ML20005B990
| ML20005B990 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Calvert Cliffs, Oconee, Palisades, Saint Lucie, Crystal River, Midland, Fort Calhoun, Crane  |
| Issue date: | 09/14/1981 |
| From: | Cutchin J NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Atomic Safety and Licensing Board Panel |
| Shared Package | |
| ML19291D364 | List: |
| References | |
| NUDOCS 8109160287 | |
| Download: ML20005B990 (13) | |
Text
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STAFF 9/14/81
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S b[I,,f f UNITED STATES OF Ai! ERICA NUCLEAR REGULATORY COMMISSION A
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BEFORE TH'E ATOMIC SAF(TY AND LICENSING BOARD 15BMm3l N "Wm7 $
9 pf)N In the Matter of
)
1, N
METROPOLITAN EDISON COMPANY, ET AL.
Docket No. 50-289
)
(Restart)
(Three Mile Island, Unit 1)
)
NRC STAFF'S RESPONSE TO LICENSING BOARD'S ORDER TO NRC STAFF 0F AUGUST 25, 1981 On August 25, 1981 the Licensing Board issued its " Order to NRC Staff Regarding Board Notification of Unsatisfactory Test Results of Safety Valve."
In that Order the Board indicated that it had become aware, 3ir a board notification that was filed in another proceeding,M of some unsatisfactory test results for a safety valve of the type installed at THI-1.
Not having received such a notification in the captioned proceeding, the Board requested the Staff to inform it promptly whether notification of this matter by the Staff would have been appropriate in this proceeding, and if not why not.
~
Also, the board directed the Staff to explain the significance of the unsatisfactory safety valve test results in the context of the proposed findings and issues in this proceeding.
The Board expressed a particular ir.terest in the effect, if any, of these test results on the Staff's position that the PORV and associated block valve are not required to mitigate the consequences of any design basis accidents because the pressurizer rafety valves provide the required protection.
l y
HRC Board Notification No. 81-20, dated August 11, 1980, that was 0,
filed in the McGuire proceeding.
p 8109160287 810914
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. i The fiRC Staff's response to the Licensing Board's Order is set forth in two documents:'
1.
The "!iRC Staff's Report on Board's Comments Regarding Board riotification of Unsatisfactory Test Results of Safety Valves" that was prepared by John F. Stoltz and Dominic C. Dilanni.
2.
The "t4RC Staff's Report to the Board on Safety Aspects of EPRI Test Data on Relief and Safety Valves" that was prepared by Edgar G. Hemminger and Walton L. Jensen, Jr.
Copies of those documents and their attachments and copies of the affidavits of Messrs. Stoltz, Dilanni, Heminger and Jensen are enclosed.
Respectfully submitted, Y
^
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~
James M. Cutchin, IV Counsel for f4RC Staff Dated at Bethesda Maryland this 14th day of September,1981 L
1 UNITED STATES OF AMERICA NUCLEAR REGULATORY C0ff4ISSION BEFORE THE AT0li!C SAFETY AND LICENSING BOARD In the Matter of HETROPOLITAii EDIS0N COMPANY, Docket No. 50-289 ET AL.
(Restart)
(Three flile Island, Unit 1)
CERTIFICATE OF SERVICE I hereby certify that copies of "NRC STAFF'S RESPONSE TO LICENSING BOARD'S ORDER TO NRC STAFF 0F AUGUST 25, 1981" in the above-captioned proceeding has been served on the following by deposit in the United States mail, first class or, as indicated by an asterisk, by deposit in the Nuclear Regulatory Commission's internal mail system or, as indicated by a double asterisk, by hand-delivery, this 14th day of September,1981:
- Ivan W. Smith, Esq., Administrative Ms. Marjorie M. 'Aamodt Judge R.D. #5 Atomic Safety & Licensing Board Panel Coatesville, PA 19320 U.S. Nuclear Regulatory Commission Washington, D.C.
20S55 Mr. Thomas Gerusky Bureau of Radiation Protection
~
- Dr. Walter H. Jordan, Administrative Dapt. of Environmental Resources Judge P.O. Box 2063 881 W. Outer Drive Harrisburg, Pennsylvania 17120 Oak Ridge, Tennessee 37830 Mr. Marvin I. Lewis Dr. Linda W. Little, Administrative 6504 Bradford Terrace Jufge Philadelphia, Pennsylvania 19149 5000 Hermitage Drive Raleigh, North Carolina 27612 Metropolitan Edison Company ATTN:
J.G. Herbein, Vice President, George F. Trowbridge, Esq.
P.O. Box 542 Shaw, Pittman, Potts & Trowbridge Reading, Pennsylvania 19603 1800 M Street, N.W.
Washington, D.C.
20006 Ms. Jane Lee R.D. 3; Box 3521 Karin W. Carter, Esq.
Etters, Pennsylvania 17319 505 Executive House P. O. Box 2357 Walter W. Cohen, Consumer Advocate Harrisburg, Pennsylvania 17120 Department of Justice Strawberry Square,14th Floor Honorable Mark Cohen Harrisburg, Pennsylvania 17127 512 D-3 Main Capital Building Harrisburg, Pennsylvania 17120
e-l i
Thomas J. Germine Deputy Attorney General Division of Law - Room 316 1100 Raymond Boulevard Newark, New Jersey 07102 Allen R. Carter, Chairman John Levin Esq.
Joint Legislative Committee on Energy Pennsylvania public Utilities Conn.
Post Office Box 142 Box 3265 Suite 513 Harrisburg, Pennsylvania 17120 Senate Gressette Building Columbia, South Carolina 29202 Jordan D. Cunningham, Esq.
Fox, Farr and Cunningham Robert Q. Pollard 2320 North 2nd Street 609 Montpelier Street Harrisburg, Pennsylvania 17110 Baltimore, Maryland 21218 Louise Bradford Chauncey Kepford Three Mile Island Alert Judith H. Johnsrud 315 Peffer Street Environmental Coalition on Nuclear Power Harrisburg, Pennsylvania 17102 433 Orlando Avenue State College, Pennsylvania 16801 lis. Ellyn R. Weiss Harmon & Weiss f
?!s. Fr ieda Berryhill, Chairman 1725 I Street, N.W.
Coalition for Nuclear Power Plant Suite 506 Postponement Washington, D.C.
20006 2610 Grendon Drive Wilmington, Delaware 19808 Mr. Steven C. Sholly
(?nion of Concerned Scientists Gail P. Bradford 1725 I Street, N.W.
ANGRY Suite 601 245 W. Philadelphia St.
Washington, D.C.
20006 York, Pennsylvania 17401
- Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.
20555
- Atomic Safety and Licensing Board Panel q
U.S. Nuclear Regulatory Commission s
Washington, D.C.
20555 gmu.,
WW James M. Cutchin, IV
- Secretary Counsel for NRC Staff U.S. Nuclear Regulatory Commission ATTH: Chief. Docketing & Service Br.
Washington, D.C.
20555 William S. Jordan, III, Esq.
Harmon & Weiss 1725 I Street, N.W.
Suite 506 Washington, D.C.
20006 i
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UNITED STATES OF N* ERICA NUCLEAR REGULATORY C0!Vl!SS10N BEFORE THE ATOMIC SAFETY AND LICENS!HG BOARD In the matter of
)
)
METROPOLITAN EDISON CO., ET AL.
)
Docket No. 50-289 (Three Mile Island Nuclear
)
Station, Unit 1)
)
(Restart)
NRC STAFF'S REPORT TO THE BOARD ON SAFETY ASPECTS OF EPRI TEST DATA ON RELIEF AND SAFETY YALVES By order dated August 25, 1981, the Board directed the staff to explain the significance of unsatisfactory safety valve test results in the context of the proposed findings and issues in this proceeding. The Board is particularly interested in the effect of the test results on the staff's position regarding the PORY and associated block valve.
In a letter dated November 26, 1980 from R. H. Vollmer (NRR) to R. C. Youngdahl (EPRI), the Office of Nuclear Reactor Regulation (NRR) provided commer.ts and requested additional information regarding EPRI's
" Proposed Program Plan for the Performance Testing of PWR Safety and Relief Valws", Revision 1, dated July 1,1980. In that letter, we requested that tbr PWR Owners make " provision for expeditious transmittal of test results from the PWR Owners to the NRC as individual valve tests are completed" so that we could continuously monitor the progress of the test program. The mechanism agreed to for regular transmittal of results is the EPRI Weekly Report. The report is usually issued on Friday and includes c summary of tests conducted at the various test facilities for the week from the previous Monday through the date of the report. One such report is the one dated June 26,1981 referred to in the Board's August 25, 1981 Order to the Staff.
1
[
.-l 3
2 The utilities with assistance from the NSSS vendors have the primary responsibility for evaluating the safety significance of a given test result for their specific plant They are responsible under the regulations to advise NRC if information obtained from the test program reveals an unreviewed safety question for their plant. NRR with assistence from RES and its con-sultant, EG8G, is reviewing and evaluating each reported test result for potential generic safety significance. The NRC and consultant personnel reviewing the test results are familiar with the basic valve types being tested, a general knowledge of valve and related piping installations in PWR plants and a knowledge of the conservatisms used to design PWR Overpressure Protection Systems. Actions to be taken based on a review of test results that fail a test screening criterion range from consideration of relevance and materiality for Board notification to shut down of plants. An example of a test result with obvious safety significance would be failure of a safety valve to open during a given test sequence. As stated in SECY-81-491 dated August 17,1981 (attached) although some test screening criteria have not been met, the testing to date has not uncovered problems with safety or l
relief valves which are considered significant to the safety of operating plants. This same conclusion is applicable to the TMI-l restart.
In response to the Board's August 25, 1981 Order TMI-1 plant specific evaluation of the significance of the EPRI test results to date is as follows.
For Dresser relief valves (PORVs) of the type installed at TMI-1, the reported preliminary test results indicate that although the test acceptance criterion were not met for water seal type installations, the PORV's will function i'n the primary mode (pressure relief) as required. The test results to date indicate that the Dresser PORV's experienced a delay of as much as
- f.
. )
i 70 seconds in closing time due to low or ambient water seal temperatures.
The valves closed on their own, however, and on disassembly and inspection no damage was obsened which might affect their ability to open or close on demand. These results do not indicate a safety concern with respect to TMI since the TMI plant specific piping does not contain water seals for the PORV's, and since all test results applicable to non-water seal piping configurations were satisfactory for the Dresser PORY.
F6r Dresser safety valves of the type installed at TMI-1, the pre-liminary test results indicate a need for additional information regarding the effects of inlet piping configuration, back pressure, and adjusting ring settings on safety valve operation. The test acceptance criteria with respect to flow capacity or stem position were not met for certain predetermined test conditions. Based on the worst case preliminary data point, a maximum stem position of 65% was obsen ed for a high ramp rate, high back pressure steam test with the valve set to the original manu-facturer reconinended ring settings.
If it were assumed that the TMI-1 installed safety valves were limited l
to the worst case stem position of 65%, a consenative estimate of approx-imately 405,000 #/hr. relieving capacity would be available. This estimate is based on the consenative assumption that percent flow is approximately equal to the percent stem position.
Sen.sitivity studies of the required safety valve flow capacities for design basis transients as described in topical report, BAW-10043, " Overpressure Protection for Babcock & Wilcox Pressurized Water Reactors", dated May 1972, indicate that a maximum total safety valve flow-capacity of 345,000 f/hr. is required. We, therefore, l
l
(
~ - - -.
r i
_4 conclude that sufficient safety valve relieving capacity is available at TMI-1, even based on the worst case preliminary EPRI test date and taking no credit for the,'100,000 #/hr. relieving capacity available through the PORY. The staff testimony of Jensen and staff proposed findings on the PORV and block valve are, therefore, unchanged.
It should be noted that the EPRI test data as reported on a weekly basis is preliminary in nature.
In general, no conclusions can be made on valve performance based on preliminary, individual test results.
It is neither expected nor desirable for Ltilities to be making adjustments
{
to their safety valves until all testing under all conditions has been completed with the results fully evaluated against plant spectfic configu-rations since all test results are not necessarily applicable to all reactor plants. The safety valve test data as reported to date includes only results of steam testing. The subcooled liquid and transition flow tests have not yet been performed.
August 17, 1981 SECY-81-4Sl For:
The Commissioners From:
William J. Dircks Executive Director for Operations
Subject:
REVISED SCHEDULE FOR COMPLETION OF TMI ACTION PLAN ITEM II.D.1, RELIEF AND SAFETY VALVE TESTING
Purpose:
To revise NUREG-0737 to extend the schedule for submittal of the" subject PWR valve test program results from October 1,1981 until July 1,1982 Discussion:
By letter dated December 17, 1979, Mr. William J.
Cahill, Jr., then Chaiman of the EPRI Safety and Analysis Task Force, submitted to the NRC " Program Plan for the Perfomance Verification of PWR Safety /
Relief Yalves and Systems". This proposed test program was in response to the requirements specified in NUREG 0578, "TMI-2 Lessons Learned Task Force Status Report and Short Term Recommendations". Item 2.1.2, "Perfomance Testing for BWR and PWR Relief and Safety Valves". Revision 1 of the program plan for PWR safety and relief valve tests was submitted by the industry to NRC on July 8,1980, in response to NUREG 0737. In addition, there have been several meetings during this time between the PWR utility representatives, EPRI staff and their const:ltants and NRC staff, to provide additional clarification of the EPRI/PWR safety and relief valve test program.
The sta'f reviewed both the initial and revised test l
descriptions and was in agreement that the technical requirements sf NUREG 0578 and NUREG 0737 would be met on satisfactory completion of testing. However, the p.oposed test schedule was felt by the staff to be optimistic in that it provided no margin for contingencies.
