ML20127P010
| ML20127P010 | |
| Person / Time | |
|---|---|
| Issue date: | 04/15/1985 |
| From: | Advisory Committee on Reactor Safeguards |
| To: | Advisory Committee on Reactor Safeguards |
| References | |
| ACRS-2294, NUDOCS 8505230672 | |
| Download: ML20127P010 (16) | |
Text
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S big f DATE ISSUED: April 15,1985
SUMMARY
/ MINUTES OF THE SUBCOMMITTEE MEETING ON RELIABILITY ASSURANCE MARCH 19, 1985 WASHINGTON, D. C.
A meeting was held by the ACRS Subcommittee on Reliability Assurance on l March 19, 1985. The purpose of the meeting was to discuss valve re-liability. Notice of the meeting was published in the Federal Register on March 4, 1985 (Attachment A). The schedule of items covered in the meeting is in Attachment B. The list of attendees is in Attachment C.
A list of handouts is in Attachment D. The handouts are filed with the office copy. R. Major and H. Alderman were the cognizant staff members for this meeting.
Mr. C. Michelson convened the meeting at 8:30 a.m.
Principal Attendees:
ACRS Members ACRS Staff C. Michelson, Chairman R. Major, Cognizant Mbr.
C. Wylie H. Alderman, Cognizant Mbr.
D. Ward A. Cappucci, Staff Engr.
G. Reed Consultant Others H. Jones, TVA D. M. Eissenberg, Oak Ridge A. Charbonneau, Movats, Inc.
NRC Staff K. V. Seyfrit E. J. Brown R. J. Bosnak R. J. Kiessel .
G. H. Weidenhamer F. C. Cherny J. D. Page S. D. Stadler J. Vora g 52 g 2 850415 2294 pyg DESIGNATED ORIGINAL s s L:
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Sammary/Minut:s 2 Reliability Assurance Mtg.
March 19, 1985 L Vora, Division of Engineering Technology, Office of Nuclear Regulato-ry Research Mr. Vora noted that he was involved with nuclear plant aging research.
.This includes aging and service wear assessment, detection of defects, and the characterization and degradation monitoring of electrical and mechanical components including valves.
He presented an overview of the various valve-related research activ-ities in the Office of Research. He remarked that the Division of Risk Assessment has a program called IPRDS (In-Plant Reliability Data System) which is conducted by the Oak Ridge National Lab. ORNL collects and analyzes data related to valves. This data is being used to study motor operated valves.
The Division of Engineering Technology has two branches. The Mechanical and Structural Engineering Branch, and the Electrical Engineering and Instrumentation and Control Branch. Both these branches are involved in valve related activities.
The Mechanical and Structural Engineering Branch has two valve related activities, one is related to the Qualification program and the other is a special valve test that is primarily for leak tests.
The Electrical Engineering Branch has as its major thrust nuclear plant aging. The emphasis is trying to identify the aging and service wear associated with electrical and mechanical components. The thrust is on surveillance and monitoring methods which might be effective in detect-ing defects in the incipient stage prior to failure.
G. Weidenhammer, Mechanical and Structural Engineering Branch, Division of Engineering and Technology .
Mr. Weidenhammer talked about the work on containment isolation system valves. The prime contractor for this work is the Idaho National Engineering Labs. The main concern in this area is the purge and the
W- -
Summary / Minutes 3
- R211 ability Assurance Mtg.
-March 19, 1985 vent valves. He noted that the containment integrity people feel that the most likely scenario for a containment to fail functionally under severe accidents is due to excessive leakages through penetrations rather than rupture of a shell structure.
This work is carried on under the auspices of the equipment qualifica--
tion effort. The testing is up to design basis levels for equipment qualification and then up to severe accident areas. This is a cost effective way to obtain the information. Under the general heading of penetration leakage, purge and vent valves are of considerable interest.
In addition to leakage considerations, the operability of the valves are also of great interest.
