ML20235M302
| ML20235M302 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 06/30/1987 |
| From: | Nalezny C, Pace N EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | |
| Shared Package | |
| ML20235M281 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-7739, NUDOCS 8707170145 | |
| Download: ML20235M302 (28) | |
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INFORMAL REPORT D<
TECHNICAL EVALUATION REPORT
. ' Jui TMI ACTION--NUREG-0737 (II.D.1) o
/dah0F f F RELIEF AND SAFETY VALVE TESTING i
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DISCLAIMER i
l This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of.their employees, makes'any' warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, of any information, apparatus, product or' process disclosed.in this report or represents that its use by such third party would l
not infringe privately owned rights.
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1 EGG-NTA-7739 i
TECHNICAL EVALUATION REPORT TMI ACT10N--NUREG-0737 (II.D.1)
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RELIEF AND SAFETY VALVE TESTING i
CRYSTAL RIVER UNIT 3 o
DOCKET NO. 50-302
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N. E. Pace C. L. Nalezny 3
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June 1987 1
Idaho National Engineering Laboratory EG&G Idaho, Inc.
l Prepared for the U.S. Nuclear Regulatory Commission Washington, D. C. 20555 Under DOE Contract No. DE-AC07-76ID01570 FIN No. A6492 i
ABSTRACT Light water reactors have experienced a number of occurrences of improper performance of safety and relief valves installed in the primary coolant system.
As a result, che authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions.
This report documents the review of these programs by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc.
Specifically, this report documents the review of the Crystal River Unit 3 Licensee response to the requirements of NUREG-0578 and NUREG-0737.
This review found the Licensee has not provided an acceptable response, which would reconfirm that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 have been met.
1 FIN No. A6492--Evaluation of OR Licensing Actions-NUREG-0737, II.D.1
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CONTENTS j
ABSTRACT..............................................................
ii 1.
INTRODUCTION.....................................................
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l 1.1 Background.................................................
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1 1.2 General Design Criteria and NUREG Requirements.............
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2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM................
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3.
PLANT SPECIFIC SUBMITTAL.........................................
6 4.
REVIEW AND EVALUATION............................................
7 4.1 Valves Tested..............................................
7 4.2 Test Conditions............................................
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I 4.3 Operability................................................
11 4.4 Piping and Support Evaluation..............................
15 5.
EVALUATION
SUMMARY
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REFERENCES.......................................................
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I TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (II.D.1) RELIEF AND SAFETY VALVE TESTING CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302 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 have been instances of valves opening below set 1
pressure, valves opening above set pressure, and valves failing to open or i
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 because of basic unreliability of the valve design.
It is known that the 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 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
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General Design Criteria 14, 15, and 30 require that (1) the reactor primary coolant pressure boundary be designed, fabricated, and tested so as
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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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l 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, which was issued for implementation on October 31, 1980.
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As stated in the NUREG reports, each pressurized water reactor Licensee or Applicant shall:
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1.
Conduct testing to qualify reactor coolant system relief and l
safety valves under expected operating conditions for design basis j
transients and accidents.
2.
Determine valve expected operating conditions through the use of analyses of accidents and anticipated operational occurrences i
referenced in Regulatory Guide 1.70, Rev. 2.
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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 2
are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (FSAR). The effect f
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
.I analysis.
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2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM In response to the NUREG requirements previously listed, a group of utilities with PWRs requested the assistance of the Electric Power Research Institute (EPRI) in developing and implementing a generic test program for pressurizer safety valves, power operated relief valves, block valves, and associated piping systems.
Florida Power Corp. (FPC), the owner of Crystal River Unit 3 (CR-3), was one of the utilities sponsoring the EPRI Valve Test Program. The results of the program, which are contained in a series of reports, were transmitted to the NRC by Reference 3.
The applicability of
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these reports is discussed below.
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1 EPRI developed a plan (Reference 4) for testing PWR safety, relief, and block valves under 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-and representative valves were selected for testing. These valves included a sufficient number of the variable characteristics so.that their testing would adequately demonstrate the performance of the valves used by utilities (Reference 5).