By letter dated July 1,1981, from R. C. Youngdahl l
to Harold R. Denton, enclosure 1, the PWR Owners l
Group reoorted on the status of the EPRI PWR safety
Contact:
E. Hemminger, DE, NRR Ext. 29481
~
r The Commissioners.:
and relief valve test program to date and requested
~
an extension of the completion dates specified in NOREG-0737. The Owners Group stated their intention to develop an expanded test matrix in order to obtain more information with respect to the effects of inlet piping configurations and adjustments of ring settings on safety valve operation.
On July 17, 1981, the staff met with EPRI and the PWR Owners Group representatives to review the status of l
the safety and relief valve testing and to discuss
(
the expaWi test matrix. Although the exact number of addittenal tests will have to be determined as the i
program progresses, the test progrtn managers estimated that it could take from four to eight months longer than the original test completion date of July 1,1981, to complete the expanded test program.
Test Program and Status The program plan developed by EPRI is an extensive testing and analysis effort costing in excess of i
l
$17 million. Three test facilities were designated I
for testing of ten relief valves and nine safety valves.
The f'llities are located at Marshall Steam Station (Duke Power Company), Wyle Laboratories (Norco, California), and Combustion Engineering (Windsor, Connecticut).
The tast facilities at Marshall Steam Station and Wyle Laboratories have been in full operation since l
mid-1980 and have provided a substantial quantity of information on relief valve (PORV) perfomance. *he PORY test results are summarized in section 4.0 of the "EPRI/PWR Safety and Relief Valve Test Program Interim Data Report", dated July 1,1981, (enclosure 2). High pressure steam testing is reported as complete on all ten PORVs, and high pressure water.
loop seal simulation, and transition steam to water tests are reported as complete on four of the ten PORVs.
The test results for each specific valve are forwarded to utilities that are known to have these valves installed or intended for use in their facilities for purposes of perfoming any required safety eval-uation.
In addition, NRR, with assistance from RES and our contractor EG&G, has teen evaluating the l
PORV test results on a weekly basis. The reported test results indicate that, while the initial i
The Comissioners
- screening criteria were not met in some cases, all p0.RVs tested will function in the primary mode (pressure relief) as required. Additional PORY tests are being planned to evaluate the effect of variable water seal temperature on valve closure timas. The test results to date indicate that some valves experience a delay of as much as 70 seconds in closing time due to low or ambient water seal tenperatures. However, the valves closed on their own and on disassembly and inspection no damage was observed which might affect their ability to i
open or close on demand. These results do not indicate a significant safety concern in the staff's view.
The testing of safety valves to meet the NRC require-ments has necessitated the design and construction of a new facility at Combustion Engineering. This facility is the first of a kind with the capability to perfonn meaningful operability tests for large spring-loaded safety valves over a broad range of fluid inlet conditions. Although extraordinary effort, including three shift-work schedules, was l
devoted to this part of the testing program, delays in construction and shakedown testing resulted in i
significant delay in the safety valve test schedule.
As a result, test results for only two of the nine safety valves to be tested are available (enclosure 2). These test results indicated a need for addi-tional information regarding the effects of inlet piping configuration and adjusting ring settings on safety valve operation. Reporting of safety valve test results and review by affected utilities and the staff is on the same basis as for the PORY results.
Based on our review of the EPRI test program to date; we have concluded tht;. the program represents a fully responsive effort to meet Commission require-ments and that the additional testing proposed will
(
provide needed infomation to assure that the technical requirements of item II.D.1 of NUREG-0737 will be met. Since testing to date has not uncovered problems with safety and relief valves which are considered significart to the safety of operating plants, we believe that good cause has been shown i
to extend the NUREG-0737 completion date for PORV l
and safety valve testing so that the extended EPRI program may be carried to completion on an orderly basis. The latest estimated test completion date is March 31, 1982.
- ~
- 1; 1
The Comn.issioners :
A proposed general letter (enc 1cture 3) will advise al.1 licennes, applicants, and coestruction perinit holders of the revised schedule.
Recommendation:
That the Comission approve a revised schedule for completion of the PWR (EPHI) valve test program.
4 It should be noted that:
a.
The BWR valve test program is not affected by the recommended change.
b.
The change does not impose any additional reporting requirements.
Scheduling:
For early consideration.
[}
b'j William J. Dircks Executive Director for Operations Enclosures :
1.
Ltr. from R. Youngdahl to H. Denton dated July 1,1981.
2.
"EPRI/PWR Safety and Relief Valve Test Program Interim Data Report" 3.
Proposed letter to all licensees l
l i
Comissioners' comments should be pinvided directly to the Office of the Secretary by c.o.b. Tuesday, September 1,1981.
Comission Staff Office coments, if any, should be submitted to the Comissioners NLT August 25, 1981, with an information copy to the Office of the Secretary.
If the paper is of such a nature that it requires additional time for analytical review and coment, the Comissior,ers and the Secretariat should be apprised of when coments may be expected.
I DISTRIBUTION Comissioners Commission Staff Offices Exec Dir for Operations Exec Legal Director ACRS ASLEP Secretariat e
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General offices: 212 West Michtgen Avenge, Jackson, MI 49201 * (017) 788 4600 July 1, 1981 Mr Barold R Denton Director, Nuclear Reactor Regulation U S Nuclear Regulatory Co==ission Washington, DC 20555 STA~1iS OF EPRI PWR SAFETY AND RELIEF VALVE TEST PROGRAM NUREG-0737, ITEM II.D.1 In Dece=ber,1979 forty-one utilities
- vith planned or operating pressurized water reactors co==itted to be responsive to the reco==endations of NUREG-0578, Section 2.1.2 and de=enstrate the capability of safety and relief valves to operate satisfactorily under expected operating and e.ccident conditions.
By 1ctter dated July 8,1980 Revision 1 of the EPRI " Program Plan for the Perfor=ance Testing of PWR Safety Relief Valves" was sub=itted to the NRC.
Tnis revision addresses Ites II.D.1. A of NUREG-0737, which prcctided NRC clari-fications to the earlier NUPIG reco==endations.
The progra: plan developed by EPRI for the participating PWR utilities is an extensive testing and analysis effort which is utilizing three test facilitiec and will cost in excess of $20 million. The progra= has been " success" oriented with very little contingency time or funds to resolve potential problems.
Al-though the program has been very successful and preliminary results-to-date indicate that the valves tested vill perform their intended safety function, more infor=ation appears needed in selected areas. Additional tests, outside the July,1980 Plan test matrix, are being performed. These additional tests of both safety and relief valves have been infor= ally diccussed with the NRC staff. The principal area requiring more testing and evaluation of relief valves is the i= pact of variable loop seal te=perature on the valve operation.
Revisions to the safety valve test matrix are necessary to obtain a better understanding of upstream pipe / valve interaction. The impact on the overall test schedule is provided in Attach =ent 1.
By previous agree =ent (R C Youngdahl letter to D G Eisenhut, dated December 15, 1960) the PWR utilities agreed to sub=it the attached Interim Data Report.
This report provides all preliminary data collected through June 19, 1981.
Additional quick look data reports and weekly activities reports vill continue to be provided to the NBC staff until all testing is co=pleted. The PWR utili-ties still intend to.heet the co=mitment dates provided in the Dece=ber 15, 1980 letter except that the final data report v411 not be provided by October 1, 1981.
- Six external organizations have since agreed to participate in the EPRI program (Co=bustion Engineering, Framatome, Central Nuclear de Almaraz, Furnas Electricas, Electronucleair and Swedish State Power Board).
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$ 1.0 7 0 7 01 S 1
Mr Harold R Denton USERC o7/1/81 2
Separate fro = the safety and relief; valve test progra: NUREG-0737, Ite: II.D.1.B requested that utilities provide verification of block valve functionability.
During earlier =eetings with the NRC staff, the utilities participating in the EPRI valve progra= concluded that emphasis =ust be placed on the de=onstration of safety and relief valve dperability but that EPRI would be requested to develop a block valve task action plan.
The FWR utilities have reviewed a pro-posed action plan and are nov prepared to discuss the need, depth and schedule of a possible block valve progra=.
While it is recognized that the schedules to satisfy the reco==endations of NURIG-0737, Ite: II.D.1 are not totally consistent with the NRC's request, EPRI and the FWR utilities have instituted a progra= that is providing new scientific supportable data about valve operability which is not avsilable fro: any other i
source.
The utility advisory groups coordinating the test progra= and EPRI are prepar.*4 to =eet with the NRC staff to discuss the status of EPRI progra= in more detail.
I propose to =eet with you and'your staff on July 16 or 17, 1981.
.I
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ENCLOSURE 2 l
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9 COPY NU:;SER
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0 EPRI/P*n'R SAF IEF VALVE TEST' PROGRAM ORT INTERI?
A JULY i
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3.0 SUMP.ARY OF SAFETY VALVE OPERABILITY DATA A total of nine PWR pressurizer. safety, valve designs were tested under steam, water, stes:n to water (transition), and loop seal conditions.
The nine safety valves selected for testing in the EPRI Program, and the i
safety valves represented by the valves tested, are identified in Section 2.0 of this report.
The purpose of this section is to present the conditions tested and principal observations for the safety valves tested as of June 19, 1981.
- Appendix A of this report contains detailed data sheets for these tests.
These data sheets completed after each test and are designed to be self sufficien+.
w timely dissemination of that safety valve test data deemed cas adequately evaluate valve performance'. Key information el these sheets ars valve designation, tested conditions, valve o i g and closing times, maximum stem position and
~
valve flow rates.
3.1 DRESSER SAFETY VALVE @
70PMA w
n 3.1.1 Conditions Tested v
Tests were perfomed o t resser safety valve model 31709NA l
at the EPRI/CE PWR Sa
- y n elief Valve Test Facility.
Table 3.1.1 presents t ma f conditions under which the l
Dresser valve was tested.
O 3.1. 2 Summary of Principal Observ r l
A full pressure steam test (te was performed on the Dresser safety valve, model 3170 ",
h test was performed with the valve mounted on a loop seal c
.,ur n with the loop seal drained and the valve set point est s
a 2480 psig. The test was initiated with a high ramp ratt t 5 n from the pre-test pres,sure inlet pressure of of 2315 psia. The safety valve opened t v.
2465 psia. The transient continued for me of 122 seconds.
The valve chattered during most of the tes urat'o. The valve re-closed at a. pressure of 2000 psia. Several mi.n'
'ter closur,e, the pressure noted valve re-opened for a second time. The seco i
by the loop operator was approximately 2150 he valve reclosed the second time at a slightly r' educed pressure. The valve was open for about 10 seconds and chattered during this time.
After the test, a leak test wac perfomed at an inlet pressure of about 2100 psia. The valve leakage measured was about 0.5 spm. The valve was then disassembled and a pre.iminary inspection was performed.
l Galling of guiding surfaces.was found; several internal parts were i
damaged.
Detailed data sheets are contained in Appendix A, Section A-1.
22
TABLE 3.1.1 VALVE 31709NA
_"AS TESTED" COMBilSTION ENGINEERING TEST MATRIX FOR Tile DRESSER TEST IHlET PIPING TRANSIENT CONDITIONS TEST. No.
TYPE CONFICllRATION INITIAL CONDITIONS VALVE PEAK VALVE OPENING PEAK DOWN-CLOSING PRESS TANK ANK STREAM TANK TEMP PRESS RATE PRESS ESS PRESS PRESS PSIA PSIA FLUID "F
PSIA _
PSI /SEC P_
201 Steam Ioop Seal Steam Sat.
2315 340-425 8
0 (1) 2010 (Dralned) n g
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(1) Measurement i
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DRESSER SAFETY VALVE MODEL 31739A 3.2
- 3. 2.1 Conditions Tested Tests were performed on the Dresser safety valve model 31739A at the EPRI/CE PWR Safety and Relief Valve Test Faci.11ty.
Table 3.2.1 presehts the matrix of conditions under which the Dresser valve was tested.
3.2.2 Summary of Principal Observations A full pressure, iow ramp rate, low backpressure, steam test
) was performed on the Dresser safety valve (test No. 3 valve opened at a pressure within +3% of the (31739A)
A maximum stem position of 58% of rated lift
. val ve - t t a pressure less than 6% above the valve set was
'n alve reclosed at a pressure greater than b
pressure.
2250 psig.
are contained in Appendix A. Section A-2..-
Detailed a
CROSBYHB-BP-86,3K6(
e.vCRpplication 3.3 3.3.1 Conditions Test
- 1ater --
3.3.2 Summary of Principal Ob er then h
- later -
CROSBY HB-Bp-86, 6M6 - Loop Seal App t
3.4 3.4.1 Conditions Tested
- later -
3.4.2 Summary of Principal Observations
- later -
g 9
24
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, TABLE 3.2.1 "AS TESTED" COMlUSTION ENGINEERING TEST MATRIX FOR Tile DRFSSER SAFETY VALVE 31? D A If.ST INIET PIPING if5T. NO.