D. Eissenberg, Engineering Technology Division, Oak Ridge National Laboratory Mr. Eissenberg defined aging as time dependent cumulative degradation changes that occur within a component of a system that reduces reliabil-ity or operational readiness over time. He noted there were three types ,
of aging. Normal aging degradation which is that which would be expect-l ed under normal conditions. Premature aging results at a faster rate than what you would expect from a normal piece of equipment due to unanticipated conditions or improper practices. Accelerated aging is imposed artifically in order to define a qualified life.
HediscussedtheNuclearPlantAgingResearch(NPAR) goals. The goal is characterizing aging phenomena associated with electrical and mechanical components. This characterization would seek to understand failure modes and causes, how they occur as a result of selection of materials and stresses which they would see during nuclear plant service, and identify ways in which the degradation occurs which one could monitor.
The results of this kind of understanding would lead to recommended methods for inspection, surveillance and conditions monitoring for detecting and trending significant aging effects prior to loss of safety
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- Summary / Minutes 4 Reliability Assurance Mtg.
March 19, 1985 function. A second goal is recommending maintenance practices which would mitigate the effects of aging.
A Charbonneau, Motor Operated Valve Analysis and Test System, (M0 VATS)
Mr. Charbonneau described MOVATS as a portable test system by which you can get a mechanical and electrical "EKG" of the valve, the operator, the motor and the control service.
The MOVATS concept is based upon the higher the load demand on a valve, the more the worm will move in one direction or the other depending on whether the valve is opening or closing. The movement of the spring pack is calibrated to the actual stem-delivered load. An additional
- parameter that is measured is the motor current. A third parameter is a circuit built into the portable device that monitors the actuation of all six centrol switches.
The portable device produces a signature trace with a time tracing that permits a complete analysis of normal and degraded modes of operation.
The trace is visually displayed and stored on tape. It can then be fed into a computer for a permanent record or a hard copy produced if it is so desired.
E. Brown, NRC, AEOD Mr. Brown presented evaluations that have been completed by AEOD. The first recommendation, he noted, was that there should be a reassessment of regulatory guide 1.106. This deals with bypassing of thermal over-loads. The primary basis for this was a relatively large incidence of motor burnouts that appears to be related to the guide usage.
Another recommendation was that there needs to be improved methods and .
procedures for the setting of torque switches to be able to evaluate the valve operability and functional qualifications under accident con-ditions. The reason for this is that the valves are testing under
Summary / Minutes 5 Reliability Assurance Mtg.
March 19, 1985 specified test conditions but this doesn't demonstrate operations under accident conditions.
Another recommendation is that signature tracing conditions should be developed and tried on selected valves as part of a periodic in-service testing program. This should be helpful in detecting aging or the inadequate adjustment in maintenance.
G. Rivenback, P.M. for Hatch, NRR Mr. Rivenback discussed the Hatch #2 event of August 25, 1982. This event started with the failure of an MSIV valve. The disc dropped loose from the stem and the valve fell closed. When the valve closed, the reactor tripped because of a high neutron fit,x that occurred as a result of the collapse of the bubbles. The scram couldn't be cleared because the pressure and temperature was up in the drywell. Tne drywell cooler was out, so the pressures and temperature couldn't be reduced. Even-tually the drywell cooler high pressure trip was bypassed and the pressure and temperature was reduced. The bottom line of this event was that at the time the reactor scrammed, the discharge volume drain valve failed to close. The discharge volume header bled into the clean radwaste drain system. The radwaste drain system is a closed system but a pipe cap was left off. Steam came up through the RCIC floor, in-creased the temperature and tripped the RCIC system.
The scram discharge valve failure was caused by a bonnet that was not tightly bolted to the valve.
Subsequent to the incident, another valve was installed in the drain line and another valve in the other line, so they are protected against single failures. The drain cap has been tack welded to prevent it from being taken off easily. A switch has been installed in the control room .
allowing bypass of the high pressure trip without actually going out to the building and disconnecting wires.
. Summary / Minutes 6 R211 ability Assurance Mtg.
March 19, 1985 D. Diiani, NRR Mr. Diiani discussed generic issue 77. Generic issue 77 looks at plants licensed prior to September 1975 where the standard review plant did not cover the requirements of flooding backflow through the floor drains.
