ML20235B023
| ML20235B023 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 12/31/1988 |
| From: | Fineman C, Nalezny C, Yuan C EG&G IDAHO, INC. |
| To: | NRC |
| Shared Package | |
| ML20235B026 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-8326, NUDOCS 8902020412 | |
| Download: ML20235B023 (32) | |
Text
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'. . ENCLOSURE f
EGG-NTA-8326
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TECHNICAL EVALUATION REPORT TMI ACT10N--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING ZION, UNITS 1 AND 2 DOCKET Nos. 50-295, 50-304 6
C. Y. Yuan C. L. Nalezny C. P. Fineman December 1988 Idaho National Ercineering Laboratory EG&G Idaho, Inc.
Idaho Falls, Idaho 83415 Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE-AC07-761D01570 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
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coolant systems. 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 their results by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc.
Specifically, this review examined the response of the Licensee for Zion, Units 1 and 2, to the requirements of NUREG-0578 and NUREG-0737. This review found the Licensee has not provided an acceptable response and, thus, has not reconfirmed that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 were met.
FIN No. A6492--Evaluttion of OR Licensing Actions-NUREG-0737, !!.D.1 O
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CONTENTS ABSTRACT .............................................................. ii
- 1. INTRODUCTION ..................................................... I 1.1 Background ................................................. I 1.2 General Design Criteria and NUREG Requirements ............. 1
- 2. PWR OWNERS' 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 4.2.1 FSAR Steam Transients .............................. 8 4.2.2 FSAR Liquid Transients ............................. 9 4.2.3 Extended High Pressure Injection Event ............. 12 4.2.4 Cold Overpressure Transients ....................... 12 4.2.5 PORV Block Valve Fluid Conditions .................. 13 4.2.6 Test Conditions Summary ............................ 14 4.3 Operability ................................................ 14 4.3.1 Safety Valves ...................................... 14 4.3.2 Power Operated Relief Valves ....................... 17 4.3.3 Electric Control Circuitry ......................... 18 4.3.4 PORV Block Valves .................................. 18 4.3.5 Operability Summary ...... ......................... 19 4.4 Pi ping and Support Eval uation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 4.4.1 Thermal Hydraulic Analysis ......................... 20 4.4.2 Stress Analysis .................................... 22 4.4.3 Piping and Support Summary ......................... 24
- 5. EVALUATION
SUMMARY
............................................... 25 5.1 NUREG-0737 Items Fully Resolved ............................ 25 5.2 NUREG-0737 Items.Not Resolved .............................. 26
- 6. REFERENCES ....................................................... 28 TABLE EPRI tests on Crosby HB-BP-86 6M6 safety valve ........................ 15 lii
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TECHNICAL EVALUATION rep 0RT TMI ACT10N--NUREG-0737 (II.D.1)
RELIEF AND SAFETY VALVE TESTING ZION, UNITS 1 AND 2 p0CKET NDs. 50-295. 50-304
- 1. INTRODUCTION
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 reseat.
From these past instances of improper valve performance, it is not known whether they occurrtd because of a limited qualification of the valve or
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because of a basic unreliability of the valve design. It is known that the failure of a pcwer-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 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 (a) the reactor primary coolant pressure boundary be designed, fabricated, and tested so as to have an extremely low probability of abnormal leakage, (b) the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient ma* gin to assure that the design conditions are 1
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not exceeded during normal operation or anticipated transient events, and (c) the components which are part of the reactor coolant pressure boundary shall be con'structed to the highest quality standards practical.
To reconfirm the integrity of overpressure protection systems and .
thereby assure that the General Design Criteria are met, the NUREG-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 (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 operating conditions through the use cf 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.
- 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.
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- 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 are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (FSAR). The effect l 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 a,nalysis.
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- 2. PWR OWNERS' 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 hnplementing a generic test program for pressurizer power operated relief valves, safety valves, block valves, and associated piping systems. The Commonwealth Edison Co. (Ceco), owner of Z'ori. Units 1 and 2, 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.
EpRI developed a plan (Reference 4) for testing PWR safety, relief, and block valves ender conditions which bound actual plant operating conditions. EPRI, through the valve manufacturers, identified the valves used in the overpressure' protection systems of the participating utilities.
I 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 Npelear 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).
EpRI contracted with Westinghouse Electric Corp. to produce a report on the inlet fluid conditions for pressurizer safety and relief valves in Westinghouse designed plants (Reference 7). Since Zion, Units 1 and 2, were l designed by Westinghouse this report is relevant to this evaluation.
Several test series were sponsored by EPRI. PORVs and block valves were tested at the Duke Power Company Marshall Steam Station located in Terrell, North Carolina. Additional PORV tests were conducted at the Wyle Laboratories Test Facility located in Norco, California. Safety valves were tested at the Combustion Engineering Company, Kressinger Development Laboratory, located in Windsor, Connecticut. The results for the relief and safety valve tests are reported in Reference B. The results for the block valves tests are reported in Reference 9.
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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 i to steam, subcooled water, and steam to water transition. Additional objectives were to (a) obtain valve capacity data, (b) assess hydraulic and structural effects of associated piping on valve operability, and (c) obtain piping response data that could ultimately be used for verifying analytical
- piping models.
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 The plant specific evaluation of the adequacy of the overpressure l
protection system for Zion, Units 1 and 2, was submitted by CECO to the NRC on July 1, 1982 (Reference 11). A request for additional information was sent to CECO by the NRC on February 19, 1985 to which CECO responded on June 18, 1985 and January 22, 1986 (References 12, 13 and 14). A second request for information was ser.t to the Licensee on March 31, 1987 (Reference 15) to which CECO responded on November 19, 1987 (Reference 16).