TYPE CONFIGilRAll0N INITIAL CONDITIONS TRANSIENT CONDITIONS VALVE PEAK VALVE OPEtlING FAK DOWN-CLOSING PRESS TANK K
STREAM TANK PRESS RATE PRESS R S PRESS PRESS TEMP FLUID "F
PSIA PSI /SEC PS MA PSIA PSIA U
302 Steam Straight Steam Sat. 2300 3.75 2483 165 s2336 O
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3.5 CROSBY HB-BP-86, 6NG - Loop Seal Application _
- 3. 5.'s Conditions Tested _
- later -
3.5.1 Sur.ary of Principal Observations
- later -
CROSBY HB-BP-86, 6NS - Non-Loop Seal Aoplication 5.6
' 3.6.1 Conditions Tested
-1e 3.6.2 Su r.a / o r,cipal Observations
- late 3.7 CROSBY HB-BP-86, 6F 6 000 Seal Apolication 3.7.1 Conditions Tes
- later -
3.7.2 Summary of Frir.eipal ns v
- later -
tion 3.8 CROSBY HB-BP-86, 3K6 - Non-Loop S 3.8.1 Conditions Tested
- later -
3.8.2 Summary of Principal Observations Q
- later -
i 3.9 TARGET ROCK 69C
~
3.9.1 Conditions Tested
- later -
3.9.2 Summary of Princioal Observations
- later -
25
SUMMARY
OF RELIEF VALVE OPERABILITY DATA 4.0 The EPRI program calls for the testing of ten PWR pres simulation conditions.
The ten relief valves selected for testing in the EPRI P of this report.
The purpose of this section is to present the test matrices and principal 19, 1981. Appendix B observations of the relief valves tested as of June These data
' of this report contains detailed data sheets for these tests.
sheets are complet. after each test and are designed to be self-sufficient
. ination of that relief valve test data deemed to allow timely, necessary to
.e i
evaluate valve performance. Key information e s are valve designation, tested conditions, included or, "s
valve opening and..
g times and valve flow rates.
o 4.1 DRESSER RELIEF VALVE 9.
4.1.1 Conditions Test Q
tW esser relief valve model Tests were performe
' n and curing Phase II and Phasa III at the Marshall Stea.
s 4.1.la, b, and c present the of the Hyle Test Prog h)this valve model was tested at matrix of conditions und
.1 Marshall, Wyle (Phase II, a 1
hase III), respectively.
Summary of Principal Observatd 4.1. 2 v
Marshall Steam Station e
The valve fully opened on dema ly closed on demand, e
for each of the ten (10) evaluat' cycles. During the
've pilot stem.
evaluation tests, steam leaked pa i
to have several Upon valve disassembly, the bellow led with a new o
partially failed welds. The valve s
bellows and cycled 16 more times with aryi
. ilot back-.;
and closed pressures up to 900 psig. The valve fully eak. Upon id on demand for each cycle a td the bello isible cracks.
disassembly, the bellows did not have a In all test cases, the valve fully opened on demand and closed on demand even thouah the bellows was damaged during some tests.
~
Based on this input and the manufacturer's assessment of valve performance with the observed damage, thu damace was determined to have no potential impact on valve operction.
Detailed data sheets for the evaluation tests are contained in A'ppendix B, Section B-la.
27 c
~b V"
e Wyl e Phase II The valve fully opened on demand and fully closed on demand for each of the five '(5) test cycles.
Detailed data sheets are contained in Appendix B, Section B-1b.
e Wyle Phase III The valve fully opened on demand and fully closed on demand for nine (9) test cycles. The valve fully opened on demand and did not close on demand during the three (3) water seal simula!?on tests; numbers 16-DR-6W 22-DR-9W/W and 24-DR-6W.
Each tesA was a 2500 psia pressure test with low temperature 0
upstream of the valve followed by 650 F water.
water s
/
0
+ s( n'. er 16-DR-6W, the low temperature water was at 103 F.
D-in t est, the Dresser valve opened on demand. Upon the valve for closure, the valve remained open de-energ1 r until *.e J e was isolated from the test loop.
Following:
- ion, the valve closed. The valve was test v e
i m tel 40 seconds after it was signalled to isolated a a
emoved from the test facility and close.
- v i w
+e r ser representative. No damage *as disassemb1 observed whic.i.**
.ect the ability of the valve to open/close on I
/, the low temperature water was In tesc number 22-3210F. During the ee
- e valve opened on demand. Upon de-energizing the va fo osg e, the valve remained open for 2 seconds and then o d Vy.
f the tes* 16-DR-6W except Test number 24-DR-6W was
-p that the test was run to max.i h time before the valve was isolated. The water tem a
.c imediately upstream of j
the valve was 1050F. During t s
he valve opened on demand but failed to close ime n de-energizing th'e sol enoid. The valve closed on it a roximately 70 seconds, a
of approximately after the closure signal at an inl 2110 psia.
v was removed, After all tests were completed, the Dresse disassembled, and inspected. No damage '
osa d which might affect the ability of the valve to " i close on demand.
Detailed data sheets are contained in Appendix B, Section B-lc.
$ Ocenine Time The total valve opening time data for the Marshall Steam Station tests and the Wyle Phase II & III tests were obtained based cn different types of inputs. As a result, the recorded Marsnall opening times exceed the recorded '.lyle times for similar steam test conditions.
In addition, main disc ouening times of the valve could not be accurately determined at Wyle.
For that reason, the main disc opening time was not included on the Wyle phase II & I!! data sheets.
23 l
^
O G
TABLE 4.1.la "AS TESTED" MARSHALL TEST MATRIX FOR Tile DRESSER RELIEF VALVE
" NOMINAL" TEST INITIAL CONDITIONS "NOMTNAL" TEST NO.
TYPE AT VALVE INLET TRANSIENT CONDITIONS V VE i
TEMP TEST
.O IRE MAX DISCH.
1 PRESS DURATION P S PIPE B.P.
o FLUID F
PSIA (SEC)
PSIA j
1*
Steam Steam (Sat.)
2475 415 i
2-5 Steam Steam (Sat.)
2475 l
2335 415 l
6*
Steam Steam (Sat.)
2455 60 2325 175 7 - 10 Steam Steam (Sat.)
2455
\\
g5 2320 175 g
O
- Tests 1 and 6 were exte.
du to w measurement tests.
l'
~~
.l g
e
es r
t i
TABLE 4.1.ib "AS TESTED" WYLE PflA3E 11 TEST MATRIX FOR T;it DRt.SSER RELIEF VALVE,,
TRANSIENT CONDITIONS I
INITIAL CONDITIONS il SI No.
TIST AT VALVE INIET
"~
.T..Y. l..' E.
VALVE MAX l
TEST CLOSURF MAX DISCH.
PILOT LINE DURATION PRESS.
PIPE B.P.
BP 4
(SEC)
SIA)
PSIA PSIA l
ILUID
.'I, s60
- 104p, l
DR-1 5 STEAM STEAM 674 2490 510 155 21 3 e
DR 1.W WATER WATER 373 680 o
DR-5.H HATER WATiR 646 2500
+15 2300-380 680 O
'26 2' '
l nn..W mTER WATER 506 pR./W WATER WATER 2510
- 16 2120 373 333 I
.I l
.g, l
4
.e i
j I
r
1 TABLE 4.1.1c "AS TESTED' WYLE PHASE III TEST MATRIX FOR THE DRESSER RELIEF VALVE TEST TRANSIENT TEST NO.
TYPE INITIAL CONDITIONS _
CONDITIONS MAX MAX MAX (STATIC 4 DYNAMIC)
VA?VE DISCH. PILOT RENnING AT VALVE INLET IN ACCUHULATOR TEST CLO" RE PIPE LINE MOMENT TEMP PRESS.
T{HP. i'iiESS.
DURATION PR S BP RP IN0UCEn FLUID "F
PSIA FLUID F
PSIA SEC., _PSpi.
PSIA PSIA IN-LB O
10-DR-15 STEAM STEAM 668 2503 SAME AS VALVE INLET 755 830 N/A 11-DR-4W WATER WATER 647 2514 2338 620 740 N/A 12J R-3W llATER WATER 450 699 15 685 260 300 N/A 13-DR-7W WiTER WATER 451 2492 Q s10 652 420 450 NA g
14-DR-2W WATER WATER 112 689
+10 2230 s2 s2 N/A 15-DR-5W WATER WATER 643 2504 S10 2360 640 750 35,600 (preload)
V Sh R
R 652 2500
$54
$14.7 292 51 3 590,000 16-DR-6W WATER 103 l
SitiULATION-20-DR-15 STEAM STEAM 5
SAMF AS VALVE INLET s10 2110 494 760 N/A 21-DR-85/W TRANSI-STEAN 2496 WATER 641 ES03 s10 2360 660 770 N/A TION WATER 22-DR-911/W SEAL WATER 321 2490 WATER 647 2488
$17 2310 675 81 5 N/A SIMULATION
~
sli 2110 440 583 N/A 23-DR-15 STEAM STEAM 657 2505 SAME AS VALVE INLET WATER 24-DR-6W SEAL WATER 1 *]5 2505 WATER 650 2505 s85 2110 693 788 N/A SINULATION
4.2 CROSBY RELIEF VALVE
(
- 4. 2.1 Conditions Tested Tests were performed on the Crosby relief valve model at the Marshall Steam Station, and during Phase II and Phase III of tne Wy'e Test Program. Tables 4.2.la, b, and c present'the matrix of conditions under which this valve model was tested at Marshall,
~
Wyle (Phase II), and Wyle (Phase III), respectively.
4.2.2 Summary of princf oal Observations s
Marshall _ Steam Station The 11y opened on demand and fully closed on demand ia h of the ten (10) evaluation tests.
During va cycling performed prior to the evaluation tests under #
w steam conditions, the pilot bellows leaked.
s disassembled and inspected, one bellcws weld When v
fracture d and a bellows assembly part was four.d to be improcerl ed The bellows was repleced, the bellows assembly
- c' yA.achined and the valve was feassembled for further t V
The valve was s n*
cycled 44 times including the valve fully opened and closed on ten evaluation te i
demand and no bell s
=6 occurred during the tests.
ation tests are contained Detailed data sheets o a
in Appendix B, Section a
e Wyle phase II The valve fully opened on dema d
ly closed on demand for each o' the six (6) test c n disassembly after tests were completed, the pilot as found to leak.,.
Detailed data sheets are contained p
x B, Section B-2b.
I e
Wyle phase III 4
v-sed on demand The valve fully opened on. demand and fu for each of the ten (10) test cycles. Upon disassembly after
~
tests were completed, the pilot bellows was observed to be l
damaged.
> Bellows Damace In all test cases, the valve fully opened on demand and closed on demand even though the bellows had been damaged.
Based on this inout and the manufacturer's assessment of valve performance with the observed damage, the damage was determined to have no potential impact on valve operation.
- 2*
~
Openino Time The total valve opening time data for the Marshall Steam Station tests and the Wyle Phase II & III tests were obtained based on different types of inputs. As a result, the recorded Parshall opening times exceed the recorded Wyle times for similar steam test conditions. In addition, main disc opening times of the v.alve could not be accurately determined at Wyle.
For that reason, the main disc opening time was not included on the Wyle Phase II & III data sheets.
O O
C>
?
=
A 33
i
~
t TA81.E 4.2.la l
" AS TESTED" MARSilALL TEST MATRIX FOR Tile CROSBY REllEF VALVE I
I t
" NOMINAL" l
TEST INITIAL CONDITIONS
" NOMINAL" l
TEST NO.
lVPE AT VALVE INLET TRANSIENT CONDITIONS i
L TEST E MAX DISCH.
TEMP PRESS.
DURAL PIPE B.P.
I FLUID F
PSIA
,[5
(
PSIA o
j 2350 385 la steam Steam (Sat.)
2495
(
}
j 2-5 Steam Steam (Sat.)
2495 2340 380
,t 6*
Steam Steam (Sat.)
2495 1355 135 e
l O
s 80 7 - In Steam
, Steam (Sat.)
249 15 2335 120
- i. g I
i l
0
\\
9
- Tests I and 6 were exten duration flow measurement tests i
ll i
I
)
l4 v
)
I i
TABLE 4.2.1b
' AS TESTED" WYLE PHASE 11 TEST MATRIX FOR THE CROSBY REllEF VALVE.
f TRANSIENTCONblTICNS
. TEST NO.
TEST INITIAL CONDITIONS TYPE AT VALVE INLET FLUID fEMP PRESS TEST VALVE MAX MAX F
PSIA DURATION CLOSURE DISCHAR PILOT LIHE SEC PSIA ipr B.P.
l PI PSIA
_ s J-I i
i CR-1-S STEAM STEAM 672 2510 s15 1
142 945 o
u (1) 1000 1
l CR-2-S STEAM STEAM.
671 2495 2140 560 t
~
CR-3-W WATER WATER 376 68 618 244 200 4
CR-5-W WATER WATER 634 s15 2280 397 775 i
CR-6-W WATER WATER 02 s18 2100 460 438 i
550 661' CR-7-W WATER WATER 46 2510 s19 2000 p
i (1) The 1000 psia pressure sensor was ofer-ranged on this test.
e 4
4 TABLE 4.2.lc 2
"AS TESTED" WYLE PilASE 111 TEST MATRIX FOR Tile CROSBY RELIEF
- aLVE TRANSIENT IISI NO.
1[ST lYPE INITIAL CONDITIONS CONDITIONS MAXIMUM MAX MAX (STATIC +DYNAMI VALVE DISCHARGE PILOT BENDING AT val.VE INLET IN ACCUMULATOR TEST CLOSU E PIPE LINE MOMENT TEMP. PRESS.
-"-~T[HP PN'ESS.
DURATION PRES PRESS.