Nine representative plants have been selected for review. These plants will be reviewed to determine how the flooding issue was handled. Upon completion of this review, the various AE's will be contacted to see what was done in this area. Upon completion of this task, a determina-tion will have to be made if backfitting is necessary or not.
One additional concern is to make sure proper check valves have been installed in the lines to make sure there is no backflow. These check valves have to be readily available for maintenance and testing.
W. Long, P.M. , NRR, Browns Ferry Incident of March 20, 1984 On March 20th, Browns Ferry was getting high radiation in the reactor building due to a leak in a RCIC line due tc erosion. They decided to isolate the leak to perform maintenance. The first valve in the line was closed but it leaked and didn't isolate the leak. It was decided to isolate the next valve in series. The RCIC was taken out of service and the leak repaired.
Attempting to open the upstream valve slowly to repressurize the line and warm it up slowly, it was found the valve wouldn't open. It had full reactor pressure on the upstream side and was depressurized on the downstream side. The motor operator was not powerful enough to open the valve. ,
Further discussion disclosed that the motor had tripped out on thermal ,
overload. The major concern is whether the valve could close in the event of a steam line break. This issue has been assigned generic number 87 and is awaiting prioritization.
_ l
" Sumary/ Minutes 7 R211 ability Assurance Mtg.
March 19, 1985 R. Bosnick, Division of Engineering Mr. Bosnick declared that he didn't have a presentation but was prepared to answer questions.
Mr. Michelson asked about isolation valves whose function was to isolate a break downstream of the valve. His question was how are these valves-
-specified at time of purchase. Do the purchase specifications include accident conditions or just normal operating conditions? How do you know the valve works? Are they tested under accident conditions or are calculations relied upon to assure operation? How do you assure yourself that the valve parameters haven't changed during installations or maintenance?
Mr. Bosnick replied that during the 0. L. Review process the purchase specifications are reviewed. The specifications that he has seen list normal operating conditions. He noted that valve vendors tell him that they never receive the proper conditions under which the valve must operate. The assurance that the valve must close under accident con-ditions is strictly by analysis. Any testing that is done is under normal operating conditions. If the valve has degraded, it could be picked up under testing conditions but that couldn't determine if it would operate under accident conditions.
R. Baer, Chief of the Engineering and Generic Communication Branch, Office of Inspection and Enforcement Mr. Baer noted that he would summarize I&E activities associated with valve problems, particularly the issuance of information notices and bulletins. Since 1971, I&E has issued about 75 bulletins or information notices associated with valve problems. Since 1983, 21 information r
notices and one bulletin on valve problems have been issued.
The first category was maintenance. He discussed problems with Limitorque gearheads. These were drives on butterfly valves and the l
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r Sumary/Minut';s 8 Reliability Assurance Mtg.
March 19, 1985 gearheads were not installed properly and essentially not aligned properly. When the gear is misaligned you could run the wonn to where it disengaged from the gear and then it would not reengage..
i Another information notice dealt with main steam isolation nonreturn check valves. These are the valves used on a PWR to prevent blowdown of more than one steam generator in the event of a steam line break outside
. of containment. An event at Trojan resulted in the nonreturn check valves remaining open when they should have closed. Upon investigation the licensee found the basic problem was the packing was too tight and it hadn't been maintained frequently enough.
As part of the post-TMI fix at Rancho Seco, an extra handle was placed on the PORV. The handle was supposed to act as an indicator as to whether the PORV was open or not. The concept was to observe the position of the handle through closed circuit TV and tell if the valve was open or not. The problem was that the valve manufacturer was not consulted. The extra weight of the handle created an unbalance and interfered with the valve operation.
The problem was solved by removing the handle and using a white back-ground for observation.
Mr. Baer noted the continuing problems with Target Rock relief valves.
The Target Rock three stage relief valves have a tendency to blow down and stay open. The two stage target relief valve was introduced to alleviate this problem. The two stage Target Rock relief valves have exhibited a tendency not to open at the desired set point. This problem is still under study.
Another valve problem was at Rancho Seco and concerned the I; urge valves.
