ML20053B305
| ML20053B305 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 05/25/1982 |
| From: | Bridenbaugh D, George Minor MHB TECHNICAL ASSOCIATES, SUFFOLK COUNTY, NY |
| To: | |
| Shared Package | |
| ML20053B272 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8205280275 | |
| Download: ML20053B305 (42) | |
Text
4 ki UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
'i ?
, '.2 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD i'
)
In the Matter of
)
)
LONG ISLAND LIGHTING COMPANY
)
)
Docket No. 50-322 (0.L.)
(Shoreham Nuclear Power Station,)
Unit 1)
)
)
PREPARED DIRECT TESTIMONY OF DALE G. BRIDENBAUGH AND GREGORY C. MINOR ON BEHALF 0F SUFFOLK COUNTY REGARDING SUFFOLK COUNTY CONTENTION 22 SRV TEST PROGRAM May 25, 1982 820 5 2 8 0 Q ~) $ [
~.
SUMMARY
OUTLINE OF SUFFOLK COUNTY CONTENTION 22 Suffolk County contends that LILCO has not adequately demonstrated the reliability of the Safety / Relief Valves (S/RV's) used at Shoreham.
This is a safety concern because faulty S/RV's could create or extend a loss-of-coolant-accident (LOCA).
It is also possible that a S/RV failure could occur in a non-detectable mode, lending to upset conditions and safety system challenges when the valve later was called upon to operate.
A long history of S/RV reliability problems, combined with the events of the accident at Three Mile Island (TMI),
prompted the NRC in NUREG-0737,Section II.D.1, to require all operating reactors and license applicants to investigate the reliability of their S/RV's to assure that the valves performed adequately.
To comply with this requirement, LILC0 joined the BWR Owners' Group, which appointed General Electric (GE) to coordinate one generic test program for BWR S/RV's that would be applicable to all BWR plants.
GE's program included testing of the Target Rock two-stage 6R10 type of S/RV Model No. 7567F, which is employed at Shoreham, and found the valve to be operable and able to maintain structural and pressure integrity under the GE program.
Thus, LILCO reported that it had met the requirements of NUREG-0737.
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Despite LILCO's position, however, it has failed to fully meet the NUREG-0737 requirement demonstrating the-reliability of Shoreham S/RV's.
There has been no indication that LILCO has conducted a plant specific analysis comparing the piping configuration, structures, controls and instrumentation used at Shoreham to those used in GE's test program.
Such an analysis is the only way to fully assure the reliability of the Shoreham S/RV's.
Therefore, the witnesses believe LILCO should conduct a detailed plant specific evaluation of Shoreham S/RV's, piping and supports in full accordance with NUREG-0737 requirements to verify their reliability and assure the health and safety of the public.
Attachments:
1.
"An Analysis of the Reliability of Light Water Reactor Power-Actuated-Pressure-Relieving Valves and Safety (Relief) Valves and Their Component Parts Using the Nuclear Plant Reliability Data System (NPRDS) - Final Report".
Southwest Research Institute, November 16, 1981.
pp.
1-4, B11-B18.
2.
NUREG-0737 " Clarification of TMI Action Plan' Requirements,"
Section II.D.1.
ii
3.
" Analysis of Generic BWR Safety / Relief Valve Operability Test Results".
General Electric, October, 1981.
pp. 57-59, 78-83.
f i
l l
iii f
PREPARED DIRECT TESTIMONY OF DALE G. BRIDENBAUGH AND GREGORY C. MINOR REGARDING SUFFOLK COUNTY CONTENTION 22 SRV fEST PROGRAM I.
INTRODUCTION 1.
This testimony was jointly prepared and edited by Dale G. Bridenbaugh and Gregory C. Minor.
A statement of the qualifications of Messrs. Bridenbaugh and Minor has been separately provided to this Board.
II.
STATEMENT OF CONTENTION 2.
The purpose of this testimony is to address suffolk County Contention 22 as admitted by the Board as follows:
Suffolk County contends that LILCO has not adequately demonstrated that the safety / relief valves to be used at Shoreham meet the require-ments of 10 CFR 50, Appendix A, GDC 4 and 30, and 10 CFR 50, Appendix B, Sections III and XI, in that the functionability of the valves, as-t l
installed, has not been established by the generic l
test program results.
Specifically, NUREG-0737, item II.D.1, performanr ? testing of BWR relief and safety valves, requires that BWR SRV valves be tested to demonstrate that the valves will open and reclose under the expected flow conditions.
It additionally requires that ATWS testing be considered.
LILCO has not yet provided a detailed plant specific evaluation of the Shoreham safety and relief valves, piping, and supports in accordance with the NUREG-0737 requirements.
Additionally, no commitment has been made on ATWS testing.
The re fore, it has not been demonstrated at this time that the specific requirements have been met.
The results of our review of some of the ir.portant matters encompassed by this Contention are summarized in the following paragraphs.
III.
DISCUSSION OF ISSUES III.A.
BACKGROUND AND
SUMMARY
OF POSITION 3.
The essence of Contention 22 is that Safety / Relief Valves (S/RV's) used at Shoreham have not been proven reliable over the full range of operating and accident conditions. The S/RV's, in fact, may fail in a mode that could either create or extend a loss-of-coolant-accident (LOCA).
Furthermore, it is entirely possible that if such an event occurred, the status of the problem would not be known to the plant operators because there are no failure detectors / indicators on the S/RV's that would indicate passive failure.
