ML20215L711
| ML20215L711 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 05/31/1987 |
| From: | Fineman C, Nalezny C EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20215L697 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-7700, NUDOCS 8706260153 | |
| Download: ML20215L711 (30) | |
Text
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4 INFORMAL REPORT TECHNICAL EVALUATION REPORT
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;t TMI ACTION--NUREG-0737 (II.D.1) i
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RELIEF AND SAFETY VALVE TESTING National.
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EGG-h'TA-7700 TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING RANCHO SECO UNIT 1 DOCKET NO. 50-312 C. P. Fineman C. L. Nalezny I
i May 1987 Idaho National Enginedring Laboratory EG&G Idaho, Inc.
e Prepared for the U.S. Nuclear Ragulatory Commission Washington D.C. 20555 Under DOE Contract No. DE-AC07-761001570 FIN No. A6492
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i ABSTRACT Light water reactors have experier.ced a number of occurrences of improper performance of safety and relief valves installed in the primary
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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 and the responses of the licensee of Rancho Seco Unit 1 to the requirements of NUREG-0578 and NUREG-0737 by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc.
This review has found the Licensee has provided acceptable responses, which reconfirm that the General Design Criteria 14,15, and 30 of Appendix A to 10 CFR 50 have been met.
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FIN No. A6492--Evaluation of OR Licensing Actions-NUREG-0737, II.D.1 ii
CONTENTS ABSTRACT..............................................................
ii 1.
INTRODUCTION.....................................................
1 1.1 Background.................................................
1 1.2 General Design Criteria and NUREG Requirements.............
1 i
2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM................
4 3.
PLANT SPECIFIC SUBMITTAL.........................................
6 4.
REVIEW AND EVALUATION............................................
7 4.1 Valves Tested..............................................
7 4.2 Test Conditions............................................
8 4.3 Determination of Safety Valve Ring Settings at Rancho Seco................................................
1?
4.4 Operability................................................
13 4.5 Piping and Support Evaluation..............................
18 5.
EVALUATION
SUMMARY
22 6.
REFERENCES.......................................................
24 i
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I TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1) RELIEF AND SAFETY VALVE TESTING RANCHO SECO UNIT 1 DOCKET NO. 50-312 1.
INTRODUCTION 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 have been instances of valves opening below set pressure, valves opening above set pressure, and valves failing to open or reseat.
From these past instances of improper valve performance, it is not known whether they occurred because of a limited qualification of the valve or becausef basic unreliability of the valve design.
It is known that the o
failure of a power operated relief valve (PORV) to reseat 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 raexamine 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 I
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 extremely low probability of abnormal leakage, (2) the reactor coolant system and associated auxiliary, control, and protection systems be 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.
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To reconfirm the integrity of overpressure protection systems and thereby assure that the General Design Criteria are met, the NUR~.G-0578 position was issued as a requirement in a letter dated September 13, 1979, by the Division of Licensing (DL), Office of Nuclear Reactor Regulation (NRR), to ALL OPERATING NUCLEAR POWER PLANTS.
This requirement has since been incorporated as Item II.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements, 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 operating 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 pressure predicted by conventional safety analysis procedures.
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 (NRC) staff review and evaluation, including criteria for success or failure of valves tested.
7.
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 4
2 r
2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PR00 GRAM In response to the NUREG requirements previously listed, saggrou_af utilities with PWRs requested the assistance of the Electric PPower R :aarch Institute (EPRI) in developing and implementing a generic testttprogr:: for pressurizer safety valves, power operated relief valves, blockkevalve:, and associated piping systems.
Sacramento Municipal Utility Distrrict (Si JD),
the owner of Rancho Seco Unit 1, was one of the utilities spornsoring :he
~
EPRI Valve Test Program.
The results of the program, which arre tcont ined in j
a series of reports, were transmitted to the NRC by Reference n3.
The
)
applicability of these reports is discussed below.
EPRI developed a plan (Reference 4) for testing PWR safetty,. relief, and block valves under conditions which bound actual plant operatiing conditions.
EPRI, through the valve manufacturers, identifiecd.tne valves used in the overpressure protection systems of the participatiing utilities and representative valves were selected for testing. These vtalves included a sufficient number of the variable characteristics so that tiheir testing would adequately demonstrate the performance of the valves usted 'by utilities (Reference 5).
EPRI, through the Nuclear Steam Supply Systemi.(NSSS) vendora, evaluated the FSARs of the participating utilities amd arrived at a test matrix which bounded the plant transients for which over ; pressure i
protection would be required (Reference 6).
I EPRI contracted with Babcock & Wilcox (B&W) to produce at report on the i
inlet fluid conditions for pressurizer safety and relief valvees in B&W
~
designed plants (Reference 7).
Since Rancho Seco Unit 1 was cdesigned by B&W, this report is relevant to this evaluation.
Several test series were sponsored by EPRI.
PORVs and bliock valves were tested at the Duke Power Company Marshall Steam Station llocated in Terrell, North Carolina.
Additional PORV tests were conducted at the Wyle Laboratories Test Facility located in Norco, California.