EPRI, through the Nuclear Steam Supply System (NSSS) vendors, evaluated the FSARs of the participating utilities and' arrived at a test matrix which bounded the plant transients for which over pressure protection would be required (Reference 6).
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EPRI contracted with Babcock & Wilcox (B&W) to produce a report on the inlet fluid conditions for pressurizer safety and relief. valves in B&W designed plants (Reference 7).
Since CR-3 was designed by B&W, this report is relevant to this evaluation.
q 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 relief valves (SRVs) were tested at the Combustion Engineering Company, Kressinger 4
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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 j/alve capacity data, (2) assess hydraulic and structural effect. of associates piping on valve operability, and (3) obtain piping responss data that could ultimately be used for verifying analytical piping models.
l 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 FPC on July 01, 1981 (Reference 10) and August 7, 1981 (Reference 11).
Additional assessment of the Pressurizer i
Safety and Relief Valve Piping was transmitted March 31, 1982 (Reference 12), June 30, 1982 (Reference 13) and November 1, 1982 (Reference 14).
A request for additional information (Reference 15) was submitted to FPC by the NRC on October 18, 1984.
FPC responded to this request on February 17, 1986 (Reference 16).
The response of the overpressure protection system to Anticipated Transients Without Scram (ATWS) and the operation of the system during feed 1
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 CR-3 utilizes two safety valves, one PORV, and one block valve in the overpressure protection system.
Both safety valves are Dresser Model 31739A. The PORV is a Dresser Model 31533VX-30.
The block-valve is a 2-1/2 in. Velan bolted bonnet gate valve with a Limitorque SMB-00-10 operator. There are no loop seals between the pressurizer and the safety
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valves or the PORV Also, each valve is connected separately with the.
quench tank.
The Dresser 31739A safety valve used at CR-3 was one of the valves tested by EPRI; therefore the EPRI test results are directly applicable to the CR-3 safety valves.
The Dresser PORV installed at CR-3 has dash 2 internals (31533VX-30-2) and has a bore diameter of 1-5/32 in.
The test valve was also a dash 2 design but 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 difference in bore diameter will only affect capacity and not operability..
R The test valve is, therefore, considered an adequate representation of the in plant valve.
The Velan block valve used at CR-3 is a 2 1/2 in, bolted bonnet gate valve 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' Limitorque operator SMB-000-10.
FPC, in their submittal (Reference 16) compared the CR-3 block valve with both the EPRI test valves and TMI-2 block valve and stated that the CR-3 block valve is not appreciably different from the valves tested at Marshall Station and is similar to the TMI-2 block valve.
It is believed that the CR-3 block valve will operate as required because both the EPRI test valve with the SMB-00-10 operator and the TMI-2 valve with the same operator' performed satisfactorily.
In addition, the 7
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i CR-3 block valve operated satisfactorily during the February 26, 1980 transient. The 3 in. EPRI test valve requires a larger force to operate and the SMB-000-10 operator is a 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.
Based on the above and assuming acceptable answers to the four questions asked, the valves tested are considered to be applicable to the in-plant valves at CR-3 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 I
The valve inlet fluid conditions that bounc 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 those 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 J
report that are applicable to CR-3 are those identified for B&W 177-FA plants.
For the safety valves, 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 17, the maximum backpressure developed during FSAR accidents and transients for CR-3 is 520 psia. Therefore the test on a Dresser 31739A valve with a peak backpressure of 617 psia. bounds the maximum backpressure predicted to occur at CR-3.
Since CR-3 does not have loop seals upstream of the safety valves, testing of the Dresser safety valves with the short inlet piping is applicable.
Eight applicable steam tests (Tests 316, 318, 320, 322, 324, 326, 328, and 1104a) with a short inlet pipe were performed with the 31739A valve which had a peak pressure of 2720 psia and a peak pressurization rate of 333 psi /sec.
The ring settings for these tests were (-48, -40, + 11 and -48, 8
L_______-_________
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-60, + 11) as compared to (-48, -50, +11) for.the CR-3 valves.
All but Test 320 performed satisfactorily (flow > 100% of rated, opened within +3%
of set pressure, and reclosed with less than 20% blWdown). Test 320 did not reach rated lift; it had an 866 psia back pressure.