This concern for reliability of S/RV's emerged particularly from the accident at Three Mile Island (TMI) and prompted the NRC to require all operating reactors and operating license applicants to investigate the I
l reliability of their S/RV's to assure that the valves perform adequately.
While the Shoreham BWR will not be subject to the I
same failure sequence as that experienced at TMI, there is reason...
e for serious concern because of the numerous cases where S/RV's in general, and particularly Target, Rock S/RV's (the type used at Shoreham), have failed to close after being operated.
Exampics of Target Rock S/RV failures as reported to the Nuclear Power Reactor Data System, are enclosed herein as.
Target Rock valves were the subject of specific consideration in the Southwest Research study because "they have been identified as causes of unscheduled outages with a 1/
frequency high enough to be of concern..."
4.
NUREG-0737,Section II.D.1, required that performance testing of S/RV's be conducted and an associated report be submitted to the NRC by October 1, 1981.
This requirement is attached herein as Attachment 2.
Thus, the NRC required that submitted information include:
a) evidence that the valves would open and reclose under the expected flow conditions; b) dccumen-tation from each licensee and applicant substantiating that the results for the valves tested in generic test program were applicable to the in-plant valves; c) demonstration of the l
integrity of the discharge piping and supports for expected load conditions; and d) test data, including criteria for success and failure of valves tested, for the purpose of NRG Staff review and evaluation.
In addition, it required test config-urations suitable for testing of the S/RV's under ATWS conditions.
t I l l
(The TMI Action Plan specified no date for completion of the ATWS testing but it c1carly states that the test facility be designed to accommodate such conditions.)
5.
To comply with NUREG-0737, LILCO joined the BWR Owners' Group which was formed to combine the efforts of BWR owners by preparing and conducting one generic test program for BWR S/RV's that would be applicable to all BWR plants.
On behalf of this group, General Electric (GE) conducted the 2/
investigation and submitted its findings in October,1981.
6.
GE's analysis included testing of the Target Rock 2-stage, 6R10 type of S/RV, Model No. 7567F.
This particular valve was found to be operable and able to maintain structural and pressure integrity under the GE test program.
Based on the fact that this is the type of S/RV employed at Shoreham, LILCO reported that the operational adequacy of the S/RV's for the Shoreham station had been demonstrated.E/
7.
In response to a Suffolk County discovery request, LILCO provided a' copy of the GE generic test program report. A!
This non-proprietary version of the full test report (NEDE-24988-P) 5/
was transmitted via B.
R. McCaffrey's March 5, 1982 letter.
The non-proprietary version of this report is particularly inscrutable.
All of the test data have been omitted and only very general statements remain.
We are enclosing as Attachment 3 a copy of pages 78 through 83 and 57 through 59 of this report. l
These pages supposedly summarize the Test Results (6.2).
An examination of the Attachment shows that blind acceptance of LILCO's claimed results are required if this report is to be the verification of the SRV tests.
It is reported, for example, on page 80 that the stresses measured in (some) water tests were higher than those measured in the corresponding steam tests, s/
This is discounted by stating that in plant pressur-ization rates will be " slower" and that the stress levels "are low".
We have no way of judging the truth of these claims with the information provided.
8.
With regard to suitability of the tests performed to demonstrate ATWS performance capability, Question 3 of Appendix A (NED0-24988) is significant.
That question requests verifi-cation that the safety valve qualification shall include qualification of the associated control circuitry.
The Owners Group response states that the tests include all associated l
valve actuation circuity "which might be affected by the dynamic loads imposed on the plant as a result of the valve actuation under the test conditions." 1/
No mention or claim is made concerning the environmental condition's effect on the valve circuitry.
Such conditions would likely be significantly impacted by ATWS conditions.
9.
Despite LILCO's report of completion of this task, it has failed to fully meet the requirements of NUREG-0737 i
to demonstrate the reliability of Shoreham S/RV's.
- First, there has been no indication that LILCO has conducted a plant specific analysis comparing the piping configuration, structures, controls and instrumentation used at Shoreham to those used in GE's test program.
This is contrary to the NUREG-0737 requirement which provides :
Since it is not planned to test all valves on all plants, each licensee must submit to NRC a correlation or other evidence to substantiate that the valves tested in the EPRI (Electric Power Research Institute) or other generic test program demonstrate the function-ability of as-installed primary relief and safety valves.
This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the final safety analysis report (FS AR).
The effect of as-built relief and safety valve discharge piping on valve operability must also be accounted for, if it is different from the generic test loop piping.
8/
Second, neither LILCO nor the BWR Owners' Group have taken any steps to conduct S/RV testing under ATWS conditions.
In fact LILCO even states that "no ATWS conditions are required" EI in the testing of S/RV's.
10.
The NUREG-0737 requirements are especially important given their applicability to Shoreham.
Indeed the accident at TMI-2, from which this requirement emerged, clearly involved the funct'ionability and reliability of relief valves in the system.
Furthermore, BWR's are greatly dependent upon relief valves for pressure relief, ADS, and emergency core cooling.
m O
during tr 'sients, accidents and ATWS conditions.
Because LILCO's
'"se to the NUREG requirements was incomplete.,
there -
irance that the Shoreham S/RV's are suitably relial
.at the public health and safety are fully prote III.B.
CONCLUSION 11.
Based on the above, we believe that LILCO has failed to adequately demonstrate at the present time that the S/RV's used at Shoreham fully meet :NRC requirements.
