ML20195G545
| ML20195G545 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 09/30/1987 |
| From: | Fineman C, Pace N EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY |
| To: | NRC |
| Shared Package | |
| ML20151K679 | List: |
| References | |
| CON-FIN-A-6492, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.D.1, TASK-TM EGG-NTA-7863, NUDOCS 8710140348 | |
| Download: ML20195G545 (20) | |
Text
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EGG-NTA-7863 J
TECHNICAL EVALUATION REPORT TM! ACTION--NUREG-0737 (II.D.1)
DELIEF AND SAFETY VALVE TESTING WATERFORD STEAM ELECTRIC STATION, UNIT 3
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00CKET NO. 50-382 N. E. Pace C. P. Fineman C. L. Nalezny September 1987 Idaho National Engineering Laboratory EG&G Idaho, Inc.
Prepared for the U.S. Nuclear Regulatory Commission Washington D.C.
20555 Under DOE Contract No. OE-AC07-761001570 FIN No. A6492 t
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ABSTRACT Light water reactors have experienced a number of occurrences of improper performance of safety and relief valves installed in the primary coolant system.
As a result, the authors of NUREG-0578 (TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations) and subsequently NUREG-0737 (Clarification of TMI Action Plan Requirements) recommended that programs be developed and completed which would reevaluate the functional performance capabilities of Pressurized Water Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systems for normal, transient, and accident conditions.
This report documents the review of these programs by the Nuclear Regulatory Commission (NRC) and their consultant, EG&G Idaho, Inc.
Specifically, this report documents the review of the Waterford 3 Licensee response to the requirements of NUREG-0578 and NUREG-0737.
This review found the Licensee had provided an acceptable response, reconfirming that the General Design i
Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 have been met.
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FIN No. A6492--Evaluation of OR Licensing Actions-NUREG-0737, II.D.1 11
CONTENTS ABSTRACT..............................................................
ii 1.
INTRODUCTION.....................................................
1 1.1 Background.................................................
1 1.2 General Design Criteria and NUREG Requirements.............
1 2.
PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM................
4 3.
PLANT SPECIFIC SUBMITTAL.........................................
6 4.
REVIEW AND EVALUATION............................................
7 4.1 Valves Tested..............................................
7 4.2 Test Conditions............................................
7 4.3 Valve Operability..........................................
9 4.4 Piping and Support Evaluation..............................
11 5.
EVALUATION
SUMMARY
14 6.
REFERENCES.......................................................
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9 TECHNICAL EVALUATION REPORT TMI ACTION--NUREG-0737 (!!.D.1)
RELIEF AND SAFETY VALVE TESTING WATERFORD STEAM ELECTRIC STATION, UNIT 1 DOCKET NOS. 50-382 1.
INTRODUCTION
1.1 Background
Light water reactor experience has included a number of instances of improper performance of relief and safety. valves installed in the primary coolant system. There have been instances of valves opening below set pressure, valves opening above set press.ure, and valves failing to open or reseat.
From these past instances of improper valve performance, it is.not known whether they occurred because of a limited qualification of the valve or because of basic unreliability of the valve design.
It is known that the failure rf a power operated relief valve (PORV) to ressat was a significant contributor to the Three Mile Island (TMI-2) sequence of events.
These facts led the task force which prepared NUREG-0578 (Reference 1) and, subsequently, NUREG-0737 (Reference 2) to recommend that progrres be developed and executed which would reexamine the functional psrf9rmance capabilities of Pressurized Wcter Reactor (PWR) safety, relief, and block valves and which would verify the integrity of the piping systen.t f6r normal, transient, and accident conditions. These programs were ocemed necessary to reconfirm that the General Design Criteria 14, 15, and 30 of Appendix A to Part 50 of the Code of Federal Regulations, 10 CFR, are indeed satisfied.
1.2 General Design Criteria and NUREG Recuirements General Design Criteria 14, 15, and 30 require that (1) the reactor o
primary coolant pressure boundary be designed, fabricated, and tested so as to have an extremely low probability of abnormal leakage, (2) the reactor coolant system and associated auxiliary, control, and protection systems be designed with sufficient margin to assure that the design conditions are 1
not exceeded during normal operation or anticipated transient events, and (3) the components which are part of the reactor coolant pressure boundary shall be constructed to the highest quality standards practical.