The highest back pressure for the other seven tests was 676 psia. These conditions bound those expected at CR-3.
For extended HPI events (which include feedwater line breaks and steam line breaks) the safety valves will initially open 'on steam with transition to subcooled ' water calculated. A peak pressure of 2515-psia was calculated with liquid temperatures ranging from 400 F to 6400F.
A peak liquid 0
surge' rate of 11,520 lbm/ min (at 6400F) will occur.
Pressurization rates from 0 to 65 psi /sec are expected.
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For the 31739A valve, testing included a steam to water transition test at 2489 psia and saturated conditions. Three water tests at pressures ranging from 2329 to 2749 psia and with water temperatures of 4140F to 6080F were run. Durihg thes tests, the 31739A valve passed at least 1128 GPM (-8,000 lbm/ min) with 5390F water, and 2492 GPM 0
(-16,000 lb/ min) with 649 F.
The transition and water tests 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 CR-3..
These conditions are sufficiently close to the conservatively selected bounding conditions to adequately demonstrate valve performance.
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.
Additionally, 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 adequate.
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 backpressure at least as high as those expected in service should be demonstrated. The expected backpressure for the PORV was not reported by 9
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FPC. However, the PORV discharge pipe routing is similar to the safety valves. The PORV rated flow, 100,000 lb/hr, is <30% of the rated flow of l
the safety valve, 317,973 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 cenclusion is reached that the expected backpressure for the PORV is less than the 520 psia which bounds the safety valve. Testing of the valve (Reference 8) included numerous steam tests with opening pressures close to the CR-3 set pressure and back pressures as high as 760 psia which adequately bounds the expected conditions for the PORV.
I For extended HPI events (which include feedwater line breaks and steam line breaks) the initial opening of the PORV will be on 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 pressures of l
approximately 2500 psia were included in the test series. The tests were l
run using the same discharge pipe orifice which developed backpressure l
l ranging from 175 to 415 psia for the steam tests so that the expected backpressure was adequately represented.
The HPI events are, therefore, considered to have been adequately represented by the tests.
The PORV is used for low temperature overpressure protection (LTOP).
For LTOP events, the valve is required to open on 565 psia steam.
Reference 7 indicates transition and water flow will not occur at CR-3 l
during low temperature overpressurization events. Opening on steam is considered to be adequately represented by the full pressure steam tests discussed above.
l For the block valve only full pressure steam, 2480 psia, tests were perf'ormed (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 10
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1 demonstrate operability of the valve for low pressure steam and the required water conditions.
The TMI-2 valve is similar to the CR-3 valve and it functioned satisfactorily during the TMI-2 accident.
In addition, the CR-3 block valve operated satisfactorily during the February 26, 1980 event which is a positive demonstration of this valve's operability.
The test sequences and analyses described above demonstrating that the test conditions bounded the conditions for the plant valves plus the
' operating experience of the block valve,' verify that 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 I
met.
4.3 Valve Operability lhe CR-3 safety valves (Dresser 31739A) were tested by EPRI and the' test conditions enveloped the expected CR-3 valve conditions as discussed Section 4.2.
The valve ring setting used (-48, -50, + 11) falls between the two ring settings used in the eight short inlet pipe tests and are considered bounding and acceptable. Seven of those eight tests (all but Test 320) functioned acceptably; Test 320 had an excessively high back l
pressure (866 psia) and the valve did not reacfi rated lift or flow rate.
Since Test 320 back pressure is significantly higher than the CR-3 expected back pressure, this test is considered outside the required operation envelope for CR-3 and its resulting failure not pertinent.
Blowdowns for the eight Dresser 31739A safety valves tested by EPRI ranged from 7.0% to 16.9% so that the measured blowdown generally exceeded the design blowdown of 5%.
A B&W analysis (Reference 18) has shown blowdown up to 20% does not impede natural circulation due to hot leg voiding.
Therefore, having the observed blowdown exceed the design blowdown is considered acceptable.
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l The maximum bending moments applied to the discharge flanges of the Dresser 31739A test valves during the eight applicable tests was l
230,913 in.-lbs.. Valve operability was not impaired by the application of these moments. The maximum moment computed by FPC for the CR-3 safety valve is 25,099 in.-lbs (Reference 16). However, this moment does not include seismic loading, or all the FSAR transient conditions.