Safety relief valves (SRVs) were tested at the Combustion Engineering Comparny,, Kressinger 4
Development Laboratory, which is located in Windsor, Connecticut. The results of the relief and safety valve tests are reported in Reference 8.
The results of the block valve tests are reported in Reference 9.
The primary objective of the EPRI/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 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.
Transmittal of the test results meets the requirements of Item 6 of Section 1.2 to provide test data to the NRC.
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3.
PLANT SPECIFIC SUBMITTAL A preliminary assessment of the adequacy of the overpressure protection system was submitted by SMUD on June 30, 1982 (Reference 10).
Additional information on the safety valves was submitted on July 29, 1983 (Reference 11). An initial assessment of the Pressurizer Safety and Relief Valve Piping was included in this transmittal. A request for additional information (Reference 12) was submitted to SMUD by the NRC on November 21, 1984.
SMUD responded to this request on January 21, 1985 (Reference 13),
and on April 12, 1985 (Reference 14).
Review of References 13 and 14 resulted in an additional request for information which was transmitted to the licensee on December 16, 1986 (Reference 15) to which SMUD responded on March 3, 1987 (Reference 16).
The response of the overpressure protecti'n system to Anticipated o
Transients Without Scram (ATWS) 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 Rancho Seco utilizes two safety valves, one PORV, and one block valve in the overpressure protection system.
Both safety valves are Dresser Model 31759A. The PORV is a Dresser Model 31533VX-30. The block valve is a 2-1/2 in. Velan F9-454B-13MS gate valve with a Limitorque SMB-00-10 actuator.
The Dresser 33759A safety valve used at Rancho Seco was not one of the valves tested by EPRI. The 31759A valve falls between the two valves tested, the 31739A and 31709NA, with respect to size.
It is closer in size to the 31739A valve, the smaller of the two valves tested. The 31759A valve differs from the test valves in the size of the inlet and outlet flanges and in the orifice size.
These differences do not affect valve operability.
These considerations, and the fact that all Dresser valves are similar in configuration and design philosophy, indicate the test valves are representative of the Rancho Seco valves.
The Dresser PORV installed at the Rancho Seco was originally a dash 1 (31533VX-30-1) design with a bore diameter of 1-5/32 in. The test valve was a dash 2 design with a bore size of 1-5/16. The dash 2 design resulted from a need to improve the seat tightness and included modifications to the internals, the body, and the inlet flange. The body and flange modifications were not of a nature that would affect operability. The Rancho Seco valve has been modified to incorporate the changes to the internals of the dash 2 design. The difference in bore diameter will only affect capacity and not operability. The test valve is, therefore, considered an adequate representation of the in-plant valve.
The Velan block valve used at Rancho Seco is a 2 1/2 in. gate valve Model Number F9-454B-13MS, and has a Limitorque SMB-00-10 operator.
Two Velan valves, both 3 in, gate valves, Model B10-3954-13MS, were tested by EPRI (Reference 9).
One was tested with a Limitorque operator SB-00-15 and the other tested with a Limitoroue operator SMB-000-10.
The plant and test 7
valves are of the same style, internal design, and operation. They differ in size, pressure rating, and valve ends, which have no affect on operability.
The 3 in. valve requires a larger force to operate and the SMB-000-10 operator is the smaller operator with the same starting torque as the plant valve, so the tests with this operator on a 3 in. valve are a conservative demonstration of the operability of the plant valve.
The block valve at Rancho Seco is installed in a vertical position while the EPRI tests were performed with the valve in the horizontal position.
Since the plant valve and the test valves are designed for use in either orientation, the horizontal tests are considered applicable to the vertical configuration.
Based on the above, the valves tested are considered to be applicable f
i to the in-plant valves at Rancho Seco 1 and to have fulfilled that part of the criteria of Items 1 and 7 as identified in Section 1.2 regarding applicability of test valves.
4.2 Test conditions The valve inlet fluid conditions that bound the overpressure transients for B&W designed PWR plants are identified in Reference 7.
The transients considered in this report include FSAR, extended high pressure injection (HPI), and low temperature overpressurization events.
Reference 7 addresses thase transients listed in Regulatory Guide 1.70, Rev. 2, which potentially challenge the PORV or safety valves in B&W plants.
The conditions in the report that are applicable to Rancho Seco are those identified for B&W 177-FA plants.
For the SRVs only steam discharge was calculated for FSAR type transients. The peak pressure was 2677 psia and the maximum pressurization rate was 175 psi /sec. According to Reference 11, a maximum backpressure of 570 psia is developed at the SRV outlet.
Since Rancho Seco does not have loop seals upstream of the SRVs, testing of the Dresser safety valves with the short inlet piping is applicable.
8
Six steam tests with a short inlet pipe were performed with the 31739A valve which had a peak pressure of 2703 psia and a peak pressurization rate of 303 psi /sec.
Tests with backpressures as high as 866 psia were run.
With the larger 31709NA valve, five steam tests were run with a peak pressure and pressurization rate of 2697 psia and 322 psi /sec, respectively. One test was run with a backpressure of 530 psia. These conditions bound those expected at Rancho Seco.