BP INDUCED El.tllD "F
PSIA FLillD F
PSIA (SEC)
PSIA PSIA PSIA IN-LB
( )N RECORDED 865 N/A 25-CR-IS STEAM STEAM 656 2'505 SAME AS VALVE INLET 10 5
26-CR-65 STEAM STEAM 657 2505 10 NOT RECORDED 868 38,400 (Pit [10All) fl 4
l 2 /- CR-211 WATER WATER 104 694 1
520 1.0 518 N/A i
28-CR-3W WAT ER WA1ER 437 695 655 160 540 N/A I
' 29-CR-IS STEAM STEAM 656 7505 10 2050 740 865 N/A 3ti-CR-IS STEAM STEAM 656 2505 10 2060 370 780 N/A 31-Cit-4S/W litANSI-STEAM 656 2510 10 15 2313 NOT RECORDED 770 N/A q
riON 32-Ell-Sil/W llATER WATER 469
,250 646 2505 15 2290 560 740 N/A SEAL SIMULATION 13-CR-7W/W WATER WATER 294 505 WATER 648 2505 15 2300 580 840 N/A SEAL.
SlHillATIUff 34-Cit-8ti/W WATER WATER 118 2500 WATER 645 2500 15 2290 570 700 N/A St AL SlHillATION g
mm
I E
4.3 TARGET ROCK RELIEF VALVE
/
- 4. 3.1 Conditions Tested
(
Tests were performed on the Target Rock relief valve model at the Marshall Steam Station and during Phase III of the Wyle Test Program. Tables '4.3.la and b present the matrix of. conditions under which this valve model was tested at Marshall and Wyle (Phase III), respectively.
4.0.?
Summary of Principal Observations fjarshall Steam Station e
'ully opened on demand and fully closed on demand The -
f e.h the ten (10) evaluation test cycles.
Detai a sheets are contained in Appendix B, Section B-3a.
o Wyle The valve f h pe on demand and fully closed on demand in eleven elve (12) test cycles. The valve did not close o i
$ the full pressure 2500 ps'1, water seal simulati e-est number 7-TR'-7W) was performed. The water just ups r or he valve was 110 F water. For this 0
test, the valve e
. demand. Upon de-energizing the valve ned opened for approximately for closure, the e
e g
The valve was removed from the
\\
12 seconds and then o
w 1 d Wthe Target Rock representative.
test facility and di s
No damage was observed h h..i Yt affect the ability of the valve to open/close on er Detailed data sheets are to e e
Appendix B, Section B-35.
Q e Opening Time The total valve opening time data f hall Steam Station tests and the Wyle Phase II &
t.e were obtained based on different types of inputs. As a i t, the recorded Marshall opening times exceed the recorded 3,es for similar steam test conditions.
In addi n
n disc opening times of the valve could nbt be accurate y deterr.ined at Wyle.
For that reason, th: main disc opening time was not included on the Wyle Phase II & III data she.ts.
e 27
i a
1 TABLE 4.3.ls "As TESTED" MARS!1ALL TEST MATRIX FOR Tile TARGET ROCK RElf EF VALVE,
d 1
0
" NOMINAL" i
TEST INITI AL CONDITIONS
" NOMINAL" l
T[ST,JIO.
,1YPl; AT VALVE INLET TRANSIENT CONDITIONS V
TEST I
I MAX DISCII.
TEMP l
PRESS DURATI P
PIPE 8.P.
FLUID
,F PSIA _
psig
,S h
2335 475 la Ste:am Steam (Sat.)
2435 2-5 S *.qm Steam (Sat.)
2435 5
2300 475 6*
Steam Steam (Sat.)
2445 D 60 2315 165 7 - 10 Steam Steam (Sat.)
15 2320 165 t>
h i
\\
9 e
- Tests I and 6 were extended doration flow measurement tests
.+
I
s TABLE 4.3.1b i
"As TESTED" WYLE PHASE III TEST MATRIX FOR THE TARGET ROCK RELIEF VALVE _
i i
i TRANS!ENT TEST NO.
TEST CONDITIONS TYPE INITIAL CONDITIONS l
I MAXIMUM MAX (STATIC + DYNAMIC)
V l.VE 15 CHARGE BEllDING AT VALVE INLET IN ACCUMULATOR TEST U
E M0 MENT MP PRESS. DURATION P
S.
INDUCED 1
fEMP. PRESS. FLU ]ID F
PSIA (SEC) 51 SIA IN-LB FLUID F
PSIA hl32 320 N/A 1-TR-IS STEAM STEAM 660 25kl SAME AS VALVE INLET 7
2134 330 N/A 2-TR-1S STEAM STEAM 669 2504
+
3-TR-3W WATER WATER 447 715 OD 639 N/A S15 2293 450 N/A.
(
4-TR-5W WATER WATER,6315 2515 l
g-l 5-TR-2W WATER WATER 114 690 s10 616 s1 N/A
> 0 l
6-TR-4W WATER WATER 448 2545 S10 2'196 395 N/A if 7-TR-7W WATER 11 2
~R 656 2506 s27 '
2172 520 N/A SIMULATION l
8-TR-5W WATER WATER 64 SAME AS VALVE INLET S10 2320 430 N/A 2302 425 16,400 j
9-TR-6W WATER WATER 645 2490 s10 i
l (PRELOAD)
~
S10 2028 325 N/A i
17-TR-15 STEAM STEAM 657 2510 slo 2620 315 36,600 i
18-TR-65 STEAM STEAM 658 2S05 (PRELOAD) 19-TR-95/W TRANSI-STEAM 656 2500 WATER 642 2504 slo 2310 435 N/A TION i
0 4.4 CONTROL COMPONENTS RELIEF VALVE
(
4.4.1 Conditions Tested Tests were performed on the Control Components relief valve model at the Marshall Steam Station and during Phase III of the Wyle Test Program. Tables 4.4.la and b present the matrix of conditions under which this valve model was tested at Marshall and Wyle (Phase III),1espectively.
4.4.2 Summary of Principal Observations Marshall Steam Station s
.11y opened on demand and fully closed on demand The a f
f the tan (10) evaluation test cycles.
Detaile sheets are contained in Appendix B. Section B-4a.
e Wyl e e
,s l
The valve 1
bea on demand and fully closed on demand 4Atest cycles performed through for each o
. ar June 19,1981.
V l
Detailed data s. a ontained in Appendix B, Section B-4b.
(
0
(
(
?'
l JC
l l
TABLE 4.4.la j
"AS TESTED" MARSHALL TEST MATRIX FOR TIIE CONTROL COMPONENTS RELIEF VALVE.
)
"N0HiNAL" I
TEST INITIAL CONDITIONS
" NOM 4,"
TEST NO.
TYPE AT VALVE INLET TRANS_IENTMNDhl0NS vhb
. OS MAX DISCH.
c TEMP PRESS I
PRESS.
PIPE B.P.
FLUID
'F PSIA (PSIA)_ PSIA 1*
Steam Steam (Sat.)
243S 0
2175 61 5 218'.i 61 5 2-5 Steam Steam (Sat.)
2435 Q5 6*
Steam
., ' Steam (Sat.)
243 60 2185 21 5 3
3 15 2185 21 5 7 - 10 Steam Steam (Sat.)
. s 0
9
?*.
.4
- Tests 1 and 6 were exten d durttion flow measurement tests t
e-e g
TABLE 4.4.lb "AS TESTED" WYIE PilASE III TEST MATRIX FOR Tile CONTROL COMPONENTS RELIEF VALVE 11Si l10.
If51 TRANSIENT TYPE INITIAL CONDITIONS CONDITIONS HAXI!10!!
MAX (STATIC + DYNAMIC)
VALVE DISCllARGE BEtlLING TEST CLO J PIPE M0!4E NT AT vat.VE INLET IN ACCIMILATOR fP. PRESS.
T{RP PRESS. DURATION I S..
RESS.
INDUCED Flul0 F
PSIA FLUID F
PSIA gE PSIA IN-LB
& l(7 ~
35-CC-15 Steam Steam 683 2760 Same as Valve Inl _
2330 468 M/A
~
36-CC-25 Steam Steam 683 2750 2280 416 N/A
( fai t eil Q
(
l; A.i r) 37-0C-35 Steam Sted 670 2535 4
2225 377 N/K.
(Preload
)
failed l
Air) 9 38-CC-511 Water Water 40 5
2180 400 N/A i
( f a lleil Alr) l l
- Balance of CCI Tests Completed After 6/19/81 -
l at
.c 9
1 C
.-O 4.5 MAS 0?!EILAN RELIEF VALVE
,(
4. 5.1 Conditions Tested Tests were performed on the Masoneilan relief valve model at the Marshall Steam Station and during Phase III of the Wyle Tables 4.5.la and b present the matrix of conditions Test Program.
under which this valve model was tested at Marshall and Wyle (Phase III), respectively.
4.5.2 Summary of Principal Observations _
e Marshall Steam Station ully opened on demand and fully closed on demand The 1
f e -h the ten (10) evaluation test cycles.
Detail a sheets are contained in Appendix 8. Section B-5a.
s e
Wyl e e
- la
~.:
o O
I 1
t I
(
O e
e L
43 e
e e
ey e d
TABLE 4.5.la
- "AS TESTED" HARSilALL TEST MATRIX FOR Tile MASONEll AN RELIEF VALVE.
I
" NOMINAL" l
TEST INITIAL CONDITIONS "N
NAL" TFST NO.
1YPE AT VALVE INLET TRANSIEN IDITIONS TEST R
MAX DISCH.
TEMP PRESS I
PIPE 8.P.
FLUID
'F PSIA A)
PSIA 1
1*
Steam Steap (Sat.)
2500 60 2235 535 1
2235 535 i
g5 2-S Steam Steam (Sat.)
2500 6*
Steam
, - Steam (Sat.)
25 60 2215 180 7 - 10 Steam Steam (Sat.)
5 15 2230 180 O
Q l
a
- Tests I and 6 were extended duration flow measurement tests l
l s
e e
TABLE 4.5.1b "AS TESTED" WYLE PHASE 1tI TEST MATRIX FOR THE MASONE!LAN RELIEF VALVE TRANSIENT TEST NO.
TEST TYPE INITIAL CONDITIONS CONDITIONS MAXIMU!i MAX (STATICt0YNAMIC)
VAL DISCHARGE BErlDitlG AT VALVE INLET IN ACCUMULATOR TEST CL.
PIPE MOMENT fEHP. PRESS.
IgMP PRESS. DUR N
5 PRESS.
INDUCED PSIA IN-L8 FLUID F
PSIA FLUID F
. PSIA e g O
^
S ATER -
O Q
e e
e b
g l
l l
4.5 COPES-VULCAN RELIEF VALVE (316 w/ stellite Plug and 17-4PH Cage)
~
4.6.1 Conditions Tested Tests were performed on the Coces-Vulcan relief valve model (316 w/ stellite Plug and 17-4PH Cage) at the Marshall Steam Station ano during Phase III of the Wyle Test Program.
Tables 4.6.la and b present the matrix of conditions under which this valve model was tested at Marshall and Wyle (Phase III),
respectively.
4.6.2 Su=ary of Princieal Observations MarsFMteam Station e
~
f 11y opened on demand and fully closed on demand a
f r ea "
the ten (10) evaluation test cycles.
Detai' d a.
sheets are centained in Appendix 5, Section B,-6a.
e Wyle Phas Mk lat Q
(
o l
~
C5 4
e
.=
e G
l l
'5
TABLE 4.6.la "AS TESTED" MAISHALL TEST MATRIX FOR THE COPES VULCAN RELIEF VALVE L316 w/ stellite Plug and 17-4PH Cage)
" NOMINAL" TEST INITIAL CONDITIONS "NON Al" TEST NO.
TYPE AT VALVE INLET TRANSIENT 0 ITIONS TEST C 05 MAX DISCH.
TEMP PRESS I
ON ES.
PIPE 8.P.
FLUID _A PSIA (P A)
PSIA 1*
Steam Steam (Sat.)
2435 60 2155 635 2165 635 2-5 Steam Steam (Sat.)
2435 g5 6*
Steam
, Steam (Sat.)
245 60 2145 205 7 - 10 Steam Steam (Sat.)
5 15 2165 21 5 h
1 j
- Tests I a'nd 6 were extended duration flow measurement tests l
l i
i TA8LE 4.6.lb "AS TESTED" WYLE PHASE III TEST MATRIX FOR Tile COPES-VULCAN RELIEF VALVE (316 w/ Stellite Plug anal 17-4Pff CageL TRANSIENT IEST NO.
TEST CONDITIONS T LPE, if',iAL CONDITIONS
\\
MAXIt10H (STATIC +0VNANIC)
MAX E b ISCHARGE BEtIDitlG AT VALVE INLET IN ACCUHillATOR TES w
IPE MOMENT PRESS.
INDUCED TFHli! FRESS.
qftF-~PAESS.
. IN-L8 Fluto
- F PSIA FLUID F
M
';t l
IA _
OSIA j
i o
62 O
9 i
- LATER -
l S
I 4'
a
4.h COPES-VULCAN RELIEF VALVE (17-4PH Plug and Cage)
(}
4.7.1 Conditions Tested Tests were performed on the Copes-Voican relief valve model with the 17-4Ph Plug and Cage at the Marshall Steam Station and during Phase.III of the Wyle Test Program. Tables 4.7.la and b present the matrix of conditions under which this valve model was tested at Marshall and Wyle (Phase III).
respectively.
4.7.2 Summary of Principal Observations e
Marshal,)sSteam Station Tb v v 11y opened on demand and closed on demand for ten (10) evaluation test cycles.
r After tk se sts were comp 1'eted, a new set of the same desig e
plug parts were insta11td and the valve was p1 ced "n the test facility. 1he valve was cycled to i 4 gat the cage to body gasket performance and to su 9 ab
.arshall Steam Station test functions.