Therefore,.the maximum expected moment on the plant valve may not be bounded.
For the test performance to be a valid demonstration of plant safety valve stability, the test inlet' piping must have a pressure difference at least as great es the plant.
The plant valves are mounted directly on a pressurizer nozzle and thus have the minimum pressure drop possible, and should be as stable as the test valve.
During the 4140F water test (Test 1114) the 31739A valve was stable but only achieved partial lift. The valve did not pass enough flow to prevent the test pressure from accumulating.
However, the amount of liquid that the valve discharged was more than the amount predicted to be discharged during a steam line break at 4000F.
In addition, there are two safety valves at the plant, which gives CR-3 more than sufficient relief capacity. Under conditions typical of the FWLB, 2515 psia, water flow at temperatures of 6020F and 6400F, the 31739A test valve on the short inlet configuration passed the required flow in two out of three tests, and over 80% during the third test. CR-3 has sufficient relief capacity at these conditions, because two valves are installed at the plant.
Based on the test results discussed above, demonstration of safety valve operability is considered adequate.
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The Dresser PORV opened and closed on demand for all nonloop seal testi.
Inspection of the valve after testing at the Marshall Steam Station showed the bellows had several welds partially fail. The failure did not l
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 12
25,500 in.-lb was induced on the discharge flange of the test valve without impairing operability. The maximum bending moment calculated for the CR-3 PORV is 15,552 in.-lbs (Reference 16).
However, this moment does not include seismic loading, or all the FSAR transient conditions.
Therefore, the maximum expected moment on the plant valve may not be bounded.
The CR-3 PORV is a pilot operated valve that uses system pressure to hold the disk tight against the seat. At one point Dresser Industries
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recommended the block valve be closed at system pressures below 1000 psig to avoid steam wirecutting of the PORV disk and seat. Testing by Dresser later showed the 1000 psig pressure limit to be overly conservative and that the PORV as designed was qualified to system pressures of 50 psig and below.
Below 50 psig Dresser recommends that the PORV block valve be closed to prevent leaking.
In addition, Dresser provides heavier springs to be used under the main and pilot disks to ensure closure if the plant is to operate below 50 psig.
However, leakage is not a problem at all plants, and the plant start up procedures for Cryatal River 3 (Paragraph 6.4.6.12) require that the valve be cycled twice at 205-215 psig to ensure its proper operability.
Failure of the PORV to operate properly will force the plant to remain at that pressure until a decision is made on how to restore the valve to its proper condition (Reference 16).
The valve performance during EPRI tests, under the full range of expected inlet conditions, and the CR-3 startup procedures, demonstrate that the PORV is capable of discharging the required steam and liquid flow rates.
However, since the maximum expected moment under all possible transients was not provided, PORV operability was not adequately demonstrated since valve performance might be reduced by excessive moments of the inlet and/or outlet flanges.
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 TMI-2 experience and the tests with the 3 in. Velan valve and SMB-000-10 operator conservatively 13
demonstrate the opevoility of the plant valve provided CR-3 block valve torque setting use the srme or higher.
The test valve was cycled successfully at full steam pressure with full flow.
In addition, the plant valve operated satisfactorily during a plant transient.
Based on the performance of the test valve, and the plant valve, the CR-3 PORV block valve is considered operable.
NUREG-0737 11.3.1 requires qualification of associated control
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circuitry as part of the safety / relief valve qualification.
In Reference 15, the NRC requested information demonstrating that the PORV control circuitry is qualified.
In Reference 16, FPC stated that qualification of the PORV and its control circuitry is not required by NRC regulation 10 CFR 50.49, and further stated that they did not believe that the NRC intended the PORV circuitry to be qualified under NUREG-0730 Item II.D.1.
The FPC response is considered unsatisfactory because an electrical system malfunction initiated the February 26, 1980 transient that challenged the CR-3 high pressure injection system and a safety valve (Reference 16) which indicates that the PORVs are subject to spurious actuations which can challenge the plant safety sustems.