The response of the overpressure protection system to Anticipated 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 j 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 Zion, Units 1 and 2, are four-loop PWRs designed by the Westinghouse Electric Co. Each unit is equipped with three (3) safety valves, two (2) PORVs, and two (2) PORV block valves in its overpressure protection system. The safety valves are 6-in. Crosby Model HB-BP-86, 6M6, spring loaded valves with loop seal internals. The design set pressure is 2485 psig and the rated steam flow capacity is 420,000 lbm/h. The PORVs are 2-in. Copes-Vulcan Model D-100-160 globe valves with 316 SS stellited plugs and 17-4 PH cases. The PORV open'ing set pressure is 2335 psig and the rated steam flow capacity is 210,000 lbm/h. The PORV block valves are 3-in. Velan Model B10-3054B-13MS gate valves with Limitorque SMB-00-15 motor operators.
The inlet pipe to the safety valve includes a cold loop seal; the inlet to the PORV has no loop seal.
Valves tested by EPRI included a Crosby safety valve identical to those installed at Zion, Units 1 and 2. Test results for the safety valve are directly applicable to the safety valves at Zion.
The PORV used in the EPRI tests was a 3-in. Copes-Vulcan globe valve, Model D-100-160, with 316 SS stellited plug and 17-4PH cage. The Zion PORV is an older version of the test valve which has the same design and performance characteristics except a smaller valve body. The EPRI test results are applicable to the Zion FORVs.
Valves tested by EPRI included a Velan block valve identical to those installed at Zion, Units 1 and 2. However, the test results for the block valve are not applicable to the block valves at Zion. For the block valves, the Licensee must demonstrate that the block valve operators are set to produce a torque greater ttA. ite minimum torque used in the EPRI tests in order for the test results to be directly applicable to the plant valves.
Based on the information provided by Ceco, it cannot be concluded the torque output of the plant operators is greater than the minimum torque used in the EPRI tests. Therefore, the block valve test results are not considered directly applicable to the Zion valves.
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Based on the above, the safety valves and PORVs tested are considered represent'ative of the in-plant valves at Zion, Units 1 and 2, and to have fulfilled the part of the criteria of Items 1 and 7 as identified in Section 1.2 regarding applicability of the test valves. The block valves at Zion, Units 1 and 2, are the same as the valve tested by EPRI. Therefore, -
Item 1 of Section 1.2 was met for the block valves. However, the test results for the block valve / operator combination tested by EPRI are not considered to be applicable to the block valve / operator combination at Zion, Units 1 and 2, and, therefore, Item 7 of Section 1.2, regarding applicability of the test valves, was not met for the block valves.
4.2 Test Conditions As stated above, Zion, Units 1 and 2, are four-loop PWRs designed by the Westinghouse Electric Corp. The valve inlet fluid conditions that bound the overpressure transients for Westinghouse designed PWR Plants are identifie; in Reference 7. The transients considered in this report include FSAR, extended high pressure injection, and cold overpressurization events.
The expected fluid conditions for each of these events and the applicable l EPRI tests are discussed in this section.
4.2.1 FSAR Steam Transients l
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For the Zion PWRs, the limiting events for the FSAR transients 3
resulting in steam discharge through the safety valves alone and in steam i discharge through both the safety and relief valves are the loss of load event (for maximum pressurizer pressure) and the locked rotor event (for the maximum pressurization rate).
In the case when the safety valves actuate alone, the maximum pressurizer pressure and maximum pressurization rate are predicted to be j 2555 psia and 144 psi /s, respectively. The maximum developed backpressure in the outlet piping is 670 psia (Reference 16). The loop seal temperature is approximately 1150F at the valve inlet and 6680F nine feet upstream of the valve (Reference 13).
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EPRI tests representative of the valve inlet fluid conditions for the limiting transient were selected for the plant specific evaluation. In selecting the EPRI tests, the safety valve ring settings and the pressure drop through the inlet pipe were also considered. For steam flow conditions, four loop seal discharge tests (Test No. 929, 1406, 1415, 1419) were applicable to Zion, Units 1 and 2. These tests were performed with valve ring settings representative of the typical ring settings used in Westinghouse PWRs including Zion. The ring settings used in these tests were (-71, -18) or (-77, -18). These represent the upper and lower ring positions measured from the level position referenced to the bottom of the disc ring. The safety valves at Zion use the Crosby recommended ring settings (Reference 13). Since both the test ring settings and the in-plant ring settings were determined by the valve manufacturer, the Crosby Valve and Gage Co., using the same methods and the same standard of performance, these two sets of ring settings are considered comparable to each other.
The loop seal temperature measured in the tests ranged from 90 to 3600F at the valve inlet. The maximum pressurizer (tank 1) pressures were in the range of 2675 to 2760 psia and the pressurization rate was 90 to 360 psi /s.
The backpressure developed in the tests were 245 to 710 psia. The above data show that the inlet fluid conditions and backpressure of these tests envelop the corresponding fluid data predicted for the Zion safety valves.
When both the safety valves and PORVs are actuated, the maximum pressurizer pressure *5 predicted to be 2532 psia and the maximum pressurization rate is 130 psi /s. In the EPRI tests on the Copes-Vulcan PORV, the maximum steam pressure at valve opening was 2715 psia, which bounds the predicted pressure at Zion. The backpressure developed at the outlet of the PORVs is not an important consideration, since the air operated PORVs used at the Zion plant are not sensitive to backpressure (Reference 6). Therefore the EPRI test inlet fluid conditions for the PORV in steam discharge are representative of the plant specific transient conditions.