In our opinion, the only way to fully assure the reliability of Shoreham S/RV's is for LILCO to conduct a detailed plant specific evaluation of Shoreham S/RV's, including their control, instru-mentation, piping and supports in full accordance with NUREG-0737 requirements.
Additionally, LILCO should test the valves under ATWS conditions or provide justification for not doing so.
1 REFERENCES 1/
Southwest Research Institute, "An Analysis of the Reli-
~
ability of Light Water Reactor Power-Actuated Pressure-Relieving Valves and Safety (Relief) Valves and Their Component Parts Using the Nuclear Plant Reliability Data System (NPRDS) - Final Report", November 16, 1981, p. 1,
-2/
General Electric, " Analysis of Generic BWR Safety / Relief
~
Vcive Operability Test Results, October, 1981.
3/
LILCO Letter to NRC, 7/21/81, SNRC #600.
-4/
NED0-24988, Analysis of Generic BWR Safety / Relief Valve Operability Test Results, October, 1981.
5,/
B.
R. McCaffrey to Gregory C. Minor, March 5, 1982, Letter 82-9.
6/
Ibid 1/, p.
80.
-7/
Ibid 1/, Appendix A, Justification of Applicability of Test Results to In-Plant S/RV's.
(Emphasis Added) 8/
NUREG-0737, " Clarification of TMI Action Plan Requirements",
Section II.D.1., p.
II.D.1-2.
9/
LILCO Answers to SOC Interrogatories, 3/17/82, p. 11.
e 4 I
t 1
i ATTACHMENT 1 "AN ANALYSIS OF THE RELIABILITY OF LIGHT WATER REACTOR POWER-ACTUATED PRESSURE RELIEVING VALVES AND SAFETY (RELIEF) VALVES AND THEIR COMPONENT PARTS USING THE NUCLEAR PLANT RELIABILITY DATA SYSTEM (NPRDS) - Final Report" pp. 1-4, B11-B18 4
e i
I l
- j
)
AN ANALYSIS OF THE RELIABILITY OF LIGHT WATER REACTOR POWER-ACTUATED PRESSURE-RELIEVING VALVES AND SAFETY p.e (RELIEF) VALVES AND THEIR COMPONENT PARTS USING Ud THE NUCLEAR PLANT RELIABILITY DATA SYSTEM (NPRDS)
!'M
$q FINAL REPORT r-Prepared for U.
S.' Department of Energv Light Water Reactor Safety Technology Management Center
.jp Sandia National Laboratories y/
Albucuerque, New Mexico 87185 (j.
'L.
Sponsored by C.
S.
Department of Energy j,
Division of Nuclear Power Development S
Washington, D. C.
20545 i:
k<orkperformedunderSandiaContractNo. 61-6634 i
SWRI Project No. 17-6649 I,.,
Submitted:
November 1981
{-
Printed:
February 1982 m
I b
.I Autnors:
Approved by:
j Benjamin M. Tashjian Project Manager I
+
Edgar Ocikers, Jr.
/ /'
Consultant
> - /'
//'
d6 l
Andrew G. Pickett Rawley D. Johr(son
/
l Senior Research Engineer Director of Information Sciences A. S harif.Homayoun Quality Assurance Systems Consultant and Engineering Division w
ta anp 8 SO UTHW EST R ES E A RCH INSTITUTE w
s.
/
UK
% _//
SAN ANTONIO HOUSTO N 1
T l
..,f iI l.0 TASK DEFINITION qi
".p f The basic work effort consisted of abstracting data from the NPRDS on the population and malfunction events (failures) of safety, relief, and power-actuated pressure-relieving valves in nuclear steam supply systems (NSSS) and the statistical analysis of these data.
.,y The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel
,y, Code Section III, Division 1, Appendix Article 0-1000 provides the following
[!l definitions of these items:
- E TI b
U' (1)
Safety Valve. An automatic pressure-relieving device actuated by the g.
static pressure upstream of the valve and characterized by full opening
{e.
v.
pop action.
It is used for gas or vapor service.