For extended HPI events, which include steam line and feedwater line break events, the safety valve will initially open on steam with transition to subcooled water calculated.
A peak pressure of 2515 psia was calculated 0
0 with liquid temperatures ranging from 400 F to 640 F, A peak liquid surge rate of 11,500 lbm/ min (at 6400F) will occur.
Pressurization rates from 0 to 65 psi /sec are expected.
For the 31739A valve, testing included r, steam to water transition test at 2489 psia and saturated conditions.
Three water tests at pressures ranging from 2389 to 2749 psia and with water temperatures of 4140F to 6080F were run.
During these test the 31739A valve passed at least 1128 GPM (-8,000 lbm/ min) with 5390F water, and 2492 GPM (-16,000 lbm/ min) 0 with 649 F water.
For the 31709NA valve, the test series included two steam to water transition tests at pressures of 2530 and P.545 psia and saturated conditions and four water tests with temperatures from 4150F to 6250F and pressures from 2393 to 2558 psia.
During three of these tests, the 31709NA valve passed at least 1646 GPM (-11,666 lbm/ min), during the fourth test, with 4150F water, the valve chattered and no data was obtained. The transition and water tests for both valves were run with pressurization rates from 1.8 to 3.2 psi /sec. Although these represent the lower end of the range of pressurization rates calculated for B&W plants, they are adequate to represent expected inlet conditions at Rancho Seco 1.
These conditions are sufficiently close to the conservatively selected bounding conditions to adequately demonstrate valve performance.
The PORV is used for low temperature overpressure protection at Rancho 1
Seco but during normal operation the PORV block valve is closed. The function of the PORV is limited to reactor system pressure control and the 9
1 l
PORV is not considered a safety device or part of the overpressure protection system (Reference 14). However, since the PORV is not valved out 100% of the time, and a transient may occur while the PORV is fulfilling its designed role of reactor system pressure control, the bounding inlet conditions which may occur at the plant were reviewed and compared to the test conditions.
Operability of the PORV under these conditions will be j
examined in Section 4.4.
For the PORV, FSAR events result only in steam discharge. Although Reference 7 indicated the PORV should be tested at a peak pressure higher than the opening set point, 2465 psia, the valve opens quickly enough that the increase in pressure during the opening cycle is minimal.
In addition, I
the peak pressure listed in Reference 7 was based on an analysis in which the PORV was assumed to be inoperable.
Testing with saturated steam at set pressure is, therefore, considered adeguate.
The Dresser PORV is a pilot operated valve and the back pressure developed at the outlet is of potential importance to valve operability. The ability of the valve to operate at backpressures at least as high as those expected in service should be demonstrated. The expected backpressure for the PORV was not reported by SMUD. The PORV discharge pipe routing is similar to the safety valves.
The PORV rated flow, 119,000 lb/hr, is 30% of the rated flow of the safety valve, 391,000 lbm/hr.
The 4 inch discharge pipe of the PORV has approximately 44% the flow area of the 6 inch pipe for the safety valves.
From these data the conclusion is reached that the expected backpressure for the PORV is less than the 570 psia reported for the safety valve. Testing of the valve (Reference 8) included numerous steam test with opening pressures close to the Rancho Seco set pressure and back pressures as high as 760 psia which adequately bounds the expected conditions for the PORV.
For extended HPI events the initial opening of the PORV will be c,a steam but subcooled liquid could follow.
HPI events can, therefore, result in steam to water transition and water (4000F to 6500F) discharge at a maximum pressure of 2500 psia (Reference 7). A steam to water transition test and liquid tests with temperatures ranging from 4470F to 6470F and 10
i I
pressures of approximately 2500 psia were included in the test series. The tests were run usirG the same discharge pipe orifice which developed backpressures as high as 450 to 500 psia for the steam tests so that the expected backpressure was adequately represented. The Hpl events are, therefore, considered to have been adequately represented by the tests.
The PORV is used for low temperature overpressure protection.
For low temperature overpressurization events the valve is required to operate over a range of inlet conditions. These include opening on 565 psia steam with a possible transition to saturated water and opening on subcooled water with temperatures ranging from 3380F to 4490F (Reference 7). Opening on l
steam with possible transition to water is considered to be adequately represented by the full pressure, 2496 psia, steam to water transition test and the subcooled water conditions are considered to be adequately represented by tests with pressures from 680 to 699 psia and water temperatures of 1160F to 4560F.
l I
For the block valve only full pressure steam, 2480 psia, tests were l
performed (Reference 9). The block valve, however, is required to open and close over a range of steam and water conditions. The required torque to open or close the valve depends almost entirely on the differential pressure across the valve disk and is rather insensitive to the momentum loading and, therefore, is nearly the same for water or steam and nearly independent of the flow. The full pressure steam tests, therefore, are adequate to demonstrate operability of the valve for low pressure steam and the required water conditions.
l The test sequences and analyses described above, demonstrating that the j
test conditions bounded the conditions for the plant valves, verify that i
Items 2 and 4 of Section 1.2 have been met, in that conditions for the operational occurrences have been determined and the highest predicted.
pressures were chosen for the test. The part of Item 7, which requires showing that the test conditions are equivalent to conditions prescribed in the FSAR, is also met.