4.2.2 FSAR Liould Transients The limiting FSAR transient resulting in liquid discharge through the PORVs and safety valves is the main feedline break accident (Reference 7).
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According to the Licensee, this event was not analyzed in the Zion FSAR because it was not a design transient required for NRC licensing when Zion was built. The Licensee further stated that, based on a probabilistic risk study of high pressure liquid challenge to the safety / relief valves, the calculated frequency of occurrence of water discharge was extremely low ~
(Appendix A of Reference 11). Liquid discharge in a feedline break, extended high pressure injection, or cold overpressurization event is unlikely to happen and, therefore, need not be considered.
However, the Westinghouse Valve Inlet Fluid Condition Report (Reference 7) stated that the main feedline pipe rupture event was classified'as a Class IV licensing event. That is, one which was not expected to take place but was postulated because its consequences include the potential for the release of a significant amount of radioactive material. Also, NUREG-0737 specifically requires the safety valves and PORVs be Qualified for inlet fluid conditions resulting from transients and accidents referenced in Regulatory Guide 1.70, Rev. 2. The feedwater line break is specifically defined in Regulatory Guide 1.70, Rev. 2. From the review of feedwater line break analyses of plants similar to Zion (see below), it is clear that the feedwater line break is most likely to be the limiting transient for providing high pressure liquid to the safety valves, a fluid for which they were not originally designed. Therefore, in accordance with the NUREG requirements, the feedline break event should be analyzed even though the probabilistic analysis showed that the frequency of occurrence is extremely low.
Reference 7 provides the pre'dicted fluid transient conditions for the feedline break event for most of the Westinghouse designed PWRs, but Zion, I Units 1 and 2, were not among them. Among the nuclear power plants owned by ,
CECO, Zion, Byron, and Braidwood are all four loop PWRs designed by Westinghouse. Since these plants are similar in design and capacity (Zion, 1040 MWe; Byron and Braidwood, 1120 MWe), the inlet conditions to the overpressure protection systems in all plants should be similar. The Byron and Brtidwood analyses should bound those for Zion since the ratio of asymptotic surge rate to safety valve capacity for Zion _is 1.352 while the ratio for Byrcn and Braidwood is 1.619 (see Reference 7). The results of a feedline break analysis, if performed for Zion, are, therefore, expected to 10
p 4 be similar to those of Byron and Braidwood. The Byron and Braidwood feedline break analyses indicated that the safety valves and PORVs opened on saturated steam at about seven minutes into the transient and steam te saturated liquid transition would follow at thirteen minutes into the event. Therefore, liquid discharge through the safety valves cannot be ruled out. Pending the receipt of information from Ceco on a plant specific evaluation of the feedwater line break, the Zion valves will be reviewed using conditions from the Byron and Braidwood analyses.
Reference 7 showed that in a feedline break accident at Byron and Braidwood, the maximum pressure at the safety valve inlet during liquid discharge was calculated to be 2508 psia and the pressurization rate was 3.5 psi /s. Fluid temperature at the valve inlet ranges from 615 to 6350F and the maximum liquid surge rate into the pressurizer is'569 ppm. <
In a feedlire break accident resulting in safety valve actuation, water discharge is always preceded by steam and steam to water transition flows.
Among the EPRI tests performed on the 6M6 valve, Tests 931a and 931b were performed for loop seal / steam, steam to water transition, and water discharge conditions. In addition, Test 932 was performed for water discharge only. The valve ring settings and inlet pipe configuration used in these tests were comparable to those of the in-plant safety valves.
Test 932 was performed at an inlet temperature approximately 1600F below that predicted for the in-plant valve, thus it is not applicable to the Zion safety valves. In Test No. 931a, the maximum inlet pressure was 2578 psia.
The pressurization rate was 2.5 psi /s, the inlet fluid temperature was 1170F, and the tank fluid temperature was 6350F. After the valve closed in Test 931a, the system was allowed to repressurize and the valve cycled on approximately 6400F water (Test 931b). Since the inlet temperature and pressure of the tests compare favorably with the predicted in-plant condition, the results of these tests are considered representative of the Zion safety valves.
The expected fluid conditions at the inlet of the safety valve described above was based on a Westinghouse analysis which assumed that the PORVs were not operable during the feedline break transient. If the PORVs are operable, the same fluid conditions postulated for the safety valve 11
inlet can also be expected at the PORV inlet-(Reference 6). In the EPRI tests, high temperature water discharge and steam to water transition tests were performed with the Copes-Vulcan PORV. In the water discharge test.
Test No. 76-CV-316-2W, the maximum pressure at the valve inlet was 2535 psia and the temperature was 6470F. In the transition test, Test No. 77-CV-316-75/W, tha maximum inlet pressure was 2532 psia and the water temperature was 6570F. The inlet fluid conditions for these tests compare well with the predicted maximum pressure and temperature of 2508 psia and 635DF for the Zion plant. Therefore this test is adequate to represent the in-plant PORV performance in the feedline break event.
4.2.3 Extended High Pressure Injection Event The limiting extended high pressure injection event is the spurious actuation of the safety injection system at power (Reference 7). For a tour-loop plant, both the safety valves and PORVs will be challenged. Both steam and water discharge are expected. In this event, however, the safety valves or PORVs open on steam and liquid discharge would not be observed until the pressurizer becomes water solid. According to Reference 7, this would not occur until at least 20 minutes into the event which allows ample time for operator action. Thus the potential for liquid discharge for extended HPI events can be disregarded.