To reconfirm the integrity of overpressure protection systems and thereby assure that the General Design Criteria are met, the NUREG-0578 i
position was issued as a requirement in a letter dated September 13, 1979, by the Division of Licensing (DL), Office of Nuclear Reactor Regulation (NRR), to ALL OPERATING NUCLEAR POWER PLANTS. This requirement has since been incorporated as Item II.D.1 of NUREG-0737, Clarification of TMI Action Plan Requirements, which was issued for implementation on October 31, 1980.
As stated in the NUREG reports, each pressurized water reactor Licensee or Applicant shall:
1.
Conduct testing to qualify reactor coolant system relief and safety valves under expected operating conditions for design basis transients and accidents.
2.
Determine valve expected operating conditions through the use of analyses of accidents and anticipated operational occurrences referenced in Regulatory Guide 1.70, Rev. 2.
3.
Choose the single failures such that the dynamic forces on the safety and relief valves are maximized, i
4.
Use the highest test pressure predicted by conventional safety analysis procedures.
5.
Include in the relief and safety valve qualifiestion program the qualification of the associated control circuitry.
6.
Provide test data for Nuclear Regulatory Commission (NRC) staff review and evaluation, including criteria for success or failure of valves tested.
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Submit a correlation or oth9r evidence to substantiate that the valves tested in a generic test program demonstrate the functionability of as-installed primary relief and safety valves.
Thi-correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the Final Safety Analysis Report (FSAR). The effect of as-built relief and safety valve discharge piping on valve operability must be considered.
8.
Qualify the plant specific safety and relief valve piping and supports by comparing to test data and/or performing appropriate analysis.
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PWR OWNER'S GROUP RELIEF AND SAFETY VALVE PROGRAM In response to the NUREG requirements previously listed, a group of utilities with PWRs requested the assistance of the Electric Power Research Institute (EPRI) in developing and implementing a generic test program for pressurizer safety valves, power operated relief valves, block valves, and associated piping systems.
Louisiana Power & Light, the owner of Waterford 3, was one of the utilities sponsoring the EPRI Valve Test Program. The results of the program, which are contained in a series of reports, were transmitted to the NRC by Reference 3.
The applicability of these reports is discussed below.
EPRI developed a plan (Reference 4) for testing PWR safety, relief, and block valves under conditions which bound actual plant operating conditions.
EPRI, through the valve manufacturers, identified the valves used in the overpressure protection systems of the participating utilities and representative valves were selected for testing. These valves included a sufficient number of the variable characteristics so that their testing would adequately demonstrate the performance of the valves used by utilities (Reference 5).
EPRI, through the Nuclear Steam Supply System (NSSS) vendors, evaluated the FSARs of the participating utilities and arrived at a test matrix which bounded the plant transients for which over pressure protection would be required (Reference 6).
EPRI contracted with Combustion Engineering (CE) to produce a report on the inlet fluid conditions for pressurizer safety and relief valves in CE designed plants (Reference 7).
Since Waterford 3 was designed by CE, this report is relevant to this evaluation.
I Several test series were sponsored by EPRI.
PORVs and block valves were tested at the Duke Power Company Marshall Steam Station located in Terrell, North Carolina.
Additional PORV tests were conducted at the Wyle Laboratories Test Facility located in Norco, California.
Safety valves (SVs) were tested at the Combustion Engineering Company, Kressinger 4
i Development Laboratory, which is located in Windsor, Connecticut. The results of the relief and safety valve tests are reported in Reference 8.
The results of the block valve tests 3re reported in Reference 9.
The primary objective of the EPRI/C-E Valve Test Program was to test each of the various types of primary system safety valves used in PWRs for i
the full range of fluid conditions under which they may be required to operate. The conditions selected for test (based on analysis) were limited to steam, subcooled water, and steam to water transition. Additional objectives were to (1) obtain valve capacity data, (2) assess hydraulic and structural effects of associated piping on valve operability, and (3) obtain piping response data that could ultimately be used for verifying analytical piping models.