T :q djh i
Vi';
(2)
Relief Valve. An automatic pressure-relieving device actuated by the 5 l static pressure upstream of the valve which opens further with the o increase in pressure over the opening pressure. It is used primarily for k: ,jd. liquid service. n l (3) Safety Relief Valve. An automatic pressure-actuated relieving device
- . ft suitable for use either as a safety valve or relief valve, depending on l
i application. 't g (4)' Power-Actuated Pressure-Relieving Valve. A relieving device whose ' */l movements to open or close are fully controlled by a source of power (electricity, air, steam, or hydraulic). The valve may discharge to I 4 atmosphere or to a container at lower pressure. The conditions, and such ] effects, shall be taken into account. If the power actuated pressure-I; relieving valves are also positioned in response ' to other control sig-nals, the control impulse to prevent overpressure shall be responsive only to pressure and shall override any other control function. Three valves of specific interest are considered herein because their data base is included in NPRDS and they have been identified as causes of unscheduled not anes wi t h a t re.pn ncy hi gh ennunh in he ni cimei rni T 1
(1) The spring-loaded safety (relief) valve manufactured by Crosby, Dreuuur,, Crane, and others and. designated as safety. (2) The pilot-operated, pressure-relief (safety) valve, which can also be power-actuated, manufactured by Target Ror' and designated as Target Rock. (3) The power-actuated pressure-relief (safety) valve, which can also be r actuated in response to a system pressure transducer signal, manufactured by Dresser and derignated as Electromatic. The spring-loaded safety (relief) valves are ubiquitous because they must be installed for overpressure protection in every system (or component) that is or can be irolated while temperature is increased or that may be exposed to over-pressure from other causes. These valves are passive, and malfunction is detected only if it is a leak, low pressure actuation, or surveillance test event. This analysis considered only those spring-loaded safety valves in main steam service because there is sufficient data for these valves, and not for others, in NPRDS. The Target Rock pilot-operated, power-actuated valve is reported only in the boiling water reactor (BWR) main steam system. It may serve as a passive safety valve or be part of the pressure reductiot systems and procedures and is actuated automatically or manually. The Electromatic valve is reported in some BWR main steam systems and on some pressurized water reactor (PWR) pressurizers as part of the pressure reduction ~ systems and procedures and is actuated automatically or manually. The valves in these three categories compose a significant source of outages and plant extended outage time and maintenance problems according to previous surveys and plant personnel interviewed. Also, these are the valves most likely to be included in NPRDS submittals and most likely to provide an ade-quate data base for statistical analysis. l 2
j,
- y.n
't ? o.. G.i 2.0 PROGRAM OBJECTIVES
- Lg M<
t>. Previous surveys and analyses of NSSS valve failure (malfunction)(1-7) have .'y been performed, which include reliability and failure " root-cause" studies. jT nM These were reviewed to obtain information and reduce duplication of effort. It was noted that, while failure data have been tabulated according to failure l:.Q t.p. i mode (such as leak) or part failure rate (such as pilot), they had not been y, .7 correlated with the population and basic functional elements, nor the design, a quality control, and preventive maintenance variables of these functional p-{ .!5 elements. l
- er d
Jiid One objective was to make a comparison of reliability functions and valve mod-R:, ification history to determine if changes in design had any effect on valve
- 4.. "
k reliability and, specifically, if the valve parts involved had relatively high reliability. Part of the effort was to determine if many of the malfunctions attributed to design deficiency were the result of inadequate quality control
- {
- 7 l
in manufacture, installation, or maintenance. Acother objective was to help j h.h increase the time-to-failure, because the cost in both money and radiation ! 'N exposure during primary coolant system valve maintenance and modification is b' appreciable. The approach for achieving this objective was the use of I reliability functions to choose from among optional valee modifications. ! y A A primary objective of.this task was to perform a reliability analysis of the ,C valves considered to identify root causes and remedial actions. Such an analy-sis considers a valve and its directly associated components as a system. The functions of individual parts are elements of reliability, whose collective a l reliability constitutes system (valve) reliability. Individual part function l can be accomplished by the use of different mechanisms, and part performance is 'l ' I' affected by design details, quality control, and preventive maintenance prac-l tices. Consequently, reliability comparisons can be made between the dif ferent mechanisms used to perform a function and part design details, quality control, i and preventive maintenance practice used in the construction and operation of I valves to accomplish the required part functions. The shapes of failure rates + 1 and also reliability curves are useful in assessing failure cause (random or constant-hazard, wear-out, fabrication quality control, or human factors). l 1 c i