In addition, the response to question 2 in Reference 16 clearly states that the limiting inlet conditions for the PORV include extended HPI operation following an FSAR steam line break (which would expose the PORV to a harsh environment I
during which it could malfunction and cause additional challenges to plant safety systems).
On the basis of the submittals by FPC, the PORV control circuitry is not consi:lered qualified, and therefore does not satisfy the requirements of NUREG-0737 Item II.D.1.
The presentation above, demonstrates that the valves operated satisfactorily, verifing that the portion of Item 1 of Section 1.2 that requires conducting tests to qualify the valves has been met.
However, that part of Item 7 requiring that the effect of discharge piping on operability be considered, and Item 5 requiring qualification of the PORV control circuitry have not been met.
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4.5 Piping and Support Evaluation l
l In the piping and support evaluation, the safety / relief valve piping and supports between the valve discharge flanges and the pressurizer relief tank were analyzed for the requirements of the ANSI B31.1 Power Piping Code, l
1967 Edition with code case N-7.
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The thermal-hydraulic analysis was performed with the program RELAP4/M005. The THRUST code was used to generate fluid force histories from the RELAP4/M005 output.
RELAP4/ MOD 5 has been been benchmarked against l
the output of RELAPS/M001 in Reference 16, and has been shown to provide j
I satisfactory thermal Hydraulic results.
Furthermore, the ability of RELAP5/ MOD 1 to calculate the system thermal-hydraulic response has been 1
verified through simulations of EPRI/CE tests (Reference 21).
Verification
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of the TRUST code was also provided in Reference 16. Therefore it can be concluded that RELAP4/M005 and the THRUST code will produce acceptable calculations of piping loads due to safety and relief valve discharge.
The Licensee stated that the thermal hydraulic analysis for each individual valve and associated discharge piping was performed separately.
Since the discharge piping of the safety valves and PORV are not headered together, the effects of simultaneous actuation of the valves is not an important consideration in the analyses for CR-3.
7 The bounding transient conditions for which the piping and supports must be qualified should have been based on those summarized in Reference 7.
In Reference 16, the licensee provided additional detail on the bounding transients.
The safety valves are required to pass steam, transition from steam to saturated water, and subcooled water.
The bounding transients are:
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Steam line break:
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Safety valve set point = 2500 psig.
Steam flow, transition to saturated water, followed by subcooled water.
Liquid temperature range = 5170F to 6020F.
Liquid insurge into pressurizer = 6555 lb/ min at 6020F, and 6019 lb/ min at 5170F.
Maximum steam discharge = 366,350 lb/hr (based on EPRI tests)
Rod ejection accident at HZP:
Safety valve set point = 2575psig.
l Saturated steam discharge.
l Maximum pressurizer pressure = 2662 psig.
Pressurization rate = 175 psi /s.
The bounding transients involving the PORV are presented in Reference 7.
They include FSAR transients, and low temperature overpressurization events that result in steam flow, transition from steam to saturated water, and subcooled water. The bounding conditions for the analyses are; j
FSAR steam transients; l
PORV set point = 2240 psig Maximum pressure = 2662 psig The forces generated from these conditions bound those from all other conditions expected at the plant.
In Reference 16, the licensee stated that the safety valve piping was evaluated using the inlet conditions for the February 20, 1980 transient, because subcooled water discharges are normally controlling due to the much higher discharge rate through the valve.
The inlet conditions used in the analysis are:
16
Safety valve opening pressure = 2410 psig Pressurization rate = 0, pressurizer maintained at 2410 psig Safety valve inlet temperature = 5600F (subcooled water)
Liquid discharge rate = 1,544,280 lb/hr Safety valve opening time = 0.040 s l
Based on the information presented in Reference 21, it is not clear that the liquid discharge analysed provides the bounding loads on the piping and supports.
Figure 2 in Reference 21 shows design forces computed in 1974 that are as high as those computed in 1980, with no explanation as to why the 1980 calculation are actually bounding.
In addition, EPRI test data indicates that the safety valve opening times ranged from 0.007 to 0.043 s.