Transmittal of the test results meets the requirements of Item 6 of Section 1.2 to provide test data to the NRC.
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3.
PLANT SPECIFIC SUBMITTAL l
A preliminary letter containing Louisiana Power & Light's response to a
the question of overpressure protection system adequacy was submitted on
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July 1, 1982 (Reference 10).
The assessment of the adequacy of the overpressure protection system was submitted December 20, 1982 in Reference 11 by the chairman of the CE Owner's Group. The pressurizer safety valve discharge piping analysis report was transmitted on j
j December 29, 1982 (Reference 12).
A request for additional information was l
transmitted on April 6, 1987 (Reference 13) to which the licensee responded on May 21, 1987 (Reference 14).
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The response of the overpressure protection system to Anticipated l
l Transients Without Sr. ram (ATWS) and the operation of the system during feed and bleed decay heat. removal are not considered in this review.
Neither the Licensee nor the NF.C have evaluated the performance of the system for these events.
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1 Waterford 3 utiltres t[ safety valves in the overpressure pro}tibier,e i
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system.
Both valves ar2 0cesser 31709NA safety valves which are meu ced on ' O,;
short inlet piping without loop seals. The Dresser 31M ANA safety val h was one of the valves testen by EPRI. l'c plant and test valves are idercical; o
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therefore, the test valve is represorttative of the niant( valve.
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The Waterford 3 probsure relief system is, designed witho*;t power 3'
operated relief valves (PORVt). This precludes the need fir PORV block valves in the plant.
Because these two types of Maves are not uset in; ti>s
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Based on the above, the nalvd,'.ested is considerad p h acpil:abit to 1
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criteria of Items 1 and 7 alii entified in Section L7 regarding' J
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7 4.2 Test Conditions, h
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f The valve inlet fluid conditions that bound the overpressure unnsiv<rts for CE designed PWR plants are identified in Reference,7.
The trinaients '
considered in this report include FSAR, extended high presxte ig er.tlon (HPI), and low temperature overpressurization events.
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N, For the SVs only steam discharge was calculated for 'i',d tyr.e 5,
The peak calculated pressure us $3 psia ani tSiaximu.n l
calculated pressurization rate was 104 psi /stc. A maximum calculated k
1 backpressure of 353 psia is developed at che SV outlet.
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has the SVs mounted clow to the pressurizer nozzles, the Orpsssr 31709NA safetyvalvewM1beevalSatedusingtheshortinlekpipingtests. Chi b 4
l Waterford3plantsafetyvalveringsettingsare(-48(wnt),-r,yimidd14,s i
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and 0 [ lower)) an! ( 48
-20, 0), which have been qualified in CEM-127 i,
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Fourteen tests were performed using the short inlet pipe configuration and threa different middle ring positions (Reference 8).
Seven of these tests were done with pertinent ring settings and steam inlet conditions; the ring settings (-48, -60, 0) were used for tests 603, 606, 611, and 618 and the ring settings (-48, -20, 0) were used for tests 615, 620, and 1305.
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test conditions v.,ed in these tests completely bound the predicted Waterford 3 peak sp tem pressure, pressurization rate, back pressure, and valve external moment.
The short inlet piping configuration used in the tests (6.5 ft) is slightly longer than that used at Waterford 3 (4.8 ft) which it' conservative also.
For tests performed at the valve ring settings of (-48, -60, 0) and
(-48, -?3, 0), the set pressure used in the tests (2515 psia) is only
- lightly above the 2500 psia set pressure being used at Waterford 3.
The range of peak pressures obtained during testing was 2505 to 2697 psia comstrea to the predicted peak pressure of 2688 psia, and the range of pressur*zation rate?. used during testirg was 2.7 to 322 psi /sec ccmpared to the calculated ran g of 33 to 104 psi /sec.
The peak backpressure range found during testing was 174 to 530 psia..ompared to the calculated range of 285 to ?53 psia.
Therefore, the Watorford 3 inlet conditions are bounded by
',these tests.
Review of the CE inlet conditions report (Reference 7) showed '5at water did not reach the valve during FSAR transients or an extended high pressure injection (HPI) event.