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4.3 Determination of Safety Valve Ring Settings at Rancho Seco Determination of the applicability of a particular EPRI test to a plant submittal usually included a comparison of the plant valve ring settings to the ring settings used in the test.
This comparison was needed to assure the plant valve will respond in a manner similar to that observed in the test. This comparison was not made in Section 4.2 because of the different 1
approach used by SMUD in the Rancho Seco submittal.
For Rancho Seco the setting for the middle ring, which controls most of the important valve performance parameters in Dresser valves, was determined by use of the COUPLE code (References 17 and 18).
COUFLE was developed during the EPRI tests by Continuum Dynamics, Inc. (CDI) to model the dynamics of spring loaded safety valves like the Dresser valves.
In Reference 18, COUPLE was shown to reliably predict the valve performance observed during the EPRI i
tests for both Dresser valves tested as a function of the middle ring setting.
The ability of the COUPLE code to determine a middle ring setting which results in acceptable valve performance has also been independently assessed. The GPU Nuclear submittal for Three Mile is.iand, Unit 1 (TMI-1) j used COUPLE to determine the middle ring setting for the Dressor 31739A safety valves used at TMI-1. The Dresser 31739A valve was one of the valves tested by EPRI.
Review of the EPRI test results showed that for tests where the middle ring setting bounded that determined by COUPLE for the TMI-1 valves, the 31739A valve had stable performance, passed rated flow, and closed with a reasonable blowdown under conditions similar to those expected at TMI-1 (Reference 19).
Therefore, it was concluded the COUPLE code can be used to determine a middle ring setting for Dresser valves which will result in acceptable valve performance.
The ring settings selected for the 31759A valves at Rancho Seco were:
upper = -48, middle = -93, and lower = +8.
The rationale for these settings was discussed in References 11 and 14. The upper setting was the same as that used in all the EPRI tests.
EPRI testing showed the upper ring in the Dresser design did not have a significant influence on valve performance.
The lower ring setting is set to give the quick popping 12
action of the valve as it opens.
This ring is usually set when the valve is cold.
Setting of the lower ring at +8 when the valve is cold will ensure the lower ring does not expand on heating and contact the disc holder forcing the valve'open.
It will also put the lower ring a nominal eight' notches above the seat.
Dresser feels this is adequate for correct operation (Reference 20).
Also, EPRI tests used lower ring settings from
-13 to +11 which bound the Rancho Secc setting.
The middle ring setting of
-93 was selected based on the most negative middle ring position actually tested with the 31739A valve during EPRI tests.
This value was -80 notches from one test, number 1008.
The -93 setting for the 31759A valve corresponds to the same scal _ed geometric position as -80 for the 31739A valve. As will be shown in Section 4.4, this ring setting resulted in acceptable valve performance as determined by COUPLE.
4.4 Valve Operability As discussed above, the middle ring setting for the Dresser 31759A valves used at Rancho Seco was determined using the COUPLE code. Use of COUPLE to determine valve ring settings and consequent valve performance is considered an acceptable approach for the reasons discussed'above, COUPLE analyses of the Rancho Seco valve for steam discharge (Reference 18) showed the valve was stable, achieved 85.4% of rated lift,.
and passed 100% of rated flow. Valve blowdown was calculated to be 8.6%.
The COUPLE analysis indicated the system pressure had to reach 7%
accumulation before the lift and flow results noted above could be achieved because of the high backpressure at the plant.
Error bounds on the COUPLE results reported in Reference 18 were blowdown, -1.9%, +2.1%, overprediction of stem lift by 2.4%, and underprediction of mass flow rate by 13.9% for the EPRI test data.
Plant valve performance within the error bounds of the' i
COUPLE analysis is considered adequate.
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The Rancho Seco submittal identified one EPRI test as applicable to the plant valves to support the assertion that acceptable valve performance will result with the COUPLE determined ring setting. This was test 1008 run with the smaller 31739A.
The middle ring setting in test 1008 was the one selected to set the middle ring in the Rancho Seco valve. This test was a steam discharge test with a peak pressure of 2680 psia, 275 psi /sec pressurization rate, and 617 psia backpressure. This test was run with the long inlet pipe used to characterize loop-seal plants, but this configuration conservatively bounds the Rancho Seco SRV installation without loop seals.
In this test the 31739A valve opened within +3% of the set pressure, reached 80% of rated lift, and passed 111% of rated flow at 3%
accumulation. The peak backpressure during this test was 617 psia which exceeded the maximum predicted backpressure of 570 psia for Rancho Seco.
The valve closed with 14.2% blowdown.
During the test the valve had stable performance.
It should also be noted that during test 618 with the larger 31709NA valve, 123% of rated flow and 101% lift were achieved at 3%
accumulation with a peak backpressure of 530 psia. This performance is considered acceptable and provides additional support to the validity of the COUPLE analysis, and shows that the Rancho Seco 31759A valve will achieve at least rated flow at 3% accumulation.