The valve
/
c.
n demand and fully closed "on demand for the first ow cycles.
During the next seven cycles, the va to within at least 88% of the full closed position.
T
. e did not fully close on demand.
f Disassembly showe 1
f the cage and plug guiding i
(
surfaces.
Detailed data sheets a.
n' i d in Appendix B. Section B-7a.
l e
Wyle Phase III t
- later -
(:3>
\\
4 49 l
I
TABLE 4.7.la
- "AS TESTED" MARSilALL TEST MATRIX FOR Tile COPES-VULCAM MLIEF VALVE (17-4Pil Plug and Cage) t
" NOMINAL" TEST INITIAL CONDITIONS "N
AL" TEST N,0.
TYl'E AT VALVE INLET TRANSIENT 0 ITIONS T
M X DISCH.
TEMP PRESS PIPE 8.P.
FLUID
- F PSIA (PSIA)
PSIA 1*
Steam Steam (Sat.)
2445 60 2155 595 2200 61 0 2-5 Steam Steam (Sat.)
2445 g
15 6*
Steam
.. Steam (Sat.)
24 60 2155 195 7 - 10 Steam Steam (Sat.)
15 2190 195 O
Q
- Tests 1 and 6 were extended duration flow measurement tests I
I
(
l
TABLE 4.7.1b "AS TESTED" WYLE PHASE III TEST MATRIX FOR Tile COPES-VULCAN RELIEF VA'LVE (17-4PH P! in and Cage)
TRANSIENT TEST NO.
TEST TYPE _
INITIAL CONDITIONS CONDITIONS MAXI'10!!
MAX (STATIC +0YNN4IC) h L
DISCHARGE BENDING PIPE MOMENT AT VALVE INLET IN ACCUMULATOR TE T-2 TEMP. PRESS.
IgMP PRESS.
PRESS.
INDUCED FLUID F
PSIA FLUID F_
. PSIA
(
_ l SIA PSIA IN-LB C_ _
O e
O
- LATER -
l
- 4 5
e
4.8 MUESCO CONTROLS RELIEF VALVE 4.8.1 Conditions Tested Tests were performed on the MUESCO Controls reitef valve model at the Marshall Steam Station and during Phase !!! of the Wyle Test Program. Tables 4.8.la and b present the matrix of conditions under which this v'alve model was tested at Marsha11'and Wyle (Phase III), respectively.
4.8.2 Sumary of Princical Observations prshall Steam Station e
The v d fully opened on demand and fully closed on demand fo -
the ten (10) evaluation test cycles.
F th t ts were performed on the valva with a replacement stem, pl'
- d gaskets. These parts exhibited wear during the first
.ests and a second set of tests was recomended w.
by MU
.0
.t
's to, infomation purposes. The valve ful.1y epn and f;11y closed on demand for each of the opened on d y
e.
Similar wear patterns were found.
evaluatio Detailed data t
. contained in Appendix S, Section B-8a.
e Wyle Phase III
(
- later -
h C>
O I
\\
r s
g
TABLE 4.8.la l
"AS TESTED" MARSHALL TEST MATRIX FOR THE MUESCO RELIEF VALVE
" NOMINAL" TEST INITIAL CONDITIONS
" NOMINAL" TEST NO.
TYPE AT VALVE INLET TRANSIENTgNDITIONS V
MAX DISCH.
TEST C
a TEMP PRESS DURATI i FC PIPE 8.P.
O l
FLUID F
PSIA _
s PSIA I
1*
Steam Steam (Sat.)
2435 I
2395 235 (2485)**
(2395)** (255)**
2-5 Steam Steam (Sat.)
2435 4
0 6*
Steam
, Steam (Sat.)
24 60 7 - 10 Steam Steam (Sat.
2 5 Q
15 0
f
- Tests 1 and 6 were extended duration flow measurement tests
- (Second set of Tests) e.
e
TABLE 4.8.la "As TESTED" MARSHALL TEST MATRIX FOR THE MUESCO REllEF VALVE
" NOMINAL" TEST INITIAL CONDITIONS
" NOMINAL" TEST NO.
TYPE AT VALVE INLET TRANSIENTMNDITIONS V
TEST C
MAX DISCH.
s TEMP PRESS DURATI I E'.
'IPE B.P.
FLUID
,F PSIA S
PSIA 1+
Steam Steam (Sat.)
2435
}
2395 235 (2485)**
(2395)** (255)**
l 2-5 Steam Steam (Sat.)
2435 2400 235 (2375)** (255)**
g C
6*
Steam
. Steam (Sat.)
24 60 2395 80 (2415)**
(80)**
(
7 - 10 Steam Steam (Sat.
2 5 Q
15 2380 80 (2470)**
(80)**
C'
- Tests 1 and 6 were extended duration flow measurement tests
- (Second set of Tests)
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'4. 9 FISHER CONTROLS RELIEF VALVE
- 4. 9.1 Conditions Tested
(
Tests were perforn.ed on the Fisher controls relief valve model at the 'tarshall Steam Station and during Phase III of the Wyle lest PNgram. Tables 4,9.la and b present the matrix of conditions under which this valve model was tested at Marshall 'and Wyle (Phase III), respectively.
4.9.2 Summary of Principal Observations e
Marshall Steam Station The v vc fully opened on demand and fully closed on demand for e the ten (10) evaluation test cycles. At the f the test, the valve was disassembled and galling nu on the plug and cage mating surfaces.
wso a r' In add *6 the evaluatio'n tests, three other sets of i
cycles ver rmed on the valve. The first two sets of?
cycles wez rmed on a set of cage and plug parts which did not re e orrect Fisher Controls design for the PORV appli ar g the cycles, the valve closed on demand to wit n t 96% of the full closed position
+
on each cycle.
e cycles were completed, the valve was disassemb?.e d a ng was observed on the plug and cage mating surfa s galling was more severe than the
(.
evaluation test cy e a pattern.
The evaluation test wa rmed on a set of cage and plug parts with correct r c s.
These are the tests discussed in the first pa a
this section and they represent Fisher Controls lied to PWR plants with i
l the correct internals.
1 A fourth set of cycles were pe a set of trim with-the correct design clearances.
fully opened on..
demand and fully closed on demand ycl e.
A galling pattern similar to that observed in tion test was observed. Again, it was less severe th attern obser,ved wher, the valve did not fully close on dem,a Detailed data sheets are contained in
. x B, Section B-9a.
e Wyle Phase III'
- later -
ss l
t
s 1
TABLE 4.9.la "AS TESTED" MARSilALL TEST MATRIX FOR Tile FISilER CONTROLS REllEF VALVE,
" NOMINAL" TEST INITIAL CONDITIONS
" NOMINAL"
,1[ST NO.
TYPE AT VALVE INtET TRANSEENT ANDITIONS TEST C,
1 MAX DISCil.
TEMP PRESS DU TI P
PIPE B.P.
FLUID "F
PSIA _
_ PSIA 1a Steam Steam (Sat.)
2455 d
2255 485 l
2-5 Steam Steam (Sat.)
2455 15 2255 485 6*
Steam Steam (Sa t. )
2415 D 60 2235 155 I
15 2255 155
" Steam (Sat.)
2fl 7 - 10 Steam g
h I
- Tests I avid 6 were extem duration flow measurement tests a.
e n
TABLE 4.9.lb "AS TtSTED" WYLE PHASE III TEST MATRIX FOR THE FISilER CONTROLS RELIEF VALVE TEST HD.
TEST TRANSIENT TYPE INITIAL CONDITIONS CONDITIONS MAXI!!U!!
l MAX (STATIC + DYNAMIC)
VAL DISCHARGE BEilDING AT VALVE INLET IN ACCUMULATOR TEST L5 PIPE HOMENT TEMP. PRESS.
T{!iP PRESS. DUR PRESS.
INDUCED PSIA IN-LB y\\
l FLUID F
PSIA FLUID F
PSIA g\\
Q O
I C
i 9
- t.,
- LATER -
i e
.. ~
s e
e
4.10 GARRETT RELIEF VALVE
~
4.10.1 senditions Tested Tests were performed on the Garrett relief valve model at the Marshall Steam Station and during Phase III of the.Wyle Test Program. Tables 4.10.la and b present the matrix of conditions under which this valve model was tested at Marshall and Wyle (Phase III), respectively.
4.10.2 Summary of Princioal Observations e Marshal}sSteam Station i 11y opened on demand and fully closed on demand Tb v v f the ten (10) evaluation test cycles.
e, Additior
. les oere perfo'rmed on the valve. During these bonnet gaskes leakage developed.
In all cycle,
.c-fully closed on demand. Disassembly showed cycles, th v wash-out f ge +o body gasket. As a result of the test ncorporated design modifications into observati
=r the test va
.'r yh>PhaseIIItestsandintov'alvesbeing supplied to P' s ontained in Appendix B, Section B-10a.
Detailed data sh. a
(
e Wyle Phase III
- later -
s i
s*
O WW
l TABLE 4.10.la "AS TESTED" MARSHALL TEST MATRIX FOR TliE GARRETT RELIEF VALVE
" NOMINAL" TEST INITIAL CONDITIONS
" NOMINAL" TEST NO.
TYPE AT VALVE INLET
- TRANSIENT CONDITIONS I
VALVE i
TEST CLOSURE MAX DISCH.
TEMP PRESS DURATION PRESS.
PIPE B.P.
FLUID F
PSIA (SEC)
(PSIA)
PSIA g
1*
Steam Steam (Sat.)
2445 60 2015 815 2-5 Steam Steam (Sat.)
2445 15 2045 815 t
6*
Steam
, Steam (Sat.)
2615 60 2035 335 7 - 10 Steam Steam (Sat.)
2615 15 2465 345 Tests I and 6 were extended doration flow measurement tests G
S
o I
TABLE 4.10.lb "AS TESTED" WYLE PilASE Ill TEST MATRIX FOR THE GARRETT RELIEF VALVE TRANSIENT TES1 110.
TEST CnNDITIONS TJPE INITIAL CONDITIONS E
MAXI!10!!
MAX (STATIC + DYNAMIC) l
'l VALVE DISCHARGE BENDING AT VALVE INLET IN ACCIMILATOR TEST CLOSURE PIPE MONENT RP PRESS. DURATION PRESS.
PRESS.
INDUCED
~Flulu }fNP. PRESS.
FLil]lD F
PSIA _,
F fSIA (SEC)
PSIA PSIA IN-L8 i
- LATER -
g.
I t;
a
.e II s
O
~
II.D.1 PERFORMNCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR RELIEF AND SAFETY VALVES (NUREt:-0578 SECTION 2.1.2)
Position Pressurized-water reactor.and boiling-water reactor ifcensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves 'under expected operating conditions for design-basis transients and accidents.
Changes to Previous Requirements and Guidance A.
Safety and Relief Valves and Piping--The types of documentation required for safety and relief valves and piping and the specific submittal dates are considered to be a clarification of item II.D.1 as described in NUREG-0660. The submittal of infomation was implied but not explicitly discussed in that report.
B.
Block Valves--Qualification of PWR block valves is a new requirement.
Since block valves must be qualified to ensure that a stuck-open relief valve can be isolated, thereby terminating a small loss-of-coolant accident due to a stuck-open relief valve. Isolation of a stuck-open power-operated relief valve (PORV) is not required to ensure safe plant shutdown. However isolation capability under all fluid conditions that could be experienced under operating and accident conditions will result in a reduction in the number of challenges to the emergency core-cooling system. Repeated unnecessary challenges to these system are undesirable.
C.
ATWS Testing--Testing of anticipated transients without scram ( ATWS) for later phases of the valve qualification program was noted in item II.D.1 of NUREG-0660. The clarification below provides updated l
information on PWR ATWS temperature and pressure conditions and clarifies that ATVS testing need not be accomplished by July 1981.
l Clari fication Licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70 Revision 2.
The single failures applied to these analyses shall be chosen so that I
the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis l
procedures.
Reactor coolant system relief and safety valve qualification shall include qualification of associated control circuitry, piping, and supports, as well as the valves themselves.
3-72 l
l
~
8 9
w O
ENCLOSURE 3 M
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act:
~9, UNITED STATES
,[*
c() cI W ASHINGTON, D. C. 20555 NUCLEAR REGULATORY COMMISSION I
s.;.... /
TO ALL LICENSEES OF OPERATING PLANTS AND APPLICANTS FOR OPERATING
~
LICENSES AND HOLDERS OF CONSTRUCTION PERMITS Gentlemen:
e.
SUBJECT:
REVISED SCHEDULE FOR COMPLETION OF TMI ACTION FLAN ITEM II.D.1, RELIEF AND SAFETY VALVE TESTING On October 31, 1980 the NRC staff tr'ansmitted a Clarification of TMI
~
. Action Plan Requirements (NUREG-0737). Item II.D.1 of that document
" Relief and Safety Valve Test Requirements" set forth implementation schedules of 7/1/81 for ampletion of the RV & SV test program and 10/1/81 for the submittal of plant specific reports.
We have completed our review of a request for senedule relief for completing that oortion of the item related to the PWR (EPRI) testing program. The Comiission has approved a revised schedule in response to thh request. The revision, as indicated in the enclosed page changes to NUREG-0737, extends completion of the test program until April 1,1982 and of the plant specific reports until July 1,1982.
Sincerely.
Darrell G. Eisenhut, Director Division of Licensing Office of Nuclear Reactor Regulation
Enclosure:
NUREG-0737 Revised Pages 1-5, 2-6, 3-72, 3-74 i
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2-6 Clarlfl-Implemen-Plant Require-Clartfi-Preleple- *ostimple.