The cutoff hind for the Waterford 3 HPI i
pumps is below the safety valve.seteoint so that an extended HP! event would not challenge the safety valves.
There was a concern that the extended safety valve blowdown (blowdown grater than 5%) observed during the EPRI tests could result in the pressurizer liquid level increasing to the safety valve inlet.
CE reanalyzed the Loss of Load Event (LOLD) assuming a 20% blowdown. Other conservative assumptions were als: made to maximize pressurizer liquid j
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level swell. The LOLD Event was chosen because it was the design basis for the safety valves in many of the CE designed plants. The 20% blowdown is conservative since the maximum blowdown observed in the applicable EPRI tests was 14.2%. This analysis showed the pressurizer liquid level did not reach the inlet to the safety valves.
Thus, the steam inlet condition was maintained.
Reference 7, Table 5-17, does not list a LOLD transient for Waterford 3, but does list a Feedwater System Pipe Break which appears to be the limiting transient with respect to peak pressure and the time the safety valve retains open.
Further discussion of this is contained in Section 4.3 below.
The test sequences and analyses described above, demonstrating that the test conditions bounded the conditions for the plant valves, verify that Items 2 and 4 of Section 1.2 have been met, in that conditions for the operational occurrences hhve been determined and the highest predicted pressures were chosen for the test.
The part of Item 7, which requires showing that the test conditions are equivalent to conditions prescribed _in the FSAR, is also met.
4.3 Valve Operability As discussed in the previous section, the Dresser 31709NA safety valves at Waterford 3 are required to operate with steam i.7
% conditions only.
The FFRI test program tested the Dresser 31709NA valve for the required raig> of conditions.
The seven steam tests all showed stable performance of the safety valve with it opening at < 3% over the set pressure and reclosing with 7.5 to 14.2% blowdown.
Rated lift was normally obtained and greater than rated flow was always obtained at 3 and 6% accumulation. The percent of rated steam flow measured was 114 to 125% at 3% accumulation and 130% at 6%
accumulation.
Rated flow is 507,918 lbm/hr and the rated lift is 0.588 in.
for the Dresser 31709NA safety valve with a set pressure of 2515 psia.
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A maximum bending moment of 200,000 in-lb was applied to the Dresser 31709NA valve discharge flange during test 1305 without impairing valve operation.
This bounds the maximum expected bending moment of 106,400 in-lb at the plant.
For a test to be an adequate demonstration of safety valve stability, the test inlet piping pressure drop should exceed the plant pressure drop.
This is the situation with Waterford 3.
The plant valves are mounted close to the pressurizer nozzles (4.81 ft) and the Dresser 31709NA safety valve tests were performed with a short inlet piping configuration (6.5 ft) which was longer than the plant piping.
The plant valves inlet pressure is predicted to be less than 94 psi; the calculated values for tests 606 and 1305 are 96 and 103 psi, respectively.
In addition, the Dresser 31709NA safety valve performed stably during the applicable tests. Therefore, it can be concluded that the plant valves will also operate stably.
As noted esrlier, the valve blowdown for the Dresser 31709NA safety valve during the applicable tests ranged from 7.5 to 14.2%. Ths Waterford 3 analysis with 20% blowdown showed that the pressurizer liquid level would not reach the safety valve inlet for the LOLD transient (References 7 and 11). This bounds the blowdown observed in the test. Also, the hot leg remained subcooled during the LOLD transient analyzed with the extended blowdown indicating adequate core cooling was maintained.
NUREG-0737 II.D.1 required qualification of associated control circuitry as part of the safety / relief valve qualification. Waterford 3 does not have PORVs; therefore, this requirement does not apply.
The presentation above demonstrates that the valves operated satisfactorily and verifies that the portion of Item 1 of Section 1.2 that i
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requires conducting tests to qualify the valves and that part of Item 7 requiring that the effect of discharge piping on operability be considered have been met, i
4.4 Piping and Support Evaluation
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The piping from the pressurizer to the safety valve discharge flanges was de'igned and analyzed to the ASME Code,Section III, Class 1.
The safety valve piping and supports between the valve discharge flanges and the quench tank were analyzed to the ANSI B31.1 piping standard and was also seismically analyzed. The load combinations and acceptance criteria were equivalent to those recommended by the EPRI piping subcomittee as presented in Reference 15.