? This method of data presentation, in addition to providing information useful, to increasing valve reliability by redesign or procedure change, will benefit-the NPRDS program as suggested improvements are implemented in reporting proce-dures, as well as defining supplementary information that needs to be acquired for a complete reliability analysis. Specific differences in manufacturer, unique plant design details, quality control, preventive c:aintenance practice, and inservice modifications are examples of supplementary information needed to g explain differences in failure experience both in time and between plants. j Reviews of United States Nuclear Regulatory Commission (NRC) regulatory guides. ( t ( bulletins, circulars, notices, and NSSS service information letters indicate that unreported valve modifications have occurred that should change failure h rates. Valve malfunctions that were detected in bench tests during scheduled outages also generally fall outside the scope of reporting to NPRDS. ? I i According to the Rasmussen Report (WASH-1400),(8) an important missing factor f in probability-based safety analyses of nuclear power plants is the actual reli-ability function for components such as valves. The limited number of nuclear I; power plants, relatively short service experience, and diversity of designs used limit the statistical sample. However, this effort also can evaluate the [ usefulness of the reliability analysis approach to obtain component reliability, functions with such a limited sample. i l In summary, the objectives were: l l .(1) To perform typical reliability analyses of selected NSSS components. (2) To learn what the NPRDS data are indicating concerning performance of valves in service. (3) To determine the adequacy of the NPRDS data base sample for improvement of reliability functions for safety analysis. (4) To identify reliability critical design details, parts, maintenance prac-tices, etc. (" root causes of valve failures"), to aid in increasing valve reliability by selecting remedial actions to reduce the number of fas due to these most frequent causes. 4
~ TABLE B-5. TARGET ROCK TilREE-STAGE MODEL 67F FAILURE EVENTS flassification Component Event Date [ Category Reactor Location Day /Mo/Yr EFPY Cause/ Comment .A.1.a.(2) llatch-1 11. 20/05/78 0.4 Cracks developed in seat weld .A.I.c. Pilgrim 10 23/04/75 1.2 Leak, oxidation cleaned, lapped .A.l.c. Pilgrim 133 29/04/75 1.3 Leak, oxidation cleaned, lapped .A.I.c. Fitzpatrick A 31/03/76 0.45 .A.l.c. Brown Ferry-1 19 05/74 0.0 Wire drawn main seat .A.1.c.(2) Mi -I C 22/05/75 2.4 Foreign material under seat .A.I.c.(2) Br.nswiu-2 K 16/07/76 0.4 Light steam cuts on disc, dirt on seat .B.I.d.(3) Monticello A 09/71 0.2 Main disc steam galled .B.1.d.(3) Monticello B 09/71 0.2 Main disc steam galled .B.I.d.(3) Monticello D 09/71 0.2 Main disc steam galled .D.2.b.(3) Monticello A 07/73 0.8 .D.2.b.(3) Honticello B 07/73 1.2 Condensate collection behind main valve piston .D.2.b.(3) Monticello C 07/73 1.7 Condensate collection behind main valve piston .D.2.b.(3) Monticello D 07/73 0.6 Condensate collection behind main valve piston .D.2.b.(3) Monticello E 07/73 1.45 Condensate collection behind main valve piston .D.2.b.(3) Monticello F 07/73 1.45 condensate collection behind main valve piston .D.2.b.(3) Monticello G 07/73 1.45 condensate collection behind main valve piston .D.2.b.(3) Monticello 11 07/73 1.45 Condensate collection behind main valve piston
TABLE B-5. TARGET ROCK TilREE-STAGE MODEL 67F FAILURE EVENTS (Cont'd) Classification Component Event Date Category Reactor Location Day /Mo/Yr EFPY Cause / Comment II.A.I.c. Monticello A 23/02/77 0.6 Crud on second stage disc and seat II.A.l.c. Browns Ferry-3 4 28/08/78 1.3 Leakage II.B.1.b. Peach Bottom-2 A 12/73 0.1 Called steam binding in bushing II.B.l.b. Peach Bottom-2 D 01/74 0.3 Called steam binding in bushing II.C.1.b.(2) Pilgrim 8 05/77 2.08 Delamination air piston diaphragm II.C.I.b.(2) Pilgrim 10 05/77 0.85 Delamination air piston diaphragm IIcc l.b.(2) Pilgrim 116 10/05/77 0.75 Delaminated diaphragm II.D.2.a.(1) Pilgrim 116 09/73 0.62 Broken air pipe nipple II. D.2. b. (2 ) Monticello A 07/72 0.45 Rust particles could have plugged orifice III.A.I.b. Millstone-1 F 26/u2/79 2.05 Disc steam cut III.A.I.c. !!atch-1 A 06/10/77 2.2 III.A.I.c. llatch-1 E 02/77 0.1 III.A.L.c. Hatch-1 G 09/01/77 1.9 Pilot leak III.A.I.c. Ila tch-1 K 02/77 0.05 III.A.l.c. lIn tch-1 L 01/02/77 1.95 IlloA.l.c. Millstone-1 A 03/78 4.5 Steam cut III.A.1.c. Millstone-1 B 20/05/75 0.8 Pilot blowby III.A.l.c. Hillstone-1 B 01/74 1.6 Worn preload spacer