Tech Licensee cation Shortened tation Appilca-ments cation mentation mentation Spec Submittel Itsa Title Description Schedule btlity Issued Issued Approval Review Reg.
Req. by Asiaarts 11.D.1 Relief 8 safety-
- 1. Describe program Fuel load All 9/27/19 11/9/19 Ro valve test a tr%edule requirements
- 3. Block Valve Tests Fuel load or NR 11/9/19 by 7/1/82, inct 3 whichever is later II.D.3 Valve position Install in control a
.All 9/27/79 11/9/79 Tes indication room Entt 3 II.E.1.1 Avallfary Fee hater
- 2. Modifice* N Full pouer NR 4/24/80 Mone As required ll.E.I.2 Auxillary fee &ater I. Initta e
system initiation (a) Coe.t,91 grade Fuel load
/WR 9/27/79 11/9/79 Yes and flow (b) $afety grade a
NR 9/27/79 11/9/79 Tes
- 2. Flow ladication (a) Control grade Fuel load PWR 9/27/79 11/9/79 ves (b) Safety grade a
WR 9/27/19 11/9/79 ves ll.E.3.1 Emergency pouer for Installed capability 4 mes prior to Pwn 9/27/19 11/9/79 Tes pressurizer heaters issuance of SER tact 3
!!.E.4.1 Dedicated hyeTogen
- 1. Design a
All 9/27/79 11/9/79 he penetrations
- 2. Review & revise Fuel load All 9/27/19 Inti 3 no H7 control proc
- 3. Install 7/1/81 or prior All 9/27/79 Enct 3 no to issuante of OL Requirement formally 1: sued by this letter i
AFour month; before operating 1(cense js issued or 4 months before date Indicated
- 4.
g
I I. D.1 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSbRIZED-WATER REACTOR RELIEF AND SAFETY VALVES (i4UREG-0578, SECTION 2.1.2)
Position Pressurized-water reactor,and boiling-water reactor licensees and applicants shall conduct testing to qualify the reactor coolant system relief and safety valves under expected operating conditions for design-basis transients and accidents.
Chances to Previous Requirements and Guidance A.
Safety and Relief Valves and Piping--The types.of documentation required for safety and relief valves and piping and the specific submittal dates are considered to be a clarification of item
~
II.D.1 as described in NUREG-066D. The submittal of information was implied but not explicitly discussed in that report.
B.
Block Valves--Qualification of PWR block valves is a new requirement.
Since block valves must be qualified to ensure that a stuck-open relief valve can be isolated, thereby terminating a small loss-of-coolant accident due to a stuck-open relief valve.
Isolation of a stuck-open power-operated relief valve (PORV) is not required to ensure safe plant shutdown. However isolation capability under all fluid conditions that could be experienced under operating and accident conditions will result in a reduct4on in the number of
~
challenges to the emergency core-cooling system.
Repeated unnecessary challenges to these system are undesirable.
C.
ATWS Testing--Testing of anticipated transients without scram ( ATWS) for later phases of the valve qualification program was noted in item II.D.1 of NUREG-0660.
The c.larification below provides updated information on PWR ATWS temperature and pressure conditions and clarifies that ATWS testing need not be actomplished by July 1981.
Cla ri fi cation Licensees and applicants shall determine the expected valve operating conditions through the use of analyses of accidents and anticipated operational occurrences refcrenced in Regulatory Guide 1.70, Revision 2.
The single failures applied to these analyses shall be chosen so that the dynamic forces on the safety and relief valves are maximized.
Test pressures shall be the highest predicted by convent'ional safety analysis procedures.
Reactor coolant system relief and safety valve qualification l
shall include qualification of associated control circuitry, piping, and supports, as well ns the valves themselves.
3-72 Y
d k
4 '
l A.
Performance Testing of Relief, and' Sa'fety Yalves--The following information must be p ovided in report form by October 1,1981 for BWRs and July 1,1982 for PWRs.
(1)
Evidence supported by test of safety and relief valve function-ability 'for expected operating and accident (non-ATWS) conditions must be provided to NRC.'
The testing should' demonstrate that the valves will open and reclose under the expected flow conditions.
Documentation Required Preimplementation review will be based on EPRI, BWR, and applicant submittals with regard to the variout test programs. These submittals should be made on a timely basis as noted below, to allow for adequate review and to ensure that the following, valve qualification dates can be met:
Final PWR (EPRI) Test Program--July l',1980 Final BWR Test Program--October 1,1980
~
Block Valve Qualification Program--January 1,1981 Postimplementation review will be based on the applicants' plant-specific '
submittals for qualification of safety relief valves and block valves. To properly evaluate these plant-specific applications, the test data and ~
results of the various programs will also be required by the following dates:
BWR Generic Test Program Results--July 1,1981 PWR (EPRI) Generic Test Program Results--April 1,1982 Plant-specific submittals confirming adequacy of safety and relief valves based on licensee / applicant preliminary review of generic test program results--July 1,1981 for BWRs; April 1,1982 for PWRs Plant-specific repor+s for safety and relief. valve qualification--
October 1,1991 for BWRs; July 1,1982 for PWRs Plant-specific s6mittals for piping and support evaluations--
January 1,1982 for BWRs; July 1,1982 for PWRs Plant-specific submittals for block valve qualification--July 1,1982 Technical Specification Changes Required No technical specification changes are r_equired.
References NUREG-0578 NUREG-0660, Item II.D.1 3-74 e
'~
UNITED STATES OF AMERICA NUCLEAR REGULATOM COM'ilSSION BEFORE THE AT0f11C SAFETY AND LICENSING BOARD
)
In the Matter of
)
METROPOLITAN EDISON CO. ET AL.
Docket No. 50-289 (Three Mile Island Nuclear (Restart)
Station, Unit 1)
JOINT AFFIDAVIT OF EDGAR G. HEMMINGER AND WALTON L. JENSEN, JR.
Edgar G. Hemminger and Walton L. Jensen, Jr., state under oath as follows:
1.
I, Edgar G. Hemminger, am a inechanical Engineer in the Division of Engineer-ing, Mechanical Engineering Branch, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
I am responsible for review and evaluation of structural integrity, operability, and functional capability of safety related mechanical equipment, which includes ev&luation of unsatis-factory safety and relief valve test results. A copy of my professional qualifications is attached.
2.
I, Walton L. Jensen, Jr., an a senior engineer assigned to the Reactor Systems Branch, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Commission.
I am currently responsible for the branch review of Tiil-l. A copy of my professiont.1 qualifications is attached.
3.
The "NRC Staff's Report to the Board on Safety Aspects of EPRI Test Data on Relief and Safety Valves" was prepared by us and is true and correct to the best of our knowledge and belief.
s W
v4D A,
Edgar G) Hemminger
[
i f* ton L. jet 1sen,/Jr.
u/
Wal
./
Subscribed and sworn to before me this 3 rJ day of September, 1981.
dVndan. EL No'tary Public g-My Commission Expires:
/, /Tg)
EDGAR G. HEMMINGER 0FFICE OF NUCLEAR REACTOR REGULATION U. S. NUCLEAR REGULATORY COMMISSION PROFESSIONAL QUALIFICATIONS I am a Mechanical Engir$eer in the Division of Engineering, Mechanical Engineering Branch, Office of Nuclear Reactor Regulation of the United States Nuclear Regulatory Comission.
I am responsible for review :.nd evaluation of the structural integrity, operability, and functional capability of safety related mechanical equipment and components.
I hold a. Bachelor of Science Degree in Mechanical Engineering from Ohio University and a Master of Science Degree in Mechanical Engineering from Drexel University and am a licensed Professional Engineer in the State of New York.
From 1965 thru 1979, I was employed by the General Electric Company at the Knolls Atomic Power Laboratory in Schenectady, New York. My work experience was in the area of therual and stress analysis of reactor plant components and equipment.
I have specifically evcluated steam generators, reactor vessels, nozzles, closure heads, pumps and piping systems. Using finite element computer methods, I have modeled the vessel closure Fead and core barrel bolt up region to determine preload relaxation and lift off for various operating and accident conditions.
I have also used results of the above type calculations ir. conjunction with fracture mechanics methods to establish safe heat up and cooldown pressure and temperature limits for normal l
plant operation.
l In 1973, I corpleted a one year training program for test and start up of naval reactor plants aboard ship. Fron 1973 thru 1979, I contributed to the construction, start up and power rance physics testing of eight reactor plants aboard ship. My primary duties were to review the test procedures and test data for acceptance testing of naval reactor plants aboard ship and to provide technical support to the shipyard in resolution l
of equipment problems dealing primarily with valves, pumps, and heat l
exchangers.
I joined the NRC in October, 1979.
l l
l
e WALTON L. JENSEN, JR.
PROFESSIONAL QUALIFICATIONS I am a Senior Nuclear Engineer in the Reactor Systems Branch of the Nuclear Regulatory Cor. mission.
In this position I am responsible for the technical ar.alysis and evaluation of the public health and safety aspects of reactor systems.
Frcs June 1979 to Deccaber 1979, I was assigned to the Bulletins and Orders Task Force of the Nuclear Regulatory Commission.
I participated in the preparation of NUREG-0565, " Generic Evaluation of Small Break Loss-of-Coolant A:cident Eetsvior in Esbcock & Wilcox Designed 177-FA Cperating Plants."
From 1972 to 1976, I was assigned to the Containment Systems Branch of the NRC/AEC, and from 1976 to 1979, I was assigned to the Analysis Branch of the l
NRC.
In these positions I was responsible for the development,and evaluation of co puter programs and techniques to calculate the reactor system and contain;.ent system response to postulated loss-of-coolant accidents.
From 1957 to 1972, I was employed by the Babcock and Wilcox Company at Lynchburg, i
i Virginia.
There I was lead engineer for the development of 1 css-of-coolant co puter programs and the qualification of these programs by co parison with j
experir. ental data.
e
s.e from 1963 to 1967, I was employcd by the Atomic Energy Commission in the Division of Reactor Licensing.
I assisted in the safety reviews of large power reactors, and I led the reviews of several small research reactors.
I received an M.S. degree in Nuclear Engineering at the Catholic University of America in 1968 and a B.S. degree in Nuclear Engineering at I:ississippi State University in 1963.
I am a graduate of the Oak Ridge School for Reactor Techrolog3, 1963-1964.
I am a cember of the American Nuclear Society.
I am the author of three scientific papers dealing with the response of B&W reactors to less-of-Coolant Accidents and have authored one scientific paper dcaling with containment analysis.
I I
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-a UNITED STATES OF AMERICA NUCLEAR REGULATORY C0tVIISSION BEFORE THE AT0flIC SAFETY ANP LICE;; SING BOARD, In the Matter of METROPOLITAN EDIS0N C0. ET AL.
Docket flo. 50-289 (Three Mile Island fluclear Station, Unit 1)
(Restar'.)
NRC STAFF'S REPORT ON BOARD'S COMMENTS REGARDIrlG BOARD NOTIFICATION OF UN5ATISFACTORY TEST RESULTS OF SAFETY VALVES In its Order dated August 25, 1981, the Atomic Safety and Licensing Board for the Three Mile Island fluclear Station, Unit 1 (TMI-1) Restart Hearing noted that the flRL staff did not notify the Board on the 1111-1 proceeding l
regarding the unsatisfactory Electric Power Research Institute (EPRI) test results for the safety valve instr.lled in TMI-1. A staff memorandum fr )m J. P. Knight to R. L. Tedesco and T. M. Novak dated July 1, 1981 enclo;ed the EPRI memorandum of June 26, 1981 which reported on the tests. The THI-1 Board became aware of the matter through NRC Board flotification No.
81-20, dated August 11, 1931, filed in the McGuire proceeding. The Board Order requested the staff, among other things, to inform t;ie Board promptly whether notification of this matter by the staff would haire been appropriate in this proceeding, and if why not.
The nanbers of the staff that prepared this report discuss their reasoning herein as to why notification of the Till-1 Board was not considered appro-priate. However, The Director, Division of Licensing, was not provided the opportunity, in accordance with current guidelines in Nuclear Reactor Regula-tion (NRR) Office Letter No.19 Rev.1 dated December 9,1980 (enclosed), to
F
-P-review the reconmend'ations and the EPRI test results to make his determination as to whether the test results were material and relevant.
In retrospect, The Director, Division of Licensing would have likely decided to notify the Till-l Board similar to the notification filed in the I'cGuire proceeding based on the Cornission's policy cited in the NRR Office Letter No.19. How-
~
ever, the staff discusses below why it believes that the unsatisfactory EPRI test results reported in the June 26, 1981 EPRI memorandum are not significant with respect to the issues in the T!11-1 proceeding.
The staff reviewed the unsatisfactory EPRI test results reported in the EPRI memorandum dated Jura 20, 1981 regarding their relevance and safety significance to the issues in the THI-1. proceeding prior to considering notifying the Tl11-1 Hearing Board. The basis for the unsatisfactory test report was that rated flow in accordance with the EPRI screening criteria was not met during a high back pressure steam test.
t This test was only one part of the early phase of the EPRI test program and although some screening test criteria have not been met, the testing i
to date has not identified a safety problem with the safety or relief valves that would affect the staff's position on the T!11-1 hearing record.
The "NRC l
Staff's Report to the Board on Safety Aspects of EPRI Test Data on Relief and Safety Valves" that was prepared by Edgar G. Hemminger and Walton L.
Jensen, Jr. provides a more detailed discussion of the valve test results.
The principal staff concern stated in the TMI-l hearing record on this matter was the need to demonstrate that the safety and relief valves can withstar,d loadings from two-phase and solid flow; Zudans (UCS 6), ff. Tr 8824, at 5; and those EPRI tests on safety valves had not yet been conducted.