Additional support for acceptable valve performance during steam discharge is found in Reference 19. This reference reviewed EPRI test results of the Dresser 31739A valve to determine if the test valve gave acceptable results with ring settings similar to those used with the 31739A valves at the TMI-1.
As noted in Section 4,3, the ring settings at TMI-1 were also determined with COUPLE. This review is relevant because Rancho Seco and TMI-1 are both B&W 177-FA plants and the SRV inlet conditions indentified in Reference 7 are the same for both plants.
Reference 20 found the test valve gave acceptable performance during steam discharge tests with middle ring settings that bounded those determined by COUPLE for the plant valve. Therefore, it can be concluded acceptable performance for the nontested 31759A valve at Rancho Seco will also result with the COUPLE determined ring setting.
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i As noted in Section 4.2, possible inlet conditions for the safety valves include steam to water transition and water flow.
Direct eviderce was not provided in the submittal to demonstrate acceptable valve performance with the plant ring settings and transition and water inlet conditicns.
Indirect evidence is available from several sources, that indicates the 31759A valves at Rancho Seco will give acceptable performance under these inlet conditions.
First, as 70ted in Reference 13, unstable behavior was uncommon with the Dresser valves tested.
Only four tests out of 47 were identified as having chatter.
For the 31739A valve, the two unstable tests were on the long inlet pipe configuration.
Therefore, these j
tests are not applicable to Rancho Seco.
For the 31709NA valve, one test was run in a loop seal configuration the other was a 4290F water test.
The 4290F water test is applicable to Rancho Seco. The test was run with the short inlet configuration and the water temperature was representative j
of a steam line break at Rancho Seco 1.
During the test the valve opened and chattered for -3 s before stabilizing, without manual actuation, and then closed. After the test the valve was disassembled and inspected.
Galled guiding surfaces were found and several damaged internal parts were found.
Since the test is applicable to Rancho Seco and damaged parts were found, if the safety valves discharge water with a temperature less then 5500F, SMUD has agreed to inspect the valves to assure valve operability has not been compromised (Reference 13}.
Additional support is found in the supplement to the TMI-1 TER noted above (Reference 19).
Reference 18 found that in EPRI transition and water tests with ring settings bounding those at TMI-1, the 31739A valve gave acceptable performance.
The transition and subcooled water inlet conditions tested were similar to those expected at Rancho Seco. Therefore it is concluded that the 31759A valve at Rancho Seco, with a middle ring setting determined with COUPLE, will have acceptable valve behavior with transition and subcooled water inlet conditions.
Blowdowns for the Dresser safety valves tested by EPRI ranged from 4.7%
to 20.2% so that the measured blowdown generally exceeded the design blowdown of 5%.
A B&W analysis (Reference 21) has shown blowdown up to 20%
15
1 does not impede natural circulation due to hot leg voiding.
Therefore, having the observed blowdown exceed the ^ sign blowdown is considered acceptable.
The maximum bending moments applied to the discharge flanges of the test valves were 241,730 in.-lbs (31739A) and 473,000 in.-lbs (31709NA).
Valve operability was not impaired by the application of these moments. The maximum expected moment at the plant for the 31759A valve is 13,212 in.-lbs.
Application of a larger moment in the EPRI tests to the smaller 31739A valve indicates the tests bound the expected plant condition.
\\
For the test performance to be a valid demonstration of plant safety valve stability, the test inlet piping must have a pressure difference at 1 east as great as the plant. The plant valves are ' mounted directly on a pressurizer nozzle and thus have the minimum pressure drop possible. The test piping included a venturi and a reducing flange and, therefore, had a higher pressure difference.
During the 4140F water test the 31739A valve was stable but only achieved partial lift.
The valve did not pass enough flow to prevent the test pressure from accumulating.
SMUD in its submittal pointed out that the flow the 31739A valve passed (which is smaller than the 31759A valves at 1
Rancho Seco) was 30% more than required for the steam line break (which results in the 4000F liquid flow case).
In addition there are two safety valves at the plant, which gives Rancho Seco 1 more than. sufficient relief capacity. Under conditions typical of the FWLB, 2515 psia, 6020F and 6400F, the 31739A test valve, on the short inlet configuraion passed
-20% less (6020F) and -7% less (6400F) than the required flow.
Rancho Seco 1 does have sufficient relief capacity at these conditions, however, because the plant valve is larger than the test valve, the test-flow was measured at a pressure less than the 2515 psia used to determine the maximum surge rate in the analysis, and, as noted above, two valves are installed at the plant.
Based on the code and test results discussed above, demonstration of safety valve operability it considered satisfactory.
i 16
The Dresser PORV opened and closed on demand for all nonloop seal tests.
Inspection of the valve afte-testing at the Marshall Steam Station showed the bellows had several welds partially fail. The failure did not affect valve performance and the manufacturer concluded the failure did not have a potential impact on valve performance.
The bellows was replaced and did not fail during any of the additional test series.
A bending moment of 25,500 in.-lb was induced on the discharge flange of the test valve without impairing operability. The maximum bending moment calculated for the Rancho Seco 1 PORV is less than 18,000 in.-lbs.