4.2.4 Cold Overpressurization Transient The PORV is used for overpressure protection during the low temperature stages in reactor start-up and shutdown operations. According to the Licensee, the low pressure setpoint of the PORV is conservatively assursed to be 500 psig instead of the Zion Technical Specification limit of 435 psig. .
The maximum liquid pressure that can be reached in a postulated cold overpressure event is 519 psig. The inlet liquid temperature ranges from ,
100 to 4570F (Reference 11).
For steam discharge through the PORV, the high pressure steam tests discussed in Section 4.2.1 would cover the low pressure steam conditions l
predicted for the cold overpressure transient. For water discharge conditions, there were two low pressure and low temperature water tests 12
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. . i performed on the Copes-Vulcan PORV with stellited plug and 17-4 PH cage.
The tests were conducted at an inlet pressure of 675 psia and water temperatures of 105 and 4420F, respectively. These conditions are representative of those at Zion. Therefore, the EPRI tests can be used to j evaluate the performance of the Zion PORV in the cold overpressurization transient.
4.2.5 PORV Block Valve Fluid Conditions The PORV block valves are required to operate over the same range of fluid conditions as the PORVs. The Velan B10-3054B-13MS block valve with Limitorque -SB-00-15 actuator was subjected to 21 cycles of steam tests
-against full flow. Steam pressure upstream of the test valve varied from 2455 psia to 2515 psia during opening cycles and from 2355 osia to 2425 psia during closing cycles. These pressures are well above the PORV opening
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pressure of 2350 psia. l 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 9) demonstrated that the torque required 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, according to friction tests cone by Westinghouse on a stellite coated specimen, the friction 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. The Velan block valves have stellite coated disks and seats. The force required to overcome disk friction in steam is essentially equal to the force required in water.
Therefore, the steam tests are adequate to demonstrate the operability of the block valves for expected water conditions.
However, as noted in Section 4.1, it could not be determined whether l
the plant block valve operatot was set to produce a torque greater than the minimum torque tested by EPRI. Therefore, although the EpRI tett conditions bound those for the plant block valves, the test results are not applicable to the Zion, Units 1 and 2, block valve / operator combination.
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l 4.2.6 Ttst Conditions Summary The test sequences and analyses described above demonstrate that the test conditions bound the conditions for the plant valves. They also verify ,
that Items 2 and 4 of Section 1.2 were met, in that conditicns for the -
operational. occurrences were 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, was also met.
4.3 Operability 4.3.1 Safety Valves The EPRI tests representative of steam discharge conditions for the Zion safety valves are the loop seal tests on the Crosby SM6 valve, Test No. 929, 1406, 1415, 1419. In all these tests (except Test No. 1415), the valve fluttered or chattered during loop seal discharge and stabilized when steam flow started. The valve opened within +2% of the design set pressure and closed with 5.1 to 9.4% blowdown. Up to 111% of rated flow was achieved at 3% accumulation with valve lift positions at 92 to 94% of rated lift. i These tests demonstrated that the valve performed its function in spite of the initial chatter during loop seal discharge.
In Test 1419, the valve chattered on closing and the test was terminated after the valve was manually opened to stop the chatter. This result does not indicate a valve closing problem for the Zion safety valve since an identical test (Test 1415) had already demonstrated that the valve performed satisfactorily and exhibited no sign of instability. The closing chatter in Test 1419 may possibly be a result of the repeated actuation of the valve in loop seal and water discharge tests. As shown in the table on ;
the next page, the 6M6 test valve was subjected to seventeen steam, water,
- I and transition tests. In the first four or five tests, the valve fluttered and chattered during loop seal discharge but stabilized and closed successfully. After Test 913, there were four instances in which the test was terminated due to chattering on closing. Galled guiding surfaces and damaged internal parts were found during inspection and the damaged parts l
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- t. e e .= e == e e % M. G 4.s l U I= =J -K *=
l 15 ,
i
_ _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ . _ _ _ _ _ _ _ _ . _ - - - - _ _ - _ _ - _.i
were refurbished or replaced before the next test started. The test results showed that the valve performed well after each repair, but the closing chatter recurred in a subse,tuent test. Test 1415 was performed immediately j i
after valve maintenance 6rd the valve performed stably. The next test I (Test 1419) encountered chatter in closin) even though it was a repeat of ~
Te:it 1415 at similar fluid conditions. This suggests that inspection and maintenance are important to the continued operability rf the valves. The Licensee should tilvelop a formal procedure requiring that th? safety valves be inspected after each actuaticn and this procedure incorporated into the plant operating procedures or licensing documents such as the plant technical specifications.
The blowdown in these tests (5.1 to 9.4%) were in excess of the 5%
value specified by the valve manufacturer and the ASME Code. Westinghouse performed an ahalysis, " Safety Valve Contingency k,alysis in Support of the EPRI Stfety/ Relief Valve Testing Program--Volume 3: Westinghouse Systems."
EPRI NP-2047-LD, October 1981, on the effects of incransed blowdown and concluded that no adverse effects on plant safety occurred in that the reactor core remained Lovered. Therefore, the amount of increased blowdown occurred in the Crosby 6M6 steam tests is considered acceptable.
