ML19241A882
| ML19241A882 | |
| Person / Time | |
|---|---|
| Issue date: | 05/18/1979 |
| From: | Stolzenberg M NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | Bennett G NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| FOIA-79-98, REF-GTECI-A-09, REF-GTECI-SY, TASK-A-09, TASK-A-9, TASK-OR NUDOCS 7907110157 | |
| Download: ML19241A882 (10) | |
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NUCLEAR REGULATORY COMMisslON f, g WASHINGTON, D, C. 20S55 c,, v j EY13 1979 i
i MEMORANDUM FOR: Gary L. Bennett, Chief Research Support Branch l
FROM:
M. D. Stolzenberg l
Research Support Branch
SUBJECT:
SAFETY AND RELIEF VALVES At the request of. D. F. Ross, NRR, this memorandum and enclosures, suranarize infonnation available regarding relief and safety valve flow discharge (water and two-phase), operation, and testing. This memorandum completes EDO-TMI Action #18, as amended by the May 7, 1979 telephone request from D. F. Ross to you.
RELIEF / SAFETY VALVE FLOW I
To support NRR's evaluation of the flow through the safety valves en the B&W and GE plants under ATWS conditions RES contracted with ETEC (Energy Technology Engineering Center) to conduct a literature search to deternine if there were any addit {onal data generated in the interim between the 1975 report on ATWS and 1978.
The final report, Study of Safety Relief Valve Ooeration Under ATWS Conditions (NUREG/CR-068) was completed in January 1979 and puollsned in Marcn 1979, thereby completing the first phase of the research requested in Reference 1.
I Memorandum from E. G. Case to S. Levine,
Subject:
Request for Confirmatory Research Related to the Behavior of the Pressurizer Safety / Relief Valves During Subcooled Discharges (RR-NRR-78-10),
dated May 10, 1978.
2 Anon., " Status on Anticipated Transients Without Scram for Westinghouse Reactors," Appendix 0, Division of Systems Safety, U.S. Nuclear Regula-tory Ccmmission,, December 9,1975.
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Gary L. Bennett MAY l 3 979 In summary, the ETEC report states:
" Flow capacity data was not found in the literature or industry survey for safety relief valves of any size operating under ATWS conditions.
Considerable data was found on nozzles and other devices at pressures to 2500 psia with various degrees of subccoling. This data included in Table 1 (attached to this memo)... illustrates the tendency of the IHE model* to underpredict (flow is greater than predicted) flow. The underprediction is due to metastable (superheated liquid) flow condi-tions. Certain conditions may, however, cause the IHE model to over-predict: dissolved gases which may prevent metastable flow conditions, subcooling in. excess of 100*F where discontinuity of the sonic velocity occurs at the saturated liquid line, and the nozzle geometry where increased diameter reduces the critical mass flux."
3 This conclusion is further substantiated in a separate BNL report where it is stated for pipe breakflow:
... both the hemogeneous-equilibrium model (HEM) and the Moody model underpredict the critical flow rate data considerably."
i ETEC is currently working on a supplemental report which will include consideration of supercritical behavior. Contact has been made with staff members at Westinghouse, Combustion Eng%eering and General Electric.
Westinghouse supplied the following documents which have been informally transmitted to NRR:
1.
" Flow of a Flashing Mixture of Water and Steam Through Pipes and Valves,"
by W. F. Allen, Jr., Transactions of the ASME (April,1951).
2.
" Flow of Subccoled Water Through Nozzles," by A. W. Powell, WAPD-PT(V)-90 (April 12,1961).
3.
" Capacity Rating of Nozzle Type Safety Valves on Saturated Hot Water," by F. A. Custer and O. E. Buxton (January 2,1971).
1 IHE = Isentropic hcmogeneous equilibrium model (ASME Coda,Section III).
3 Pradip Saha, A Review of Two-Phase Steam-Water Critical Flow Models with Emohasis on Thermal Nonecullibrium, USNRC Feport NUREG/CR-0417 (Septemoer 12).
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Gary L. Bennett ~,'
3 iS73 I
It is my understanding that C-E has informed you that the ETEC report did not include Reference 3 (probably due to an overlap in publication) and a U.K. report "A Study of the Critical Flow Models Used in Reactor Blowdown Analysis." It is also my understanding that C-E may have a valve testing capability soon.
1 Contact has also been made with representatives of Philadelph.ia Electric which operates a fossil-fueled super-critical power plant (Eddystone).
Unfortunately, the safety valves were not tested for flow rate in the plant.
Discussions with W. S. Farmer pointed out tnat Pratt and Whitney planned super-critical netzle tests but these were never performed.
QA TC:/9AEFTY VAO/E FATU!RE 4
A recent NRR report cites over 100 examples of failure of pressure relief and safety relief valves, and indicates that " operating experience with spring leaded sr.fety valves has been essentially failure free." The significance here is that the pressure relief and safety relief valves are of the pilot operated type while the safety valves are spring loaded and sel f-acti 4ated.
The Nuclear Safety Information Center will soon publish a bibliographic report on valve behavior.
FOREIGN RESEARCH The following sections smanarize programs in France, Germany (FRG) and Japan on relief and safety valves with comments.