The transient condition analyzed was limited to discharge of saturated steam. The safety valve opening pressure in the analysis was 2574.25 and the flow rate used was 576,720 lbm/h., which is <14% above the ASME rated capacity (504,874 lbs/hr).
In the EPRI/CE tests the Dresser 31709NA safety valve actually disc %arged up to 125% of ASME rated flow at 3% accumulation.
Use of a lower valve flow rate will result in lower calculated valve discharge loads.
However, the calculcted piping strecses are well within the allewables, and the increase from using a larger valve discharge would not result in an overstress condition and is therefore not a safety conc.cn.
The thermal-hydraulic analysis was performed with the program RELAp5/M001. The ability of RELAPS to calculate the thermal-hydraulic transient has been verified through simulations of EPRI/CE SV tests, as shown in Reference 16.
A RELAP5 model for the safety valve piping from the pressurizer to the quench tank was developed. A valve opening time of i
12 msee was used and the control volume size (0.67 to 1.06 ft) used is according to Reference 14 guidolines.
The maximum time step used was 2x10-4 seconas.
This selection satisfies the criteria that no front
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(whether pressure or fluid) may traverse the length of a control volume in one time step. Th3 opening of the two safety valves was assumed to be
' simultaneous; this should produce the worst case loading condition, 11
Significant effort was taken to compare the RELAPS/CALPLOTF III predicted loads with the original design loading conditions which were predicted using the RELAP4 and PIPESHOCK codes in conjunction with CALPLOTF III.
Verification of CALPLOTF III was provided in Reference 14. The verification of CALP:.0TF III was accomplished by comparing computed results against EPRI/CE test 1411.
The comparison was good and verified the capability of the CALPLOT III code for this type of problem.
Based on the above, the results of the RELAP5/CALPLOTF III analysis are considered to be satisfactory.
To confirm the design adequacy a structural model was developed and an analysis performed using the PIPESTRESS 2010 computer code which performs a "generalized response analysis" of the systea which is "subjected to simultaneous application of the transient hydraulic loadings."
If the predicted stresses are too high the code is capable of "selected model superposition time history analysis of pipe segments" to predict less conservative stresses. Details of the functioning of this computer code was provided in Reference 14. PIPESTRESS 2010 is an Ebasco proprietary piping analyis code that performs linear elastic analyses of three-dimensicnal piping systems in accordance with ASME-IIIclasses 1, 2, and 3 and ANSI B31.1 Codes as specified.
PIPISTRESS 2010 has been extensively benchmarked against standard published data and is considered to be a verified code.
The results of the piping analysis showed the pipe stresses to be less than their allowables and the support loads to be less than their ratings.
The Class 1 piping (pressurizer to safety valve discharge) analysis was done according to the ASME Class 1 piping requirements and the safety valve discharge to quench tank piping was analyzed to ANSI B31.1.0.
The determination of piping supports adequacy was based on Ebasco Specification of General Power Piping. Adequacy was based on a support deflection criteria, and allowable stress limits of the individual support component 3 Allowable strtss limits were ASME Sestion III,1971 thru Winter 1972 or, if not acceptable, ASME Section III, Appendix XVII-2200 of the 1974 edition.
In Reference 14, the licensee provided a tabulated comparison of 12 l
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the calculated support loads for the various loading conditions with the maximum allowable loads during the normal, upset, and faulted conditions.
The maximum calculated loads were all within the allowables.
The analysis discussed above with its results which are within the allowables, verifies Item 3 of Section 1.2 has been met and the analysis of che piping and support system verifies Item 8 has also been met.
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EVALUATION SUtHARY The licensee for Waterford 3 has provided an acceptable response to the requirements of NUREG-0737, reconfirming that the General Design Criteria 14, 15, and 30 of Appendix A to 10 CFR 50 have been met with regard to the safety valves. The rationale for this conclusion is given below.
The licensee participated in the development and execution of an acceptable relief and safety valve test program to qualify the operability of prototypical valves and to demonstrate that their operation would not invalidate the integrity of the associated equipment and piping. The subsequsnt tests were successfully completed under operating conditions which, by analysis, bound the most probable maximum forces expected from anticipated design basis events.