As discussed in Section 4.2.2, the limiting FSAR transient resulting in liquid discharge is the main feedline break accident. Tests 931a and 931b with typical plant ring settings of (-71, -1B) simulate the expected Zion feedwater line bretk conditions. Test 931a was a loop seal / steam / water transition test. The 6M6 valve initially opened, fluttered or chattered in a partial lift position during loop seal discharge, then popped open, stabilized on steam, ano closed with a 12.7% hlowdown. Test 931b was a saturated water test. The SM6 valve opened on 6400F water, chattered, and ,
then stabilized. The valve closed with 4.8% blowdown. For these tests the valve opened within -1% and +3% of the set pressure. The maximum calculated surge rate at Zion, Units 1 and 2, during the fsealine brer.k transient was estimated to be 569 gpm. The 1,icensee did not provide information on i
feedline break accident for Zion. The above surge rate is based on the I
predicted value for Byron /Braidwood given in Table 5-2 in Reference 7.
Zion, Byror and Braidwood are similar plants; their feedline break analyses should produce similar results. The 6M6 valve tested by EpRI passed i 16
2355 gpm at 2415 psia and 6410F which is much higher than the predicted s
flow rate for Zion. The above results demonstrate that the Zion safety valv?s would be adequaw to perform the required water relief function.
Bending roments as high as 298,750 in-lb (Test 908) were induced on the discharge flange of the Crosby 6M6 test valve and had no adverse effect on valve performance. This applied moment is 6% less than the maximum estimated bending moment of 315,328 in-lb for the Zion 182 valves (Reference 16). Because the test valve operated properly with the applied moment and because the plant specific moment is less than 6% more than the test moment, the performance of the plant valves is also expected to be unaffected.by bending moments imposed during discharge transients.
One way to estimate the stability of the in-plant valves is to compare the plant specific inlet pining pressure drop on valve opening to the inlet presrure drop in the EpRI tests. The Crosby 6M6 safety valve was tested on an inlet piping configuration with a pressure drop of 263 psid on valve opening and 181 psid on valve closing. CECO, in Reference 16, provided the Zion inlet pressure drop in the form of a pressure plot from one of the RELAPS analyses completed by its consultant as part of the structural knalysis of the pressurizer piping. This figure showed the pressure at the valve inlet dropped almost 800 psi when the valve opened. This is significantly larger than the pressure drop for the EPRI test
- configuration. Therefore, the information provided by CECO has not provided evidence assuring the valves will operate stably in the plant specific piping configuration.
4.3.2 power Operated Relief Valve The EPRI tests on the Copes-Vulcan PORV with 316 SS stellited plug and 17-4 PH cage demonstrated that the valve opened and closed on demand in steam, water, and steam to water transition conditions. The opening and closing time were within the 2.0 second opening and closing time normally
-> quired for Westinghouse PWRs. The lowest steam flow rate observed in the tests was 232,000 lb/h which exceeded the rated flow of 210,000 lb/h for the Zion PORVs.
j 17
[ 4
The predicted value of the maximum bending moment induced at the Zion PORV discharge flange was not provided by the Licensee. The Licensee was requested to provide this information in Reference 15 and Ceco stated in Reference 16 that the moment was contained in an attached consultant report. However, the maximum calculated bending moment on Zion PORVs was ,
not found in the report. Therefore, operability of the Zion PORVs with the plant specific bending moments cannot be assured.
l 4.3.3 Electric Control Circuitry NUREG-0737 II.D.1 required qualification of the associated control circuitry as part of the safety and relief valve qualification task. The specific electric circuits under consideration are the control circuits of the PORVs. The Nuclear Regulatory Commission staff agreed that meeting the licensing requirements of 10 CFR 50.49 for this circuitry is satisfactory and that specific testing per NUREG-0737 is not required. According to the Licensee, all of the electrical components in the PORV control circuitry that can be exposed to a harsh environment (solenoid valves, limit switches, cable, terminal blocks, electrical penetrations) were included in the 10 CFR 50.49 Environmental Qualification Program (References 13 and 17).
The inclusion of the PORV control circuitry in the 10 CFR 50.49 review will provide adequate verification that the control circuitry will function properly and, therefore, specific testing to meet NUREG-0737 is not needed.
4.3.4 PORV Block Valves The Velan block valve was subjected to 21 cycles of steam tests against full flow at 2355 to 2515 psia. The valve opened and closed on demand and the stroke time was recorded at 9.7 to 9.9 s. The test pressures were above the Zion PORV opening pressure of 2350 psia and the stroke time war,within the specified stroke time of 10 s for the SMB-00-15 valve operator.
The plant specific block valve has a Limitorque SMB-00-15 actuator which is similar to the SB-00-15 actuator used in the EPRI tests. The difference between these two actuators is that the SB actuator incorporates a spring pack compensator on the stem nut which enhances the performance of the actuator in high speed and high temperature service. As far as the 18
i 1
1 plant specific application at Zion is concerned, the SB and SMB operators are practically the same. Therefore, the Zion PORV block valves are .}
expected to provide performance similar to the test valves with steam l
{
discharge conditions.
f 1
Tests for water flow for the Velan block valve were not performed in the EPRI test program. As explained in Section 4.2.5 of this report, the !
valve behavior under the water flow condition is expected to be similar to that of the full pressure steam tests. Therefore, the operability of the valves for liquid flow condition was indirectly demonstrated.