TABLE B-5. TARGET ROCK THREE-STAGE MODEL 67F FAILURE EVENTS (Cont'd) Classification Component Event Date Category Reactor Location Day /Mo/Yr EFPY Cause / Comment III.A.I.c. Millstone-1 D 17/06/77 3.85 Pilot leak, steam cutting III.h.l.c. Millstone-1 F 06/76 3.15 Pilot leak III.A.l.c Monticello A 22/11/78 0.6 Pilot did not seat correctly, steam cutting III.A.l.c. Monticello D 04/05/78 3.35 Pilot steam cutting III.A.I.c. Monticello E 05/11/74 0.75 Leaks, foreign material III.A.I.c. Monticello F 23/02/77 2.4 Steam eroded III.A.I.c. Monticello G 11/11/74 0.75 Foreign material on pilot seating surfacee III.A.l.c. Monticello 11 23/11/77 2.7 Steam cutting III.A.I.c. Monticello 11 01/12/78 1.05 Crud on pilot seat III.A.I.c. Pilgrim 9 09/75 1.55 Steam cutting III.A.I.c. Pilgrim 9 10/78 1.4 III.A.l.c. Pilgrim 10 14/11/77 0.3 Pilot leak Ill.A.I.c. Pilgrim 10 12/72 0.15 III.A.I.c. Pilgrim 10 10/78 0.5 III.A.I.c. Pilgrim 116 20/07/75 0.83 Pilot valve leaka'ge III.A.l.c. Pilgrim 116 09/73 0.62 III.A.I.c. Pilgrim 133 17/11/77 1.2 III.A.I.c. Brunswick-1 F 13/03/78 0.65 Pilot leak ~
TABLE B-5. TARGET ROCK TilREE-S' AGE MODEL 67F FAILURE EVENTS (Cont'd) Classification Component Event Date Category Reactor Location Day /Mo/Yr EFPY Cause/ Comment III.A.l.c. Brunswick-1 G 01/81 2.0 III.A.l.c. Brunswick-1 J 14/04/79 1.2 Pilot leak III.A.l.c. Brunswick-2 A 16/07/76 0.4 Steam cuts due to wear and dirty seat III.A.1.c. Brunswick-2 B 16/07/76 0.36 Dirty and pitted pilot dise III.A.I.c. Brunswick-2 E 07/79 1.8 Leak, steam cutting III.A.l.c. Brunswick-2 G 16/07/76 0.4 Steam cuts and some dirt between disc and seat III.A.l.c. Peach Bottom-2 D 06/01/77 1.33 Pilot leak ? Z III.A.I.c. Peach Bottom-2 F 14/11/76 1.6 III.A.I.c. Peach Bottom-2 K 06/01/77 0.9 Pilot leak III.A.I.c. Peach Bottom-2 K 06/75 0.82 III.A.I.c. Peach Bottom-2 L 04/11/74 0.55 Pilot valve disc leakage, machined, lapped III.A.l.c. Peach Bottom-3 B 12/07/76 1.05 Pilot leak III.A.l.c. Peach Bottom-3 E 07/76 1.05 Pilot leak, open III.A.1.c. Peach Bottom-3 E 12/76 0.3 III.A.1.c. Peach Bottom-3 F 12/76 1.35 Pilot leak, open III.A.1.c. Peach Bottom-3 G 20/07/76 1.05 Pilot leak III.A.l.c. Peach Bottom-3 L 13/06/79 2.95 III.A.l.c. Fitzpatrick B 22/11/76 0.9
TABLE B-5. TARCET ROCK THREE-STACE MODEL 67F FAILURE EVENTS (Cont'd) lassification Component Event Date Category Reactor Location Day /Mo/Yr EFPY Cause/ Comment III.A.1.c. Fitzpatrick E 11/76 0.25 FII.A.1.c. Fitzpatrick 'E 07/76 0.65 II.A.l.c. Fitzpatrick F 19/11/76 0.9 II.A.I.c. Browns Ferry-1 22 26/02/75 0.05 Leaks, wire drawn II.A.I.c. Browns Ferry-1 23 26/02/75 0.05 Leaks, wire drawn II.A.l.c. Browns Ferry-1 4 26/02/75 0.05 Leaks, wire drawn II.A.I.c. Browns Ferry-1 5 26/02/75 0.05 Leaks, wire drawn II.A.1.c. Browns Ferry-1 5 21/04/77 0.3 II.A.1.c. Browns Ferry-2 5 13/02/78 1.05 II.A.L.c. Browns Ferry-3 30 17/08/78 1.3 ~ II.A.l.c. Browns Ferry-3 31 21/04/77 0.35 II.A.I.c. Browns Ferry-3 34 17/08/78 1.3 -II.A.l.c. Browns Ferry-3 41 17/08/78 1.3 II.A.l.c.(2) Peach Bottom-2 E 16/10/74 0.55 Pilot valve disc leakage, machined, lapped
- II.A.l.g.
11a tch-1 C 01/02/77 0.05 Did not reseat II.A.l.g. Browns Ferry-2 41 0S/02/78 1.05 Did not reseat
- II.A.l.g.
Browns Ferry-3 31 15/04/78 0.75 Did not reseat II.B.I.d.(3) Monticello A 14/06/76 1.8 Overtightening of solenoid plunger
TABLE B-5. TARGET ROCK Ti!REE-STAGE MODEL 67F FAILURE EVENTS (Cont'd) ' Classification Component Event Date Category Reactor Location Day /Mo/Yr EFPY Cause /Cocunent
- III.C.1 Ita tch-1 C
09/08/77 2.2 Setpoint set incorrectly III.C.1 Ilatch-1 D 06/10/77 2.3 Setpoint drift
- III.C.1 liatch-1 E
06/10/77 0.4 Setpoint dr'ift III.C.1 llatch-1 .G 06/10/77 0.35 Setpoint drift III. Col IIntch-1 11 06/10/77 2.3 Setpoint drift .III.C.1 llatch-1 J 06/10/77 2.3 Setpoint drif t .III.C.1 Ilatch-1 K 06/10/77 0.35 Setpoint drift
- III.C.1 Itatch-1 K
03/01/77 1.9 Setpoint drift III.C.1 Honticello G 20/09/77 2.05 Serpoint drift
- III.C.1.b.(3)
Monticello D 07/72 0.6 IIII.C.I.b.(3) Monticello E 10/05/77 1.8 Bellous O ring leak suspected III.C.1.b.(3) Peach Bottom-2 C 03/74 0.22 Bellows Icaks III.C.1.b.(3) Peach Bottom-2 D 03/74 0.1 Bellows leaks
- III.C.1.b.(3)
Peach Bottom-2 C 03/74 0.22 Bellows leaks III.C.L.b.(3) Peach Bottom-2 11 03/74 0.4 Bellows leaks III.C.1.b.(3) Peach Bottom-2 J 26/10/74 0.55 Bellows leaks III.C.1.b.(3) Peach Bottom-2 K 10/73 0.03 Bellows leaks III.C.l.c.(4) Pilgrina a 11/72 0.12 Nitrogen setpoint
TABLE B-5. TARGET ROCK TilREE-STAGE MODEL 67F FAILURE EVENTS (Cont'd)
- 1assification Component Event Date Category Reactor Location Day /Mo/Yr EFPY Cause/ Comment II.C.2.c.(1)
Brunswick-2 B 4 or 5/75 0.04 Dislodged 0-ring, burr on plunger, heat II.C.2.f. liatch-1 B 16/05/78 2.7 Sticking solenoid II.C.2.f. llatch-1 C 09/06/78 0.5 Solenoid plunger dirty, out of adjustment, and/or partially damaged during handling II.C.2.f. Hatch-1 L 09/06/78 0.75 Solenoids sticking II.C.2.f. Pilgrim 116 07/80 2.3 Locktite on plunger t II.D.2.a.(1) !!onticello A 03/02/78 0.7 Pilot inlet filter plug seat cut by steam II.D.3. Brunswick-2 B 15/07/77 0.35 Electrical II.D.3.a. Brunswick-2 E 09/80 0.5 Broken solenoid coil wire II.D.3.b. Peach Bottom-3 J 11/74 0.1 D.C. system grounds II.E. Hatch-1 E 11/76 1.8 Failed bellows pressure switch II.E. Hillstone-1 A 02/79 0.7 Cracked sensing tube II.E. Fitzpatrick J 02/78 1.55 Cround. due to moisture on switch I