The EPRI tests, therefore, bound the expected plant condition.
The Rancho Seco 1 PORV is a pilot operated valve that uses system pressure to hold the disk tight against the seat. At one point Dresser Industries recommended the block valve be closed at system pressures below 1000 psig to avoid steam wirecutting of the PORV disk and seat. A later recommendation by Dresser was to install heavier springs under the main and pilot disks to ensure closure at low pressures.
Rancho Seco 1 has modified-its PORV to include the heavier springs and, therefore, the PORV will be available for low temperature overpressure protection.
Based on the valve performance during EPRI tests, under the full range of expected inlet conditions, and based'on the modification using the heavier springs, the demonstration of relief valve operability is considered adequate.
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
~
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-lbs (Reference 9).
Since Rancho Seco uses a
-1 17-
Limitorque SMB-000-10 operator which was shown by test to produce adequate torgue, the block valve and operator are considered qualified.
In Reference 16, SMUD stated that the torque setting of the plant Limitorque SMO-000-10 operators have been set to produce a torque of 87.6 ft-lbf. which is sufficient to close the PORV " lock valves under expected operating conditions.
NUREG-0737 II.D.1 requires qualification of associated control circuitry as part of the safety / relief valve qualification. The Nuclear Regulatory Commission staff has agreed, however, that meeting the licensing requirements of 10 CFR 50.49 for this electrical equipment is satisfactory and that specific testing per the NUREG-0737 requirement is not required.
In Reference 16, SMUD stated that The PORVs are not required to perform any safety function to mitigate the effects of any design basis accident when exposed to a harsh environment. The licensee stated that as part of the 10 CFR 50.49 environmental qualification program, the PORV solenoid valves has been passively qualified to ensure that the PORVs remain in the closed position when exposed to a harsh environment.
Thus failure of the solenoid will not result in a spurious actuation of the PORV as a result of short circuits or loss of power in accident conditions.
Based on the licensees I
submittal, it can be concluded that the PORV circuitry meets the qualification requirements of NUREG-0737 Item II.D.1.
The presentation above, demonstrating that the valves operated satisfactorily, verifies that the portion of Item 1 of Section 1.2 that requires conducting tests to qualify the valves and that part of Item 7 requiring that the effect of discharga piping on operability be considered have been met.
Also, the qualification of the PORV circuitry under 10 CFR 50.49 is considered to satisfy Item 5 of Section 1.2.
4.5 Piping and Suoport Evaluation In the piping ano support evaluation, the safety / relief valve piping between the valve discharge flanges and the pressurizer relief tank were analyzed for the requirements of the ASME,Section III, Code. The pipe 18
supports were analyzed for the requirements of the AISC Code, Sections 1.5.1.3, 1.6.1, and 1.6.2.
The load combinations and acceptance criteria were equivalent to those recommended by the EPRI piping subcommittee as presented in Reference 22.
The transient conditions analyzed were based on Reference 7 and included discharge of saturated steam or 4000F water at assumed valve opening pressures of 2575 psig (saturated steam) and 2500 psig (4000F
)
~
water) for the SRVs and 2450 psig (saturated steam) and 2500 psig (4000Fwater) for the PORV.
For the saturated steam analyses a pressurization rate of 175 psi /sec was assumed in ramping the pressurizer pressure from the set point to the maximum pressure (SRVs.open--2662 psig, PORV open--2575 psig).
The forces generated from these conditions bound
]
those from all other conditions expected at the plant including discharge of f
water following a feedwater line break combined with High Pressure Injection (Reference 14).
In addition, in Reference 16, SMUD stated that a reanalysis
)
effort is currently underway to improve the model and fully verify system i
adequacy under all design basis loading ir.cluding subcooled water cases.
i l
The thermal-hydraulic analysis was performed with the RELAP5 code.
Reference 21 has shown RELAPS is a suitable' tool for the calculation of discharge loads.
The program RELAP-FORCE was used to produce. force time histories for each piping segment from the output of the RELAPS analysis.
Verification of the RELAP-FORCE calcuations was provided in Reference 16.
A RELAP5 model for the safety and relief valve piping from the valve discharge to the relief tank was developed.
Average control volume lengths in the model ranged from 1.25 ft to 2.10 ft.
These lengths we're 2.5 to 4.2 times larger than the maximum size recommended in Reference 21. The reference recommended a maximum volume length of 0.5 ft and 5 to 10 volumes per bounded segment. The longer volume lengths were used in the Rancho Seco analysis because of computer limitations.
The submittal indicated the lack of sufficient detail in the RELAPS model was compensated for by applying intensification factors to the forces computed on pipe segments with too.few-19 I
j
i control volumes.
These intensification factors were developed from a nodalization study performed with the model of one SRV discharge line which had twice the number of volumes as the same line in the complete model.
Further clarification of the methodology used to develop the intensification factors was presented in Reference 16.
Based on the information presented by the licensee, it can be concluded that the analytical results conservatively bounds the discharge piping loads.
In the piping model, the key parameters of choked flow locations, and valve opening times were reviewed and found to be acceptable. discharge case.