As discussed in Section 4.1, in order to show the EPRI block valve tests are directly applicable to the Zion block valves, the Licensee must show the plant operator is set to produce a torque greater than the minimum torque used in the EPRI tests. Based on the information provided by the i Licensee it cannot be concluded this is the case. I 4.3.5 Operability Summary The above discussion verifies that the part of Item 1 of Section 1.2 that requires conducting tests to qualify the valves was met for the safety valves, PORVs, and block valves in so far as the applicable test valves are concerned. However, because of differences between the in-plant and test valve installation, the tests do not qualify the safety valves, PORVs, or block valves. This is because the Licensee failed to meet Item 7 of Section 1.2. For the safety valves, this item was not met because the information provided by CECO indicated the inlet piping pressure drop at Zion 1&2 was significantly larger than that in the EPRI tests. Therefore, stable operation of the safety valves in the plant specific configuration could not be assured. Also, the Licensee must document a formal procedure for the inspection of safety valves as discussed in Section 4.3.1. The part of Item 7 that requires the effect of discharge piping on operability be considered was met for the safety valves but not for the PORVs. This is because the Licensee did not provide a comparison of the maximum calculated bending moment for the plant PORVs to that in the EPRI tests. For the block valves, as noted in Section 4.3.4, the test results are not directly l
19
applicable to the in-plant valves. -The licensing action for 10 CFR 50.49 is considered to satisfy Item 5 of Section 1.2. -
4.4 piping and Support Evaluation This evaluation covers the piping and supports upstream and downstream of the safety valves and PORVs extending from the pressurizer nozzle to the relief tank. The piping system was designed for deadweight, internal pressure, thermal expansion, earthquake, sod safety valve and PORV discharge conditions. The same analysis method and valve discharge conditions were used for Units 1 and 2. The calculation bi the time histories of hydraulic forces due to valve discharge, E.e methot vi structural analysis, and the load combinations and stress evaluation are discussed below.
4.4.1 Thermal Hydraulic Analysis Pressurizer fluid conditions were selected for use in the thermal hydraulic analysis such that the calculated pipe discharge forces would bound the forces for any of the FSAR, Hpl, and cold overpressurization events, including the single failure that would maximize the forces on the valve.
Information on the thermal hydraulic analysis of the safety valve and PORV piping system was provided for loop seal / steam discharge throu,gh safety valves only (see response to Question 12, Reference 14). Water discharge through the safety valves, as in the feedwater line break, was not addressed. However, the loads resulting from the loop seal / steam discharge would bound those from a water discharge transient. For the loop seal / steam discharge analysis, the loss-of-load (maximum pressure) and the locked rotor (maximum pressurization rate) events were chosen as the limiting conditions ~
that would generate the highest piping loads. The three safety valves were assumed to open simultaneously. This approach is reasonable because the -
three safety valves are identical and have the same set pressure. Maximum forces in the common header could theoretically occur when the valve opening sequence is such that the initial pressure waves from valve opening reach the common header junction downstream simultaneously. This event is unlikely, however, because the valves would be required to open at times 20
4 perfectly spaced to compensate for differing piping lengths leading to the common junction. Thus, the assumption of simultaneous valve opening is acceptable.
Steam and water discharge through the PORVs was not addressed by Ceco or its consultant in the NUREG-0737, Item II.D.1, submittals reviewed. In Reference 16. CECO stated that only the loop seal / steam discharge through the safety valves was analyzed because the existing system configuration was previously evaluated for all other cenditions. This analysis showed the system met the applicable code requirements for the other possible conditions. The information provided by CECO did not discuss what was meant -
by all other conditions. Also, references were not provided so that the previous work could not be reviewed. Thus, the discussion below only evaluates the loop seal / steam analysis provided by CECO.
The thermal hydraulic analysis was performed using the RELAP5/ MODI computer code. RELAP5 calculates the thermal hydraulic properties of the fluid as a function of time in each control volume and at each junction of the piping model. The RELAP5 results are then used as input to a post processor, REPIPE, to calculate the force time histories acting on the piping system. RELAPS is widely used in the industry and was shown to be an adequate tool for predicting piping discharge loads (Reference 18).
Verification of the REPIPE code, part of the CDC CYBERNET system, was reviewed as part of the D. C. Cook, Units 1 and 2, submittal (Reference 19). This review found the code adequate for converting the RELAP5 output to piping forces.
In the RELAPS analysis of the steam discharge condition, the safety valves were assumed to actuate at a set pressure of 2499.7 psia. The valve opening time was assumed to be 0.88 s and 0.0145 s for the simmer and pop open period respectively, which are representative of the simmer and pop times for Crosby 6M6 valves with cold loop seals. The inlet fluid conditions used in the analysis were based on Zion FSAR analysis results instead of the data given in Reference 7 for a 4-loop Westinghouse reference plant. The peak pressure at the safety valve inlet was assumed to be 2532 psia and the pressurization rate was assumed to be 80 psi /s. These values are lower than the predicted peak pressure of 2555 psia and 21
pressurization rate of 144 psi /s given in Reference 7. The loop seal temperature was assumed to be 1150F at the safety valve inlet and 6680F at the upstream end of the loop seal.
The analysis used a safety valve flow area of 0.025 ft2 and a valve -
discharge coefficient of 0.8 s. These modeling assumptions resulted in a steady state flow rate of 465,480 lbm/h which corresponds to 111% of the safety valve rated flow of 420,000 lbm/h. Therefore, the ASME Code requiremer.t f or 90% derating of the safety valve was acccunted for acceptably in the analysis. Node spacing was kept small enough to prevent underestimating the piping forces due to numerical smearing. A maximum time step of 0.001 s was used in the analysis. This maximum time step would seem to be to large for the nodalization used downstream of the safety valves where volumes less than 0.5 ft long were used. However, Ceco stated that the maximum time step input was chosen based on past experience and preliminary results of the maximum velocity in the piping. Therefore, the thne step chosen, and the thermal hydraulic analysis in general, are considered adequate.