Therefore, it can not be concluded that the bounding transients that provide the maximum expected loads on the piping and supports due to a safety valve discharge have been considered.
i In Reference 16, the licensee stated that the thermal hydraulic analysis for the PORV discharge piping was based on the following design conditions.
l Pressurizer pressure = 2315 psia Pressurizer temperature = saturated steam at 2315 psia PORV flow rate = 117,000 lb/hr Pressurizer pressure = constant 2315 psia.
Subsequent to the analysis it was determined that the PORV flow rate is 165,900 lb/hr. However, the analysis was not redone, on the assumption that the piping loads are proportional to flow rate.
Again, few details were provided by FPC on the calculations performed.
It did not appear that a liquid discharge case was analyzed for the PORV piping, which according to FPC would produce the maximum piping and supports loads.
17 i
The piping structural analysis was performed using the computer code
)
PIPDYN II. This code was originally developed by the Franklin Institute.
The licensee did not provide verification of this code to the EPRI test I
1 data. It was stated that the code has been verified to an alternative structural problem with a published solution; however, data comparing the results of these analyses were not provided.
The licensee stated that the results of the piping load cases that were analyzed were combined by absolute summation. This is generally more conservative than the SRSS combination method recoramended by EPRI for combining transient and seismic l
l loads. 'The stress information provided implies that the results will envelope the load cases recommended by EPRI in Reference 19.
- However, complete information concerning piping modeling information and results for all recommended EPRI load cases was not provided so structural adequacy for all load cases could not be confirmed. The applied moments on the valves l
{
due to the transient load case analyzed by the licensee were well within the
)
values applied to similar valves during the EPRI tests.
However, it is not clear that these are bounding moments.
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The information supplied indicates only that the piping supports were evaluated.
It was stated that two supports would require modification to withstand the loads imposed during the 2-26-80 transient (which is used as the limiting transient).
However, since this load was not considered to be l
a design condition and since no physical damage was apparent after the 2-26-80 event, no modifications were made.
No detailed information was provided regarding analytical methods utilized for supports analyses. The licensee has not demonstrated support structural adequacy for all load cases recommended by EPRI.
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18 L
5.
EVALUATION
SUMMARY
The licensee for Crystal River Unit 3 has not provided an acceptable response to the requirements of NUREG-0737, which would reconfirm 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 i
conclusion is given below.
l The licensee participated in the development and execution of an i
acceptable relief and safety valve test program to qualify the operability l
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 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 currectly and safely for all steam and water discharge events specified in the test program that were applicable to Crystal River Unit 3 and that the pressure boundary component design criteria were not exceeded.
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.
However, the plant specific piping has not been shown by analysis to be acceptable.
In addition, the effect of in plant piping on operability of the plant safety valves and PORV has not been adequately considered.
The following requirements of Item II.D.1 of NUREG-0737 have been met:
Items 1-4, and 6 in Paragraph 1.2.
These requirements ensure that the reactor primary coolant pressure boundary components (safety valves, PORV and block valve) will have a low probability of abnormal leakage (General Design Criterion No. 14).
However, the reactor primary coolant pressure boundary and its associated components (piping, and supports) have not been shown by analysis to have been designed with a sufficient margin so that design conditions are not exceeded during relief / safety valve events (General Design Criterion No. 15).
19 l
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The prototypical tests and the successful performance of the valves and
{
associated components demonstrated that this equipment was constructed in I
)
accordance with high quality standards, meeting General Desigr. Criterion i
No. 30.
The PORV control circuitry, piping and supports have not been shown to satisfy General Design Crition No. 30.
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20
6.
REFERENCES 1.
TMI-Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578, July 1979.
2.
Clarification of TMI Action Plan Requirements, NUREG-0737, November 1980.
3.
Letter, R. C. Youngdahl, Consumers Power Co., to H. D. Denton, NRC,
" Submittal of PWR Valve Test Report, EPRI NP-2628-SR," July 1, 1981.
4.
EPRI Plan for Performance Testing of PWR Safety and Relief Valves, July 1980.
l 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 Reoort, EPRI NP-2628-SR, December 1982.
9.
EPRI/ Marshall Electric Motor Operated Block Valve Test Data Report, EPRI NP-2514-LD, July 1982.