Testing to date involving two-phase and solid flow for the Dresser type power operated 1
- relief valve as used on TMI-l does not show unacceptable results. Therefore, for the reasons stated above we did not believe the failure of a safety valve to meet EPRI screening criteria during this steam test to be significant with respect to the issues in the Till-1 proceeding.
In addition to the EPRI test report of June 26, 1981, the staff received other EPRI test reports on relief and safety valves of the TMI-l type, some of which show test results that deviate from the EPRI screening criteria.
In the cases discussed below, the staff also concluded that the results were not material to the TMI-1 hearing record issues:
1.
EPRI test report dated May 15,1981 (enclosed in Staff memorandum from J. P. Knight dated May 19, 1981 to Tedesco and Novak) noted unsatisfactory test results on a Dresser (power operated relief valve (PORV) of the type used at THI-1.
In that test, the un-satisfactory results were associated with the effects of an upstream simulated water seal. Since the PORV at THI-l does not have a water i
seal feature, the staff concluded that the water seal test effects should not be representative of THI-1 valve behavior.
2.
CPRI test report dated July 2,1981 (enclosed iri Staff memorandum
" rom J. P. *; night dated July 16, 1981 to Tedesco and Novak).
3.
EPRi test report dated July 10,1930 (enclosed in Staff memorandum i
from J. P. Knight dated August 6, 1981 to Tedesco and Novak).
l Reports 2 and 3 included results of tests on the Dresser safety valves of the type used at TMI-1.
In those tests, rated flow was achieved but valve closing pressures were below the EPRI screening criteria for valve
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The staff does not believe that the valve closing
,losure pressures.
pressure test results are material to the TMI-l hearing, since the valve Also the acceptably performed its minimum relief capacity function.
deiayed closure is not an unreviewed safety concern, and further, does not correspond to a pressure level that would challenge the plant's The test results would not affect the staff's engineered safety features.
position on the THI-l hearing record.
Copies are enclosed for the Board's information of the four nenoranda cited I
i in this report from J. P. Knight to R. L. Tedesco and T. H. Novak that enclosed the EPRI memoranda reporting on the valve tests.,
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The staff does not believe that the valve closing closure pressures.
pressure test results are material to the TMI-1 hearing, since the valve Also the acceptably performed its minimum relief capacity function.
delayed closure is not an unreviewed safety concern, and further, does not correspond to a pressure level that would challenge the plant's The test results would not affect the staff's engineered safety features.
position on the THI-1 hearing record.
Copies are enclosed for the Board's information of the four memoranda cited in this report from J. P. Knight to R. L. Tedesco and T. H. Novak that enclosed the EPRI memoranda reporting on the valve tests..
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UNITED STATES i-NUCLEAR REGULATORY COMMISSION actofy Test ~Results' bn~
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MEM0MNDUM FOR:
Darrell G. Eisenhut, Director Division of Licensing.
'd Richard H. Vollmer, Director, Division of Engineering 4
Stephen H. Hanauer, Director Division of Human Factors Saiaty Denwood F. Ress, Director, Division of Systems Integration 4
-6 Themas E. Murley, Director, Division of Safety Technology Eernard J. Snyder, Program Director, TMI Program Office M
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. Harold R. Denton, Director Office of Nuclear Reactor -
.3 FROM:
Regulation U..,,
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SUBJECT:
KRR OFFICE LETTER NO.19, REVISION 1
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PROCEDURES FOR NOTIFICATION TO LICENSING BOARDS OF
.3 RELEVANT AND MATERIAL NEW INF0TdATION G1
!k Effective i=ediately, all NRR. personnel will use the following revised procedures for assuring prcmpt and appropriate action on notifying Licensing
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-Soards, Aopeal Panel and the Comission of new infomation which is considered iN by thsi stiff to be relevant and material to ene or more licensing proceedings.
These revised procedures reflect the experience we have gained since issuing l
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the original Office Letter No.19 on July 6,1978.
NRR staff members to be alert This, Office Letter places an obligation on all to the significance of new infomation that is developed in the course of their review and to consider whether this infomation could reasenably be regarded as putting a new or different light up:n an issue before Boards or as raising a new issue after publication of the staff's principal evidentiary documents.
This is the central theme of the procedures and requires the exercise.of g6ed judgment to assure that Boards will not be burdened with material beyond that potentially significant to the individual licensing proceedings.
Harold R. Denten, Diretter Office of Nuclear Reactor Regulation
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Enclosure:
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Ecard Notification A;_
Procedure l
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Ty-Following Comission approval of its Board Notification policy on May 4 e
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1978, the Of fic~e of Nuclear Reactor Regulation issued NRR Office Letter
- .f No.19, dated July 6,1978, which centained Board Notification procedures gy to be ir
- plemented by NRR. The term " Board Kotification" refers to new infcrmation which is considered to be relevant and rsterial to
-C T,y' one or more licensing proceecings', i.e., material relating to an issue before a Licensing Board, Appeal Panel, or ti.e Comission which can reascnably be regarded as putting a new or different light on that
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issue, or raising a new iss::e. (Note that the term ' Board" will be a-p.
used in this procedure to refer to Licensing Boards, Appeal Panel and
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Comission.)
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,d In a memorandum dated May 10, 1978, the Comission requested that an' evaluat of the Board Notification policy be prepared when approximately one year l R'z..-
of experi ence wr.s available.
To this end, Comission Paper SECY-80129, f.. ; '
dated Maren 10, 1980, provided an assessment of then current procedures l'.i '
anc prcposed changes to those procedures to correct problems encountered in carrying out the Board Notification policy.
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DISCUSSION b
There were three significant changes to the Board Notificatfon procedures "reccmenced in SECY-80-129 and approved by the Comission:
C 1.
Change the time threshold for initi,ating the formal Soard Not1fication procedures from the issuance of the ACRS Supplement and FES No N days before *.he start of the evidentiary hearing.
2.
Eliminate the routine transmittal to the Beards of staff cerr'esponcence and notic,es to applicants and ifcens.ees.
Staff correspe'ndence and notices to~ applicants and licensees sould be sent to the Beard only if it is determined to meet the guidelines for Board Motification.
~
i.e., new information considered material and relevant.
3.
Incorpcrate the guidelines for staff aparaisal and evaluation of Board Notification matte. set forth in ALAB-551, as follows:
supoly an exposition adequate to allow a ready appreciation of the a.
precise nature of the Board Notification catter; supply an exposition adequate to allow a ready appreciation of the l
. i.,
3.
cx: ant to what the Scard Notificattan matter might have a bearing upen the particular facility before the board; in the event a conclusion with regard to the safety or environmenta j
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significance of the. Board Nctification matter is presented, set for l
the reasoning underlying that conclusion sufficient to allow the
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board to make an informed judgment on the validity of the conclusic 7
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where the heard has limited jurisdictiori, spell out tlie pSssible relationship between the subject matter of the rotification and one or mre of. the issues Mfore the board.
3 C.
DETERMINATION OF RECOM.MENDATIONS FOR BOARD NOTIFICATION BY TECHNICAL REVIEW GROUPS AND PROJECT MANAGERS The Board Notification policy is applicable to operating license proceedings as well as construction permit proceedings.
In these proceedings the staff will send new information relevant and : s.,
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raterid to safety or environemntal issues to the Boards regardless of the specific issues which have been placed in controversy. This practice includes. proceedings for the conversion of provisional to full-term operating licenses.
In hearings concerning operating license amendments Board Notification is limited to the issues under consideration in the hearing. All staff members are responsible for reviewing all information received in the course of their assigned tasks, including reports identified by the Research and Standards Coordination Branch as being appropriate for consideration for Board.:
Notification, to determine whether it may be related to licensing proceedings and ray represent relevant and material new information jhich should be provided to appropriate Boards._
Information received from outside sources and censidered to be suitable for Board Notification should be handled in an expeditious manner. Sone examples of inferration from outside sources are:
(1) the reporting of errors discovered in a vendors.Errergency Core Cooling System (ECCS) models or codes which could result in changes to analyses previously evaluated and discussed in the SER, (2) t.he reporting of geological features which cou1d result in significant changes to those pre.viously repersed by the applicant and evaluated by the staff as discussed in the SER, and (S) those reports identified by the Research and Standards Coordination Branch as being appropriate for consideration for Board Notification.
Internally generated information that could reasonably be regarded as l
putting a new or different light upon an i. sue before Boards should also be reported as expeditiously as practicable. However, the Comission's policy recognize's the difficulty of determining the point when an individual staff. member's perceived concern has developed into!.
a staff issue of sufficient importance that Boards.are to be notified.--
In accordance with the Comission's policy, internally generated information should be provided to Boards at the point when the staff determines that.it is necessary to get more information about a problem from a source external to the staff. That.is, if such new information is detennined to be of sufficient inportance to seek further informatio6 analyses, tests, etc., from licensees or ventors, NRC contracts, or others outside the NRC staff, then the issue has developed to the point where concerned Boards should be informed.
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As for internally generated information, technical papers and journal articles should be provided to Boards at a point when the staff determines that (1) such information is of sufficient importance to call into question staff positions and criteria or (2) the staff has cetermined to seek further information, analyses, tests, etc., frca licensees, vendors, NRC contractors or others outside the staff.
1.
Staff members shoulo provide promptly the fo11cwSg information, through their ranagement, to the Director. Division of Licensing:
a.
The item recomended for notification of Boards.
b.
An exposition adequate to allow a*rea(y appreciation of the precise nature of Board Notification matter.
c..
Considerations regarding relevancy and materiality; i.e.,
l putting a new or different light upon an issue before the Soard or raising a new issue.
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An exposition adequate to allow a ready appreciation of the l
extent to what the Scard Notification matter might have a bearing upon the particular facility before the Board.
l e.
A statement as to the perceived significance of the information as it may affect current staff positions.
(A clear assessment of the significance is not required at this time and the recomendation should not be celayed in order to permit lengthy determinations.
If a clear assessment and final resolution is available, it obviously provides for a clean Board submittal.
For all recom.endations which do not contain a final resolution followup action is required to inferin the Boards as to the ultimate staff cisposition.)
f.
In the event a conclusion with regard to the safety or environmental sienificance of the Board Notification matter is i
i presented, set forth the reasoning unde'rlying that conclusion sufficient to allow the Board to make an informad judgment on l
the validity of the conclusion.
Where the Board has limited jurisdictien, spell out the possib.le g.
relationship between the subject matter of the nctification and one or mere of the issues before the 5 card.
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2.
relevant and material to Boards considering facilities 1 Sensed under Part 70 and under the cognizance of the Office of Nuclear Material Staff members should make any such Safety and Safeguards (NM55).
recomendations through their management to the Director, Division of The information provided should, to the extent possible, Licensing.
conform to that listed in Item 1. above. The Director, Division of Licensing..will forward the Board Notification material to the 09ector, Office of Nuclear Material Safety and Safeguards...
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Recomendations may be Judged by the Direct.o.r. Divis. ion of Licen31nsi.
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bot to be caterialmd r.gJ.gyf.nLJtad a merqrn_ndom to that effect wil t,be If the originator still feels that the ~
provi ded t o tu ^H ehtra, information should be provided to Boards, he or she should so state in
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Such a followup recomendation will be a followup reconcendation.
Although processed through the normal Board Notification channels.com.
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information not to be' relevant and material, it will be forwarded to the Board.
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- k. Board Notifications on differing professional opinions will follow the procedures of NRC Manual Chapter 4125, " Differing Professional Opinions. "
PROCESSING OF BOARD NOTIFICATION RECOMMENDATIONS D.
The key to comencement of' Board Notifications on a specific case is the 1.
establishment of the date for the beginning of evidentiary hearing and issuance of related notice by the Board. Prior to 30 days before the hearing, new material which is considered material and relevant to a proceeding is presented to the Boards via SER supplement or o documents.
disposed of, a sum.ary list is to be provided by the project managerFor cases within to the Board 30 days before the start of the hearing.
30 days of (or during) the evidentiary hearing new material found eterial and relevant shall be forwarded promptly to the Bo'ard according to these procedures.
OELD will provide DL with periodic updates of a list of current -
2.
proceedings for facilities under the cognizance of DL, indicating whether the Licensing Board, Appeal Board or Comission has jurisdiction over proceedings.
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The Dffice of the Director, DL wil' :stablish and raintain the' record-1.eeping system related to all Board Notification ratters. "
This will include a log of current proceedings and a detailed list
~
cf issues under. consideration.
4 The Director, Division of Licensing, shall review all reccamdations and determine whether they are relevant and raterial (5 working days from logging).
Recomendations containing inferration considered to be directly related to a specific case art also reviewed for applicability to other cases.
If it is determined that a recemendation is not considered to be relevant and raterial, a mecerancum to that effect is sent to the recomending parties.
If the inforra-ion and accoganying recomendatien are not clear enough for a detarmination to be rade, the Director will request clarifying information from the criginator.
5.
For instances prior to 3D days of the evidentiary hearing, the Director, Division of Licensing, shall forward a recorandum to the cognizant DL Assistant Directer (s) advising them that the item be brought to'the attentien of the Board through incorporation in the SER or as supple-
- r. ental staff testimony. A copy cf the menorandum will be sent to the criginator. The project ranager is responsible for seeing that the item is covered in evidentiary decurents unless it has been
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determined that the item has been resolved and that 5:ard Notification-i is net required.
Final dispcsition shall be reported to the Office
['
cf the Directer, DL (Scard Notification Coordinator).
5.