e TABLE B-6. TARGET ROCK TWO-STAGE MODEL 7567 FAILURE EVEllTS jassification Cooponent Event Date [ Category Reactor Location Day /tto /Yr Cause/ Comment B.I.b. Pilgrim D 10/80 Foreign material probably lodged between guide and piston rod 1.C.1. llatch-1 Seven 04/81 Setpoint drift of valve actuator Locations I.C.2.f. Fitzpatrick G 01/81 Locktite compound in solenoid valve I.C.2.f. Millstone-1 One 04/81 Particulate contamination in solenoid Location
- e d
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s s s s ATTACHMENT ~2 ~ NUREC-0737 s " CLARIFICATION OF TMI ACTION PLAN REQUIREMENTS" SECT 7 .D.1 s %-^- \\ ~ %N m s \\ )
8 1I.0.1 PERFORMANCE TESTING OF BOILING-WATER REACTOR AND PRESSURIZED-WATER REACTOR' RELIEF AND' SAFETY VALVES (NUREG-05)8, SECTION ?.l.2) Pcsi+. ion Prussuriyed-water reactor and boiling water reactor licensees and applicants ., hall conduct t.est.ing to quali ty the reactor coulant system relief and salet.y valves under expected operating conditions for design-basis transients and accidents. Changes to Previous Reouirements 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.0.1 as described in NUREG-0660. 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 terminatin0 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 4 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 11.0.1 of NUREG-0650. The clarification below provides updated information~ on PWR ATWS temoerature and pressure conditions and clarifies that ATWS testing need not be accomplishad by July 1981. Clarification 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 the dynamic forces on the safety and relief valves are maximized. Test pressures shall be the highest predicted by conventional safety analysis 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. A. Performance Testing of Relief and Safety Valves--The following information must be provided in report form by October 1, 1981: (1) Evidence supported by test of safety and relief valve functionability 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.
~ (2) Since it is not planned to test all valves on all plants, each licensee must submit to NRC a correlation or other evidence to substantiate that the valves tested in the EPRI (Electric Power Research Institute) or other generic test program demonstrate the functionability of as-installed primary relief and safety valves. This correlation must show that the test conditions used are equivalent to expected operating and accident conditions as prescribed in the final safety analysis report (FSAR). The effect of as-built. relief and safety valve discharge piping on valve operability must also be accounted for, if it is different from the generic test loop piping. (3) Test data including criteria for success end failure of valves tested must be provided for NRC staff review and evaluation. These test data should include data that'would permit plant-specific evaluation of discharge piping and supports that are not directly tested. B. Qualification of PWR Block Valves--Although not specifically listed as a short-term lessons-learned requirement in NUREG-0578, qualification of PWR block valves is required by the NRC Task Action Plan NUREG-0660 under task item II.D.1. It is the understanding of the NRC that testing of several commonly used block valve designs is already included in the generic EPRI PWR safety and relief valve testing program to be completed by July 1, 1981. By means of this letter, NRC is establishing July 1, 1982 as the date for verification of block valve functionability. By July 1, 1982, each PWR licensee, for plants so equipped, should provide evidence supported by test that the block or isolation valves between the pressurizer and each power-operated relief valve can be operated, closed, and opened for all fluid conditions expected under operating and accident conditions. 'l C. ATWS Testing--Although ATWS testing need not be completed by July 1, 1981, the test facility should be designed to accomniodate ATWS conditions of approximately 3200 to 3500 (Service Level C pressure limit) psi and ~ 700 F with sufficient capacity to enable testing of relief and safety valves of the size and type used on operating pressurized-water reactors. l Applicability This requirement applies to all operating reactors and operating license applicants. 1 Implementation l, See implementation schedules in the " Documentation Required" section. Type of Review Preimplementation review will be ptrformed for EPRI and BWR test programs with re:,lmt:1. Lo spin I li IthtLlon o f reiIeI niul f..i t e Ly va ivm.. A l '.o, I.lui npp i li:an t.s ' proposal for functional testing or qualil'ication of PWR valves will be reviewqd. Postimplementation review will also be performed of the test data and test results as applied to plant-specific situations.