- 10. Letter, P. Y. Baynard, FPC, to Harold R. Denton, NRC, "CR-3 Safety and Relief Valve Testing Program," 3-071-01, July 1, 1981.
11 Letter, W. A. Cross, FPC, to Darrell G. Eisenhut, NRC, "CR-3 Safety and Relief Valve Test Requirements," 3-081-05, August 7, 1981.
12.
Letter, David G. Mardis, FPC, to Darrell G. Eisenhut, NRC, "CR-3 Safety and Relief Valve Testing Program," 3F-0382-33, March 31, 1982.
13.
Letter, P. Y. Baynard, FPC, to Darrell G. Eisenhut, NRC, "CR-3 Pressurizer Safety / Relief Valve Operability", 3F-0682-23, June 30, 1982.
14.
Letter, Patsy Y. Baynard, FPC, to Darrell G. Eisenhut, NRC, "CR-3 Safety and Relief Valve Testing Program," 3F-1182-01, November 1, 1982.
l 15.
Letter, John F. Stolz, NRC, to W. S. Wilgus, FPC, " Request for Additional Information on Relief and Safety Valves Testing per NUREG-0737, Item II.D.1., August 26, 1985.
16.
Letter, G. R. Westafer, FPC, to John F. Stolz, NRC, "CR-3 Pressurizer Relief and Safety Valves," 3F-0286-09, February 17, 1986.
1 l
21 b
I
17.
" 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.
q l
- 18. Pressurizer Safety Valve Maximum Allowable Blowdown, B&W Report l
l 77-113-5671-00, August 1982.
l l
19.
EPRI PWR Safety and Relief Valve Test Program Guide for Application of I'
Valve Test Program Results to Plant-Specific Evaluations, Revision 2, Interim Report, July 1982.
1 l
- 20. Application of RELAP5/M001 for Calculation'of Safety and Relief Valve j
Discharge Piping Hydrodynamic Loads, EPRI-2479, December 1982.
- 21. Letter P. Y. Baynard FPC, to R. W. Reid, NRC,'" Crystal River Unit 3, NRC letter dated May 2, 1980, Requesting Information Concerning Pressurizer Safety Valves," July 17, 1980.
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g,voMm E',"328-BIBLIOGRAPHIC DATA SHEET EGG-NTA-7739 SEE INSTRUCT 10N5 0N TDet REVtH58 2 TITLE AND 5U9 f sTLE J LE Avt OLahn Technical Evaluation Report, TMI Action--NUREG-0737 (II.D.1) Relief and Safety Valve Testing, Crystal a oAre aa oar CoMauteo i
River, Unit 3 j
..A.
,,0N 1,.
June 1987 i
it Avr,,0,,,,,
. oar..tro T,ou.o N. E. Pace C. L. Nalezny June l 1987 MONrN veaa 7 v't A*0HMING ORGAmi2ATION NAME AND MaaLiNG ADont65 sinesu.sslse Casiiss S PROJtC7/T ASK/WQRK WNIT NUM8tm INEL-EG&G Idaho, Inc.
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Idaho Falls, ID 83415 A6492
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10 SPON50ReNG ORGANIZ A IION NAME AND MalLING ADDMESS (sarsesseles Comes 11a. TYPt OF REPOR T Mechnical Engineering Branch Informal Office of Nuclear Regulatory Commission s..oo Cov e..o,,,,,,,,. e Washington, DC 20555 12 SUPPLEMtNTAR Y NOTE $
i3 A35f se ACT (200 =orse er ess#
l Light water reactors have experienced a number of occurrences of improper performance of safety and relief valves installed in the primary coolant system.
l 1
As a result, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status l
Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification l
of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional perfonnance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would, e
i l
verify the integrity of the piping system for normal, transient, and accident l
conditions. This report documents the review of these programs by the Nuclear j
l Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc.
Specifically, l
l this report documents the review of the Crystal River Unit 3 Licensee response to l
the requirements of NUREG-0578 and NUREG-0737.
This review found the Licensee has not provided an acceptable response, which would reconfirm that the General Design Criteria 14,15, and 30 of Appendix A t'o 10 CFR 50 have been met.
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