For instances within 3D days of (or during) the evidentiary hearing, the Director, Divisien of Licensing, shall ferward a memcranda to the cegnizant CL Assistant Director advising them that the item
- ust be brcught procotly to the attention of the appropriate. Boards.-
The cognizant.DL Assistant Director thall assure that the item is brcught progtly to the attentich cf the Beards (5 werking days frem receipt of the Director's mercrandum).
C: pies of the Board Nctification shall be sent to the criginator, technica1' review group, Office cf the Director, DL (Scard Nctification Cecrdinater) and
~
DELD (Hearing Division Director and Chief Counsel).
7.
A finding by the Directer, Division of Licensing, with regard to Board rec:crendations shall be tviewed by the DL Assistant Directors for applicability to pr:ceedings related to applications for c:nstruction t
permits, post-CP prcceedings, a:plicatiens fer cperating licenses, as well as proceedings relating to issuanca of license arendrents..
Pr:ceedings related to research and test facilities licensed under Part 50 are to be taken int.c mndeeration also.
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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OY 191981 ME!CRANDUM FOR: Robert L.'Tedesco, Assistant Director for Li. censing. DL Thomas H. Novak, Assistant Director for Operating Reactors, DL FROM:
James P. Knight, Assistant Director for Components & Structures Eroireering, DE
SUBJECT:
REPORTING OF U C ATI! FACTORY EPRI/PWR TEST RESULTS FOR POWER OPERATE,0 RELIEF VALVES
References:
(a) EPRI memorandum, dated 5/1/81 (b) EPRI memorandum, dated 5/15/81 As described in the referenced memorandums, the Dresser PORV model no. 31533VX-30 and the Target Rock PORV model no. 80X-006-1 failed the initial loop seal simulation tests at Wyle. The valves opened on 110' F water at full pressure 2500 psi, but failed to close as water temperatJre was ramped up to 650*F, a condition similar i
tc that experienced in plant; t;'th loop seals upstream of the PORV's.
The Dresser PORV in question is believed to be installed in CE and Br.W PWR's only and Fort Calhoun specifically is known to have loop seals upstream of Dresser PORV's. The Target Rock valve is reportedly not usec' on any operating plants
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but is planned for use on some plants presently under construction.
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It is requested that operating PWR's and NTOL's be contacted to detennine what corrective action, if any,.is being taken by the' licensees and NT0Ls for which
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the above test results are applicai,le.
In addition, this information may also be It is further noted that the Target relevantforlicensingboardnotification.
Rock and Dresser PORV s in question were disassembled and, inspected and no visible The Mechanical de age was observed which would affect future operation or testing.
Engineering Eran.ch will fon ard the results of future testing of these valves as they become available.
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p mes nig t istant Director for Component's Structures Engineering hivision oLEngineering l
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R. Vollmer, DE Z. Rosztoczy, DE R. Bosnak, DE R. Woods,'IE F. Cherny, DE E. 3ordan, IE
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JOHN J. C REY s
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N The EPRI/PWR Safety and Relief Valve Test program testing ~ activities for the period of May 11 - May 15 wers as follows:
. WYLE The full pressure preload test and the 110 F/650 water seal test on the 0
Dresser relief valve ware perfomed as scheduled last Friday, May 8.
The screening criteria was met for the preload test.
The Dresser valve perfom-ance for the witer seal simulation test was similar to the Target Rock valve performance for the same test condition.
The Dresser valve opened as e'xpected.
Upon de-energizing the valve for closure, the valve remained open until the valve was isolated from the test loop.
Following test valve isolation, the valve closed.
The valve did not pass the screening criteria (failure to close on demand). The valve was removed from the test facility and disassembled by the Dresser Representative. No damage was observed that would affect future testing.
During this water seal simulation test larger than expected bending moments were measured in the upstream and downstream piping. It has been speculated that this resulted from the uneven exhausting of the 1100F water through the downstream ramshead. To eliminate the re-occurence of this during future testing the ramshead has been removed.
The Target Rock valve was reinstalled in the test loop.
A full pressure steam test was performed Wednesday, May 13. The preload test originally scheduled for Monday, May 11, was perfomed Thursday, May 14.
In addition
-- a steam to 6500F water transition test was, perfomed.
For the above tests on the Target Rock valve the screening ' criteria were met.
Resumption of' testing on the 1)resser relief valve is scheduled for H:nday, May 18 Present plans call for retesting the Dresser valve for the water seal simulation test condition.
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Pay 1.1981 W
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DISTRIBUTION FROM:
JOHN J. CAREY f
SUBJECT:
S/RV TEST Ar.TIVIT The EPRI/PWR Safety and Relief Yalve Test Program testing activities for the period of April 27 - May 1 were as follows.
WYLE Testing on tee Target Rock valve resumed this week. On &nday, the low pressure 665 psi,100 F, water test was performed. On Tuesday, the full 0
, pressure 2500 psi, 4500F, water test was perfortned.
For these tests, as
- well as the previous tests, the, Target Rock valve opened and closed as expected. The valve passed the screening criteria. On Wednesday, the full' pressure 2500 psi. loop seal simulation test was perforned. The water just upstream of the valve was 110 F followed by 650 F water.
For this test, 0
the valve opened as expected. Upon de-energizing the valve for closure.
_ _ _the valve recained opened for approximate y 12 seconds and then closed. The l
valve did not pass the screening criteria for this condition (failure to close upon decand). The valve was renoved from the test facility and disassembled by the Target Rock Representative. No darage was observed that would affect future testing.
The valve was re-installed in the test facil ity. The full pressure 2500 psi, 6500F, wa'ter preload test is scheduled for today.
CCM30STION ENGINEERING s
All work on facility construction was completed this week. Pre-test adjustraents will continue through the weekend.
The first full pressure 2500 psi, steam shakedown test is scheduled'for Manday, May 4,1981.
JJC/WJB/ad DISTRIBUTION:
D. Hoffman
- Telecopy f 517-788-0134 J. Scott
- Telecopy i201-430-6734 '
F. Cherny (NRC)
Telecopy (301-492-4994 Panafax set et 6 8M y(b)S J. Turnage R. Newton W. Jones 1 7 Q f./
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JUL 1 1381 Q,: +..._.eces.co,'; Assistant Director for
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Licensing,7 L Thomas H. liovak, Assistant Director for Operating Reactors, DL James P. Y. night, Assistant Director for Com;cnents & Structures Engineer,ing, DE FR0i4:
REPORTII G OF U! SATIS [ACTORY EPRI/FWR TEST 4P0HEliTS, INC. POWER OPERATED RELIEF YALVE' SU3 JECT:
CO:: TROL C0:
AliD DRESSER 14DDEL 31739A SAFETY VALVE 25, 1961 discusses the The attacEed cemorandum from EPRI for the tesek of June l
results of both steam and water tests perforced at Wyle-:orco on the Com;cnents, Inc. PORV and the results of steam tests at the C 1;ote that this is not the same Dresser on the Dresser 31739A Safety Valve.
safety valve discussed in our June 16, 1951 cesorandum. As de I " screening criterion" in EPRI mercrandum, each of these valves f ailed an EFFor each "f ailure",
one er more of these tests.
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as stated in the memorandum.
It is c.;r unterstandinc that the.Licer. sees anc Ccr.struction Ferdi h !!SSS that u;iii:e er plan to utili:e one or bcth of these valves and t e h
- cr.sibility ver.dcrs have been notified of these tests results an c
tr.eir riants.
Our ir.f tr ation fr s E??.} ~ indicates that the Centrb1 Ccm:cnen is being us(d er tiili te used on the following plants:
F.cGuire 1 and 2 Catawba 1 and 2 ihe Dresser 31729A Safety [alve is being used or will te used e plants:
Calver7 Cliffs i and 2 Crystal River 3 7141 - 1 Palisades lii11 stone 2
%!4idland 1 and 2
Oconee 1, 2 and 3 1>U P o f Flp7 %os9 3 -XA
.m a.
.e 2
[resultsisstillbeing Althouch the specific safety significance of these teseva i
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Jarr. esp'.Ynicht',AssistantDirectorfor Cc.gonents &_ Structures Engineering "ivisitn of Engineering v
R. Vollmer cc:
R. Levin F. Cherny E. He minger H. Gregg H. Stolzenberg
- 2. Rosztoczy R. Kiessel E. Jordan E. Brown D. Chaney s
R. Clark S. Varga
. h'. Johnston R. Sesnak e
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June 25,1951 DISTF.IEUi20N ( Attached)
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64 JO?.N J. car.EY
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. ROM:
- EU30ECT:
S/F.V TEST iCTIQilES i
for The EFRI/FR.5afcty and Relief Valve Test Frogrm testing activit
-he peri'o.d of June 22 - 25 were as follows:
gyi :
June 19, a 2500 psia, 450 F vattr test his perferried on the, Cn Friday,ipanents relief valve, utilizing the operator spring fo cc.. trol C5 The valve oper.ed ar.d closed en deCind.
for cIcsure.
e.performsd, a.gs.in' utilizing 1 Sr.turday, Juc.e 20, three addit'ional tests werThe first test was a 500 psia.
the.,cperator sp.ing fo: ce only for c* osere.The se::r.d test das a 500 p 0:
0
~o r The third its: Ots a' 100"? vite.- t est.h.h tss:s, the vr.ive c;st.id tr.d:cl: sed en deand.The U;cr. si;r.alling tl 20 2500 ;sia, 550* F viter test.
~.e vaive for cicture, the velve ver.ained e;ened for approxi.i a y Viive cicsure c:cerred at a valve inle: pressure of 2155 psi crittric re:uirine valve citsece en datzt.d was not si:t.ds.
The uive was'disasserti'ed 'and idste:ted by the CCI va'.va F EFF.I s:res.ir.:
f' arce. 'The Gs darzst ~5.s c':str,e:' that v uid effs:t future valve per crr.
i valve was ratsse:.bied and the syster readiad fcr test ng.
On Weir.e'sf ay,.ere 11, a 2750 csia, stes n estwas perfer: e F.ir pressure to open and cicse the valve.
ds.snd.
The first.
On Tr.ursday, tuc 2500 ksia, 650 F water tests were perforbed.T
- iii:ed air pressure to cpen and close the valve.The se:cnd te Durih; this test the valv.e opened on dectr.d.
4 0 se:cnds.
closed on demnd.
'iiive closure occui red at valve inlet prescura o for closure.
EFF.1 sr.reening th:
The r. ext test on criteria requiring valve. closure on de ar.d was not.ast.
the CCI Calve is schedeied for lionday., June 29..
C. n..: e 1 nh: e hm h. : e..v...,o--
m.
On Tfidry, Jupe 15, a high ramp rate, low back pressure, steam tes The valve epEned at a
- erfor.ed on the Dresser safety valve (3173?A).
A ca.ib.en s*Eri i
3*i cf t.'.e valve design set pressere.
press;;re withir. 1
-,. +,
n,, -,
~
wn-,
4 -
,m
-o-
4 less than 6%
sition of ED% of rated lift wts cbtained at a pressure.The viive re:lo tion for the a.bove test, ttcve.ne vaive dssign set pre,ssure.
d ltini creater thin 2250 psig. -(During test prepara ihe tank prersure upstream of the valve ir.edvertently it. creas resu in t sh:rt d0ratica c.ctettien of the test velve.)
h ratp rete,, low back C.. H:nday, &ne 22 e se:cnd (slightly higher) higThe valve c;ened at a pressu p, essure,.. steam test was ;ierforced, A r.axime. sto position of 73% of f the valve design
-3% of the valve design. set pressurawas obtained a e pn..ssure gretter thso 5% oF.eted flew w
=
7tud lif:
-~ :
. set pressure.
,..cr ater thsn.2250 psig.
diur., back drassure test ka,s,pericre.ed. 0
- je i
tl pressure..,,AmaQwm,'c-Oa isesify.; a high ram. p rite, !r;e The yaive opened %ithin +M' f the, ya)ve des.isn S,e.Gobtaised at t (ressurg p
. :.. sis. psitienMf 57% of r. ate.li.ft vdesOO, set bys'sure.5.Freted'.fi l
M.. f.p.c f phe v4Yve: ".. ciosic it 'a 'pr' essure'gruter thtn 2250
, vy. tpproxir.stely GO psig test vith m:dified t
On k'edr.ssdty, a high ramp rate, iow brek pres:ure, s et:-
The valve opensd Et A pressure Within valve rir.c set.ines was perforr.ed.
A r.t.xime r. ste position of 76% of Fate: iif-was obtained et e. pressuri greater then
-3' cf thi valve Sesien set pressure.
The niva clesed at a pressure F.eted flow wts achieved.*
se pressure.
gret;er then 2 50 pspi.
test, with On Tr.ursd.e.y," a hi gh ramp rate, high bs:k pres sure, stet,m Ine V21e-c:
. of rin-set:ir.cs was performed.
A maximan stem F:sition cf E31 Fa-ed lift was schiived at a pressure gretter thtr
- rigi r.ti+3; of the vilve c'tsIer set pressure.
The EF?.I screer.irg criteria F. acid flow Kas rst s:hieved."
5 psig (tirgEt -
Fetk tz:k pressdre otttined wts i,pprodittely M set crassura.
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was i.o. ret.stetty ste;c trekpressure was E35 prig).
Nddiedfcrtod!Y-Tae next s:e c tese"or the Dresser safety valve is sc i e
l
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, e.=.c oi1 prei'.mim.ry vent'.ri flow data.
1 I
+
..e.
U b m / =h =
(Distribution EttaCh!d) d h og,g M e
'*-e.
pFE.,e' e 68 8b0==
h