In Reference 16, it was stated that the maximum time step used was 2 x 10-4 sec based on the recommendations made in Reference 22. A question regarding the PORV opening time used in the analysis resulted from a typographical error in Reference 14; which stated that a PORV opening time of 0.50 see was used, while the longest Dresser PORV opening time in the test report was 0.23 sec.
In Reference 16 SMUD stated that the opening time used in the thermal-hydraulic analysis was 0.05 sec.
To account for uncertainties in valve flow rates, the flow rate in the piping analysis was conservatively adjusted. ' A conservative factor of 1.15 was included in the maximum rated valve mass flow rate for the PORV and SRVs. The conservative valve flow rates used in the analysis acceptably account for 10% ASME derating and potential error in the flow rate.
Two valve opening cases were addressed in the analysis.
One was a simultaneous opening of the two safety valves, without PORV' discharge, the other the opening of the relief valve without SRV discharge. This approach-is reasonable since the two safety valves' are identical and have the same set pressure and there is only a single PORV. Maximum forces in the common header region of the piping system could theoretically be expected when the.
opening sequence is such that the initial pressure waves from valve opening reach the common junction downstream simultaneously. This event is unlikely, however, because the valves would be required to open et times perfectly spaced to compensate for differing piping lengths leading to the common junction.
'20
r The structural analysis was performed using Stone & Webster Engineering Corporation's (SWEC) version of NUPIPE.
This is a widely used program which has been fully verified for pipe stress analysis. The NUPIPE code was benchmarked by the NRC in 1979 as part of a five plant review conducted by SWEC.
i The key parameters of lumped mass spacing, and damping aro adequate.
A cutoff frequency of 50 Hz (Reference 14), which is more conservative than
~
the NRC guideline of 33 Hz, was used for the seismic analyis.
For the thermal-hydralic analysis, a cutoff frequency of 300 Hz was used (Reference
- 16) which is consisedered satisfactory.
i The results of the piping and support analysis identified a number of j
locations where support modifications are needed to relieve overstressed locations. These include the 2 in, square bars which connect the support on the pressurizer to the pressurizer lifting lugs, ten locations where the trunion attachments of the pipe supports produce the overstress, and the need for two new reinforcing pads at the tee connections to the common l
header from the PORV discharge line and one safety valve discharge line.
SMUD has stated the necessary modifications to the support system will be i
made during the cycle 8 refueling outage.
l The analyses discussed above, demonstrating that a bounding case has been chosen for the piping configuration, verifies Item 3 of Section 1.2 has been met, and item 8, requiring the the piping and support system be qualified by analysis have been met.
1 21 1
4 l
S.
EVALUATION
SUMMARY
The licensee for Rancho Seco 1 has provided an acceptable response to the requirements of NUREG-0737, thereby reconfirmaing that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 have been met with regard to the safety valves and PORV.
The rationale for this conclusion is given below.
The licensee participated in the development and execution of an acceptable relief and safety valve test program to qualify the operability j
of prototypical valves and to demonstrate that their operation pould not invalidate the integrity of the associated equipment and piping.
The j
subsequent tests were successfully completed under operating conditions which, by analysis, bound the most probable maximum forces expected from anticipated design basis events. The test results showed that the valves
)
tested functioned correctly and safely for all steam and water discharge events specified in the test program that were applicable to Ranch Seco Unit 1 and that the pressure boundary component design criteria were not j
exceeded.
SMUD has recognized the potential effects of water discharge on valve operability and has agreed to inspect the safety valves after 0
discharge of water with a temperature less than 550 F.
Analysis and review of both the test results and the licensee justifications indicated the performance of the prototypical valves and piping can be extended to the in plant valves and piping. The plant specific piping has been shown by analysis to be acceptable.
1he licensee has stated that the thermal-hydraulic analysis is being repeated with an improved model. The licensee also stated that the revised thermal-hydraulic analysis, which will be available in April 1987, will be provided to the NRC staff.
If the reanalyis effort shows that piping or support modifications are required, the licensee should submit the proposed changes and the proposed schedule for making the modifications to the NRC for review.
Thus, the requirements of Item II.D.1 of NUREG-0737 have been met (Items 1-8 in Paragraph 1.2).
Compliance with these requirements ensures that the reactor primary coolant pressure boundary will have a low 22
probability of abnormal leakage (General Design Criterion No. 14).
In addition, the reactor primary coolant pressure boundary and its associated components (piping, valves, and supports) have been shown by design to have J
been designed with a sufficient margin so that design conditions are not exceeded during relief / safety valve events (General Design Criterion No. 15).
Further, the prototypica; tests and the successful performance of the valves demonstrated that this equipment was constructed in accordance with high quality standards, meeting General Design Criterion No. 30.
l i
4 O
23
6.
REFERENCES 1.
TMI-Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578, July 1979.
I 2.
Clarification of TMI Action Plan Recuirements, NUREG-0737. November 1930.
1 3.
R. C. Youngdahl ltr. to H. D. Denton, Submittal of PWR Valve Test Report, EPRI NP-2628-SR, December 1982.
4.