The test results showed that the valves tested functioned correctly and safely for all discharge events specified in the test program that were applicable to Waterford 3 and that the pressure boundary component design criteria were not exceeded. Analysis and review of both the test results and the licensee justifications indicated the performance of the test valves and piping can be directly applied to tha in-plant valves and piping.
The plant specific pioing also has been shown by analysis to be acceptable.
Thus, the requirements of Item II.D.1 of NUREG-0737 have been met (Items 1-8 in Paragraph 1.2) and, thereby, ensure that the reactor primary coolant pressure boundary will have a low probability of abnormal leakage (General Design Criterion No. 14).
In addition, the reactor primary coolant pressure boundary and its associated components (piping, valves, and supports) have been designed with a sufficient margin so that design conditions are not exceeded during relief / safety valve events (General Design Criterion No. 15).
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Further, the prototypical tests and the successful performance of the valves and associated components demonstrated that this equipment was constructed in accordance with high quality standards, meeting General Design Criterion No. 30.
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REFERENCES 1.
TMI-Lessons Learned Task Force Status Report and Short-Term Recommendations, NUREG-0578, July 1979.
2.
Clarification of TMI Action Plan Reauirements, NUREG-0737, November 1980.
3.
R. C. Youngdahl ltr. to H. D. Denton, Submittal of PWR Valve Test Report, EPRI NP-2628-SR, December 1982.
4.
EPRI Plan for Performance Testing of PWR Safety and Relief Valves, July 1980.
5.
EPRI PWR Safety and Relief Valve Test Program Valve Selection / Justification Report, EPRI NP-2292, December 1982.
6.
EPRI PWR Safety and Relief Valve Test Program Test Condition Justification Report, EPRI NP-2460, December 1982.
7.
Valve Inlet Fluid Conditions for Pressurizer Safety and Relief Valves in Combustion Engineering-Design Plants, EPRI NP-2318, December 1982.
8.
EPRI PWR Safety and Relief Test Program Safety and Relief Valve Test Report, EPRI NP-2628-SR, December 1982.
9.
EPRI/ Marshall Electric Motor Operated 81uck Valve, EPRI NP-2514-LO, July 1982.
10.
Letter L. V. Maurin, Louisiana Power & Light Co., to R. L. Tedesco, NRC, "Waterford 3 SES, Docket No. 50-382, NUREG-0737, Safety Valve Functionability," W3P82-1800, 3-A19.09.04.03, Q-3-B31.V2, I.02.02, j
L.09, July 1, 1982.
11.
Letter R. W. Wells, Chairman, CE Owners Group, to Harold Bernard, NRC, "Transmittal of CEN-227, Summary Report on the Operability of Pressurizer Safety Valves in CE Designed Plants," December 20, 1982.
12.
Letter L. V. Maurin, Louisiana Power & Light Co., to T. M. Novack, NRC, "Waterford SES 3 Docket No. 50-382, NUREG-0737, SER Confirmatory Issue, Performance of PWR Relief and Safety Valves," W3P82-4011, 3-A19.09.04.03, Q-3-B31.V2, I.02.02, L.09, December 29, 1982.
13.
Letter J. H. Wilson, NRC to J. G. Dewease, Louisiana Power & Light Co., "Request for Additional Information, NUREG-0737 Item II.D.1, Performanca Testing of BWR and PWR Relief and Safety Valves,"
April 6, 1987.
14.
Letter K. W. Cook, Louisiana Power & Light Co., to NRC Document Control Desk, "Waterford SES Unit 3, Docket No. 50-382, NUREG-0737, Item II.D.1, Response to NRC Ouestions," W3P87-1063, A4.05, QA, May 21,
- 1987, 16
15.
EPRI PWR Safety and Relief Valve Test Program Guide for Application of Valvo Test Program Results to Plant-Specific Evaluations, Revision 2, Interim Report, July 1982.
- 16. Application of RELAP5/ MOD 1 for Calculation of Safety and Relief Valve Discharge Piping Hydrodynamic Loads, EPRI-2479, December 1982.
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