stolz, NRC, from R. P. Crouse, Toledo Edison, " Davis-Besse Equipment Qualification Manual Rev.2," Serial No. 1009, November 29, 1983.
24 Application of RELAPS MODI for Calculation of Safety and Relief Valve Discharoe Piping Hydrodynamic Leads, EPRI-2479, December 1982.
- 25. Letter J. E. Ward, SMUD, to J. F. Stolz, NRC, " Rancho Seco Nuclear Generating Station, Unit 1, NUREG-0737, Item II.D.1 Request for Information," March 3, 1987.
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