EPRI Plan for Performance Testing of PWR Safety and Relief Valves, July 1980.
5.
EPRI PWR Safety and Relief Valve Test Program Valve Selection / Justification Report, EPRI NP-2292, December 1982.
6.
EPRI PWR Safety and Relief Valve Test Program Test Condition Justification Report, EPRI NP-2460, December 1982.
7.
Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves for B&W 177-FA and 205-FA Plants, EPRI NP-2352, December 1982.
8.
EPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Report, EPRI NP-2628-SR, December 1982.
9.
EPRI/ Marshall Electric Motor Ooerated Block Valve, EPRI NP-2514-LD, July 1982.
10.
Letter J. J. Mattimoe, SMUD, to D. G. Eisenhut, NRC, " Rancho Seco Unit 1 NUREG-0737 Item II.D.1 Relief and Safety Valve Testing,"
June 30, 1982.
11.
Letter R. J. Rodriguez, SMUD, to J. F. Stolz, NRC, " Rancho Seco Unit 1 NUREG-0737 Item II.D.1, Relief and Safety Valve Testing," July ?.9,
- 1983, 12.
Letter J. F. Stolz, NRC, to R. J. Rodriguez, SMUD, " Rancho Seco Nuclear Generating Station--Request for Additional Information," November 21, 1984.
13.
Letter J. J. Mai",noe, SMUD, to J. F. Stolz, NRC, " Response to 851121 (SIC) Request for Addl Info RE NUREG-0737, Item II.D.1," January 21, 1985.
14.
Letter J. R. Rodriguoz, SMUD, to J. F. Stolz, NRC, " Rancho Seco Nuclear Generating Station, Unit 1; NUREG-0737, Item II.D.1 Relief and Safety Valve Reliability," April 12, 1985.
15.
Letter J. F. Stolz, NRC, to R. E. Ward, SMUD, " Rancho Seco Nuclear Generating Station--NUREG-0737, Item II.D.1 Request for Information,"
December 16, 1986.
24
16.
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.
17.
" Coupled Valve Dynamic Model, for Isentropic, Two-phase and Subcooled Discharge, Technical Description," Continuum Dynamics, Inc., Tech Note 83-6, prepared for participating PWR Utilities ar.d EPRI, May 1983.
18.
" Safety Valve Dynamic Analyses for Dresser Industries' 31739A and 31759A Valves," Continuum Dynamics, Inc., Report No. 83-4, Rev. 1, prepared for B&W, December 1983.
19.
C. L. Nalezny, Supplement to Technical Evaluation Report TMI Action NUREG-0737 (II.D.1) Relief and Safety Valve Testing, Three Mile Island Unit 1 Docket No. 50-289, EGG-RST-6593 Supplement, June 1985, 20.
Letter F. P. Bolger, Dresser Ind., to J. V. McCulligan, SMUD, " Lower Ring Setting 31759 Valve, Rancho Seco Unit 1," March 30, 1983.
- 21. Pressurizer Safety Valve Maximum Allowable Blowdown, B&W Report 77-113-5671-00, August 1982,
- 22. EPRI PWR Safety and Relief Valve Test Program Guide for Application of Valve Test Program Results to Plant-Specific Evaluations, Revision 2, Interim Report, July 1982.
- 23. Application of RELAP5/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads, EPRI-2479, December 1982.
e 25
l NIC FORM 335 U 8. NUCL(AR RBOULATORY COMMI5440N i REPOH F NUMetR ease,p.ew av /80C. Joe vet Ne, et says 00 2o $2 '
BIBLIOGRAPHIC DATA SHEET f
SEE INSTRUCTIONS ON THE REVERS $
J (g Avg 3 LANE 2 TITLE AND Sug flTLS TMI Action--NUREG-0737 (II.D.1) Relief and Safety Valve Testing: Rancho Seco, Unit 1 4 OATI REPORT COMPLtTEO May 1987
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C. P. Fineman C. L. Nalezny l1987 MONT-rear May
- 7. FtRPORMtNG ORGANQAflON NAME AND MatuNG AOORESS ffaeresse de Co-t 8 PROJECT /T A$A/ WORE UNIf NUM96R 4
INEL-EG&G Idaho, Inc.
- a oa ca^ar avaaa Idaho Falls, ID 83415 A6492 d e,
" "" o* auca'
,o 5,0NSOR,~o ORoANe AT ON Naut ANO MA uNo ADORen u~~-e Mechanical Engineering Branch Informal Office of Nuclear Reactor Regulation
- " " ' ' "'"' "~~**~'
U.S. Nuclear Regulatory Commission Washington, D.C.
20555 11 $UPPLEMENT AR Y NOTES lJ t.85iR ACT UCC was er ' esso 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 the TMI Action Plan Requirements) recommended that programs be developed arid 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 and the responses of the licensee of Rancho Seco, Unit 1, to the requirements of NUREG-0578 and NUREG-0737 by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc. This review has found the Licensee has provided acceptable responses, which reconfirm that the General Design Criteria 14,15, and 30 of Apprndix A to 10 CFR 50 have been met.
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