4.4.2 Stress Analysis The structural analysis of the safety valve and PORV piping was performed using the finite element structural analysis program, ANSYS.
ANSYS is a large-scale, general purpose computer program for the static and dynamic analyses of linear and nonlinear structures. The Licensee stated that ANSYS was verified and quality assured for nuclear safety-related analyses, but did not provide de. tails on the verification of this program.
CE".o was requested to provide details on the verification of the program in Reference 15. In its reply, CECO stated that the attached consultant reports contained several verification problems pertinent to the analyses performed for Zion. However, review of these reports in Reference 16 found that they simply stated that ANSYS was verified and quality assured for -
nuclear safety related analyses. In general, more information on the verification work is needed to demonstrate a code's capability to analyze valve discharge transients. However, ANSYS is a widely known and accepted code for use is static and dynamic analysis of structures, and use of ANSYS 22
A to perform the structural analysis for pressurizer safety and relief valve piping is considered acceptable.
For the analysis of valve discharge conditions, the input forces on the piping system were obtained from the thermal hydraulic analysis results discussed in the previous section. The force time histories calculated from the REPIPE Code were applied to each pipe segment at the structural nodes where change of flow direction or cross section area occurred.
In Reference 16, CECO provided information on the key input parameters for the ANSYS analyses. CECO stated that ANSYS uses a direct integration !
method of solution. The time step used in the analyses was 0.0005 s. The structural model was reviewed and the number of piping elements used to represent the system is considered adequate. Damping of 2% was used.
The safety valve and PORV piping were evaluated against the requirements of the ASME Code,Section III, 1983 Edition with Addenda through Winter 1985. In the submittals provided by Ceco it was stated that the piping was only analyzed for the faulted condition which included deadweight, temperature, pressure, and safety valve discharge. This definition of the faulted load combination is not consistent with that provided by EPRI in Reference 10: normal operation load, earthquake (SSE),
and safety valve discharge loads. The most important difference is the lack of a seismic load in the CECO load combination. A review of the Zion FSAR indicated that at least the reactor coolant system was analyzed for a faulted load combination that included the SSE. Therefore, the load combination used by CECO is not consistent with either the Zion design basis or the EPRI recommendation.
When this load combination was evaluated using linear elastic methods, the results indicated several points in the piping system were severely overstressed. The piping system was then evaluated using nonlinear, inelastic methods. This analysis showed that all the components of the system (piping and supports) met applicable allowable stress or load limits except for those discussed below. Where the allowable stress or load was exceeded, the nonlinear, inelastic analysis showed the component was qualified except for one shock arrestor (see below). For seven elbows 23
, . , j where the elastic limit was exceeded, the maximum membrane plus bending
. strain was less than maximum allowable strain of 4%. Therefore, the elbows are cualified. All three of the fabricated elbows exceeded the elastic j limit. In this case, the maximum strain was less than the maximum allowable strain of 2%. This qualified the three fabricated elbows. Among the .
supports, nine of the shock arresters were subject to loads greater than their load limit. For eight of the arresters the maximum strain was less than the 2% membrane strain limit. The one shock arrester that exceeded the 2% membrane strain limit, RCRS-1120, was removed from the structural model and the analysis rerun. The analysis showed the integrity of the piping system was maintained with the arrester removed. Because the arrester is still physically a part of the system, this shock arrester should be examined to verify the need for replacement shoula the cafety valves lif t.
According to results of EPRI tests performed on the Crosby 6M6 safety valve, high frequency pressure oscillations of 170-260 Hz occurred in the piping upstream of the safety valve as loop seal water passed through the valve. This raises a concern that these oscillations could potentially excite high frequency vibration modes in the inlet piping that could contribute to higher bending moments in the piping. This phenomenon was not accounted for in the structural analysis of the system. The piping between the pressurizer and safety valves in the EPRI tests, however, was composed of 8-in. Schedule 160 and 6-in. Schedule XX while that at Zion is 6-in.
Schedule 160. Since the test piping did not sustain any discernible damage during pressure esci11ations occurring in the tests, it is expected that the plant piping also would not incur damage during similar oscillations. Thus,
, a plant specific analysis for th,ese pressure oscillations is not necessary.
4.4.3 Piping and Support Summary The piping analysis presented by CECO in this submittal is not considered complete because the results of an analysis of the system for PORV discharge or combining valve discharge loads with seismic loads was not presented. Therefore, Items 3 and 8 of Section 1.2 were not met. Should the safety valves lift, support RCRS-1120 should be examined to verify the need for replacement.
24 /
- 5. EVALUATION
SUMMARY
1 1
The Licensee for Zion, Units 1 and 2, has not provided an. acceptable response to the requirements of NUREG-0737. Therefore, it has not been reconfirmed that the General Design Criteria 14, 15, and 30 of Appendix A'to 10 CFR 50 were met. The rationale for this conclusion is given-below.
5.1 NUREG-0737 Items Fully Resolved Based on the following information provided by the Licensee, the requirements of Item !!.D.1 of NUREG-0737 were partir.11y met (Items 1, 2, 4 to 6, and part of item 7 in Section 1.2).
The Licensee participated in the development and execution of an l acceptable Relief and Safety Valve Test Program designed to qualify the operability of 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 inlet conditions which by analysis bounded the most probable maximum forces expected from anticipated design basis events. The generic test results and piping analyses showed that the valves tested functioned correctly and safely for all relevant steam discharge events specified in the test program and that the pressure boundary component design criteria were not exceeded. Analysis and review of the test results and the Licensee's justifications indicated, except as discussed in Section 5.2, direct applicability of the prototypical valve and valve performances to the in-plant valves and systems intended to be covered by the generic test program.