o Doc,tJmonta_ tion Reguirqcj l'ri Imp lomon t..it.lun rov line w i l l leo le.it. oil on l l'l!!, ilWit,.un t opp i le:.in t t.nf mil l.l.o t s, with regard to Llie various tes t prograitis. liiese sobriiittals 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 1, 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 those plant-specific applications, the test data and results of the various programt will also be required by the following dates: PWR (EPRI)/BWR Generic Test Program Results--July 1, 1981 Plant-specific submittals confirming adequacy of safety and relief valves based on licensee / applicant preliminary review of ge'neric test program results--July 1, 1981 Plant-specific reports for safety and ' relief valve qualification-- October 1, 1981 Plant-specific submittals for_ piping and support evaluations--January 1, 1982 Plant-specific submittals for -block valve qualification--July 1,1982 T_echnical Specification Changes Required No technical specification changes are required. References NUREG-0578 NUREG-0660, Item II.D.1 O e 4 1
t ATTACHMENT 3 " ANALYSIS OF GENERIC BWR SAFETY-RELIEF VALVE OPERABILITY TEST RESULTS" pp. 57-59, 78-83 l l l I l \\
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~ TABLE 4.2-1 Page 1 of 3
SUMMARY
OF REDUCED DATA TARGET ROCK 6X10-2 STAGE S/RV WITH LOADS I SUPPORTS Test Data
- Steam, Water, 15 F Water, 50 F Test Parameter Saturated Subcooling Subcooling Description Units Run 301 Run 303 Run 307 i
6 e a G e 0 0 e I
IAllLE 4.2-1 Page 2 of 3 SilMMAltY UI' lti()UCl.1) 1)AI A I Al(GLi l(UCK 6M10-2 SI AGL S/l(V W1Ill LUAUS 1 Sul'I'UltiS Test Data, Maximum Dynamic Values . Steam, Water, 15 F Water, 50 F Test Pararreter Saturated Subcooling Subcooling Description Units Run 301 Run 303 Run 307 58
TABLE 4.2-1 Page 3 of 3
SUMMARY
OF REDUCED DATA TARGET ROCK 6X10-2 STAGE S/RV WITH LOADS I SUPPORTS Test Data, Maximum Dynamic Values
- Steam, Water, 15"F Water, 50 F Test Parameter Saturated Subcooling Subcooling Description Units Run 301 Run 303 Run 307 O
e 2 i a 5 1 l ,0 t E 4 ,e, ,m,-
rcpresents the stress component due to pipe temperatura effects. Deviations of the actual trace above and below the mean value line represent the stress component due to dynamic loading. 6.2 Test Results 6.2.1 Description of Discharge Phenomena Following S/RV actuation for steam discharge, the pressu're within the S/RVDL increased.' Pressurization continues until the water and air initially in the S/RVDL have cleared. The sequence of the events for Run 17 (steam discharge, Crosby 8X10), also shown in Figures 6.1-9 and 10, is as follows: Time (msec) 78 i
( = 6.2.2 Pressure Sensors Data Sunmary Six pressure transducers (sensors P1, P2, P3, P4, P5 and P10) were installed on the S/RVDL to measure pipe pressure during line clearing and subsequent flow. There were also two pressure transducers (P0 and I P6) installed on the sweepolet and steam chest respectively. Locations of the sensors are shown in Figures 6.1-1 and 6.1-2. The average back pressure for each steam run reported is tabulated in Table 4.2-1. The steady state backpressures for the water runs were inconsequential 1 l 1 6.2.3 Strain Gage Data Summary Thirty five strain gages were installed on the S/RVDL, steam chest and sweepolet outlet. The locations of the gages are shown in Figures 6.1-1 and 6.1-2. The strain measurements were converted to stresses by multiplying by the modulus of elasticity. The stresses obtained from each sensor are tabulated in Table 4.2-1.
o For some strain gage locations, stresses measured in water tests were liiglier Llian Lliube ineobured Isi Llie currespunding sLuani LesL. II Is is cluu to the extremely fast pressurization rate used in these tests (0-250 psig in less than one second) which was necessitated due to facility constraints. Actual in plant pressurization rates for initiation of alternate shutdown cooling will be much slower. In all cases, however, the measured stress levels are low. 6.2.4 S/RVOL Pipe Support Load Data Summary The support loads obtained from each load cell are tabulated in Table 4.2-1. Examples of load time history plots for steam and water discharge are shown in Figures 6.1-10 and 6.2-1 respectively. The maximum loads acting on each support structure from all Load I tests are tabulated below. Crosby 8R10 Steam Discharge Water Discharge Load Load Water X 100% Run (Kips) Run (Kipsl Steam Dikkers 8R10 Steam Discharge Water Discharge Load Load Water X 100y Run (Kips) Run (Kips) Steam 80
O Crosby 6R10 Steam Discharge Water Discharge Load Load Water X 100% Run (Kips) Run (Kips) Steam Target Rock 6X10 3-Stage Steam Discharge Water Discharge Load Load Water X 100% Run (Kips) Run (Kips) Steam I 4 Target Rock 6X10 2-Stage Steam Discharge Water Discharge Load Water Load 100% Run (Kips) Run (Kips) Steam Dresser Electromatic 6X8 Steam Discharge Water Discharge Load Load Water X 100% Run (Kips) Run (Kips) Steam t I The maximum loads acting on each support structure from Load II tests are tabulated below. 81
Steam Discharge Water Discharge o Load Load Water Run Valve Type (kips) Run Valve Type (Kips) Steam X 100% 6.2.5 Pipe Thrust Load Data Summary The pipe thrust loads were calculated by equation (1) by combining the measured support loads and pipe acceleration. The calculati.ons are for Load II test only. The maximum calculated loads are tabulated below. l Steam Discharge Water Discharge Load Load Water Run Valve Type (kips) Run Valve Type (kips) Steam X 100% i 6.3 Pipe Load Evaluation Conclusion As described in the foregoing sections, the water discharge loads were far less than the steam discharge loads in all cases. This ratio is applicable to all other S/RV piping arrangements. The response of the S/RVDL for steam S/RV inlet flows were analytically predicted. In general, the analytically predicted piping and support response was comparable to the measured responses for steam. 82
O Therefore, the test and analysis demonstrated that the current S/RV discharge pipe design is adequate for the alternate shutdown cooling conditions. 1 I 8 i}}