Federal Recublic of Germany There is a research program on valves being conducted by KWU (Kraftwerk Union, Erlangen, FRG) called " Investigation on the Operational Reliability of the Pressurizer Safety Valves and Relief Valves During Blowdown of Hot Pressurized Water."
U.S. Nuclear Regulatory Commission, Technical Recort on Ooeratinc Exoerience with BWR Pressure Relief Systems, NUREG-0462 (July 1978).
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t Gary L. Bennett...
'"a I I 3 1979 l
The intent of this research program is to demonstrate experimentally 1
and analytically, the capability of the relief valve to meet ATWS conditions. This research effort will involve tests with saturated steam, hot pressurized water and the transition from saturated steam to hot pressurized water.
Testing will take place in two facilities, one in Karlstein and one in Erlangen. The testing will be limited by pressure and flow capabilities of the facilities. These tests are to investigate:
I 1.
performance of pilot and primary (main) valve, i
2.
interplay between pilot and primary valve, i
3.
effect of piping layout and operation.
1 The pilot valve has had some testing completed at Erlangen using saturated steam and hot pressurized water. These tests were conducted with the pilot valve connected to a primary valve but without any fluid connection to the primary valve. The valve performed as expected with saturated steam, but did not open as smoothly and completely with hot pressurized i
water. In additien the pressure gradient used (10 bars /sec) is higher
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than the required gradient (4 bars /sec). Additional tests will be run at the lower gradient.
The tests at Karlstein will include a blowdown through the primary valve but at about 1/10 the design flow rate.
l The remaining tests are expected to be run in the latter part of.1979.
i Comment: These tests are similar to what may be required for an NRC ATWS safety valve test but they don't go far enough; that is, the pressure is considerably below what we would require (3200 psi). There has been no attempt so far in the FRG research program to measure flow rates. Moreover, the FRG researchers are testing pilot-operated valves while the valven of interest to us are of the self-activated spring type. However, the valve being tested appears to be very similar to the relief valve- (RC-R-2) that did not reseat at TMI-2.
Additional details on the FRG research will be sent shortly by W. B. Corwin, an CRNL staff member stationed in FRG.
(See enclosed letter for more information on the current status). We can expect to have limited access to data because these tests appear to be privately funded and may not be covered by our research exchange agreement with the FRG Ministry of Research and Technology.
3 } 7.
Gary 1.. Bennett LQ73
.s France Researchers with the French EdF intend to test a 6-inch self-activated spring-loaded safety valveQe testing is to be conducted with hot pressurized water pressurized by'M.ea(in a facility near Paris.
The initial steady-state tests were conducted with the valve mechanically held open. The dynamic tests, with the valve opened by system pressure, are to be conducted from September-December 1979.
(The French CEA said they sent NRC information on valve tests but the documents. seem to have.
gotten lost within NRC).
Coment: These tests could contribute considerably to the data we are looking for to confirm the method of analysis of the flow through a safety valve under ATWS conditions. The pressures being used are lower than our ATWS conditions and the flows expected may not match, but the general conditions, size and type of valve appear to be very close to what we are looking for. Since we have no exchange agreement with EdF we must await completion of EPRI negotiations with EdF.
Jacan The Japanese have a multi-purpose facility called the Nuclear Power Engineering Test Center for testing BWR and PWR nuclear plant components.
The testing for valves is categorized _as follows:
1.
accident environment simulatior, tests, 2.
operation and leakage tests, 3.
safety and relief valve tests.
The program on valves which is called " Reliability Demonstration Tests for Nuclear Plant Valve" includes testing of safety and relief valves to confirm the functioning of the valves, service life, leakage and valve coefficients. Testing has begun and will continue through 1981, but the valves scheduled for the test all appear to be for SWRs.
Coment: This program is very broad and covers many facets of safety and relief valves. With the exception of pressure the facility apoears to have the capability to provide data wnich may be applicable to cur ATWS problem. There is no indication, as yet, that they are planning that type of test.
It is not clear that our exchange agreement with Japan covers this type of testing.
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Gary L. Bennett Enclosed for your use is a sumary table of the world-wide test facilities known to me. At present there are no U.S. facilities capable of performing ATWS-type tests; although both the ETEC and Wyle facilities could be rnodified for such testing. A sumary of the ETEC capabilities is enclosed.
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w M. D. Stolze erg Research Support Branch
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Division of Reactor Safety Research
Enclosures:
4, as stated cc w/ encl:
S. Levine T. E. Murley R. J. Mattson V. Stello D. F. Rcss L. S. Tong L. Shao S. H. Hanauer R. Vollmer G. D. McPherson E. Brown H. Ornstein A. Thadani C. Graves T. Novak L. S. Rubenstein G. Mazetis R. M. Scroggins A. W. Serkiz S. Fabic L. M. Shotkin J. Bates, ETEC E. S. Hutmacher, ETEC A. H. Spano W. B. Cottrell, CRNL/NSIC W. Corwin J. Cicerchia, C-E L. E. Hochreiter, W W. 0'Ardenne, GE ~
S. Banworth, B&W W. Loewenstein, EPRI p2 07 8
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