Therefore, the prototypical tests and the successful performance of the valves and associated components demonstrated that this equipment was constructed in accordance with high quality standards (General Design Criterion No. 30).
l I
25 l
_ _ - _ _ _ - - - - _ _ _ _ _ _ _ _ - _ - \
z l
u e
5.2 NUREG-0737 Items Not Resolved Based on the Licensee's submittal, the following requirements of NUREG-0737, item II.D.1, as shwn in Section 1.2, were not met. l Item 3: Item 3, which requires the forces on the safety and relief valves be maximized, was not met. This because the Licensee did not provide the results of an analysis of the system for PORV discharge or combine valve discharge loads with seismic loads. Therefore, it could not be concluded the forces were maximized.
Item 7' : That part of item 7 that requires consideration of the effect of as-built discharge piping on safety valve and PORV operability was not met. This is because the pressure drop when the safety valves open is larger than the corresponding values for the test valves. This suggests the Zion 1&2 safety valves may not perform stably. Also, the maximum expected bending moment on the Zion 1&2 PORVs was not supplied. Thus, operability of the p0RVs with the maximum expected applied moment could not be assured.
Item 7: Item ,, regarding applicability of the test valves, was not met for the block valves. The test results on the valve / operator combination tested by EPRI are not applicable to the block valve / operator combination at Zion, Units 1 and 2. This is because, based on the information provided by the Licensee, it cannot be concluded the torque output of the plant operators is greater than the minimum torque used in the EpRI tests.
Item 8: Item 8, which requires qualification of the piping and supports, was not met. This is because the Licensee's piping analysis did not the analyze the system for PORV discharge or combine valve discharge loads with seismic loads.
Two items require the Licensee to develop formal procedures that need to be incorporated into the plant operating procedures or licensing documents such as the plant technical specifications. The ten results demonstrated the need for inspection and maintenance of the safety valves 26
+ ~
- .s following each lift involving loop seal or water discharge. Also, should the safety valves lift, support RCRS-1120 needs to be examined to verify the need for replacement. Based on the resolution to Item 8 above, additional work may be required to qualify RCRS-1120 as well as other piping and supports.
Therefore, the Licensee has not demonstrated by testing and analysis that the reactor primary coolant pressure boundary will have a low probability of abnormal leakage (General Design Criterion No. 14) and that the reactor primary coolant pressure boundary and its associated components (piping, valves, and supports) were designed with sufficient margin such that design conditions are not exceeded during relief / safety valve events (General Design Criterion No. 15).
27
i
., o .-
o
.. 9
- 6. REFERENCES
- 1. TMI Lessons Learned Task Force Status Report and Short-Term j Recommendations, NUREG-0578, July 1979. i l
- 2. Clarification of TMI Action Plan Requirements, NUREG-0737, November .
1980.
q
- 3. D. P. Hoffman, Consumers Power Co., letter 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, January 1983.
- 6. EPRI PWR Safety and Relief Valve Test Program Test Condition Justification Report, EPRI NP-2460, January 1983.
- 7. Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Westinghouse-Designed Plants, EPRI NP-2296, January 1983.
B. EPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Report, EPRI NP-2628-SR, December 1982.
- 9. R. C. Youngdahl, Consumers Power Co., letter to H. Denton, NRC,
" Submittal of PWR Valve Data Package," June 1, 1982.
' Valve Test Program Results to Plant-Specific Evaluations, Revision 2, Interim Report, July 1982,
- 11. E. D. Swartz, Commonwealth Edison Co., letter to D, G. Eisenhut, NRC,
" Zion Station, Units 1 and 2, NUREG 0737, Item II.D.1, Plant Specific Submittal," July 1, 1982.
- 12. S. A. Varga, NRC, letter to D. L. Farrar, Commonwealth Edison Co.,
" Request for Additional Inf6rmation, NUREG-0737. Item II.D.1, Performance Testing of Relief and Safety Valves, Zion, Units 1 and 2,"
February 19, 1985.
- 13. P. C. LeB1cnd, Commonwealth Edison Co., letter to H. R Denton, NRC,
" Zion Nuclear Power Station, Units 1 and 2, NUREG-0737, Item II.D.1,"
June 18, 1985.
- 14. P. C. LeBlond, Commonwealth Edison Co., letter to H. R. Denton, NRC,
" Zion Nuclear Power Station Units 1 and 2, NUREG-0737, Item II.D.1,"
January 22, 1986.
- 15. J. A. Norris, NRC, letter to D. L. Farrar. Commonwealth Edison Co.,
" Request for Additional Information, NUREG-0737, Item II.D.1, Performance Testing of Relief and Safety Valves, Zion, Units 1 and 2,"
March 31, 1987.
/
28
,s A
- ci
, ~
- 16. P. C. LeBlond, Commonwealth Edison Co., letter to USNRC Document Control Desk, " Zion Nuclear Power Station Units 1 and 2, NUREG-0737, Item II.D.1," November 19, 1987.
17.
F. G. Lentine, Commonwealth Edison Co., letter to H. R. Denton, Zion Equipment Qualification Report, May 19, 1983.
- 18. Application of RELAPS/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads, EPRI-2479, December 1982.
- 19. C. Y. Yuan, C. L. Nalezny, and C. P. Fineman, Technical Evaluation Report, TMI Action--NUREG-0737 (II.D.1). Donald C. Cook Units 1 and 2, Docket NO. 50-3 Q and 50-316, EGG-NTA-7881, October 1987.
29
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