They are normally closed during operation and after a refueling outage, ,
they would not close properly. This was discovered when the valves failed the containment surveillance test. These valves were not lu-bricated and they were not tested prior to the outage.
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Sumary/ Minutes 9 Reliability Assurance Mtg.
harch 19, 1985 Mr. Baer noted problems with check valves in the auxiliary feed water lines. Leakage past the check valves results in steam binding of the auxiliary feed water pumps.
Torque switch settings have been a problem. The torque switch settings at Oyster Creek were changed and were lower than the original vendor specifications. It isn't clear whether the valves would have closed during postulated accident loads. The licensee didn't have any formal set point requirements or documentation or procedures to determine set points. Apparently, the set points had been lowered because it was found that they were adequate for surveillance testing.
Browns Ferry had a problem with a torque switch being set too high.
This occurred on the RHR system. One of the valves that isolate the RHR from the Reactor Coolant System couldn't be opened. With too high a torque switch setting in closing the valve, the seat was forced into the disc. Consequently, when it was attempted to open the valve, the breakaway force was too high.
Both Davis Besse and St. Lucie have had prot:1 ems with main steam safety valves due to fracture of cotter pins. A nut is loosened and the MSIV lifts. This problem is very significant on Babcock and Wilcox reactors.
Once a steam generator is blown down, you have to be very careful in cooling down the reactor system.
The next group is based upon Information Notice 84-74 and involved the potential for an intersystem LOCA. All the instances involve isolation of reactor coolant system systems from low pressure systems. There are three cases, all involving valves and the potential for overpressurizing low pressure systems.
Pilgrim had a check valve that was held imobile by rust and resulted in .
l high pressure reaching the low pressure end of the HPCI system.
At Hatch a check valve was open for four months because it was assembled wrong. The air pressure was constantly on, keeping the valve open. The
< Summary / Minutes 10 Reliability Assurance Mtg. ,
March 19, 1985 result was that for four months, there was only one valve isolating the reactor coolant system from the RHR.
At Browns Ferry the core spray line was overpressurized. A solenoid pilot operator for a check valve had reversed ports.
The next category of valve problems dealt with a hammering effect. A core spray valve failed to operate. The specific utility was not identified. Upon investigation, it was found that the valve gear housing had failed, and the valve had been repeatly hamered closed by the valve operator. One possible cause is that a limit switch is out of calibration and the valve is pounded until something fails.
The next category of information notices on valves were those involving misapplication of valves. The first case was at Crystal River. It was a valve in the high pressure safety injection system. This valve was being throttled much too much and was causing cavitation, degradation and high velocity erosion.
Another case involved Rockwell International globe valves. Again the flow was being throttled outside the range where it should be, and the disc of the valves fell off.
The formal portion of the meeting was adjourned at 5:29 p.m.
The next scheduled meeting of the subcomittee on Reliability Assurance will be on April 23, 1985.
NOTE: ,A complete transcript of the meeting is on file at the NRC Public Document Room at 1717 H St., NW., Washington, D. C. or -
can be obtained at cost from ACE Federal Reporter, Inc., 444 N. Capitol St., Washington, D. C. 20001, Telephone (202) 347-3700.
Ft .* ' la g. a ! Vt. b. G / Monday March 4.1955 / Notices 8Gr
& r , krit er. stater. . v. l b+ issuej J ne n.19c0, cuthonzed tha r:quist:d cxt:nsion g,f the ditt for ettmed c d rud' au M. s te as beense to operate the facihty o' submittal c f the upd:ted Fin:] Saf;ty C:mn.;ttu . Kaori p vi.. Le perrr.:t'ed steady-state rese:or power lesels not in Analysis Report. ne requested en!) during thew partions t f the excess of191 megawatts thermal. These extension to December 1.1985 (s monthe 4, rnertmg whtr. a transcr:Lt is being Lapt. licenses proside. 6mong other things. efter startup testing is scheduled to be cnd questaurs tra) be asi ed c,r.(.1,3 that they are subject to all rules, completed) will allow the bcensees' members of the Sabcommitiee. Its regulatmas ar.d Orders of the engineering personnel necessary and scnsultants and Staff. Peru,ns doirm; Commissior.. sufficient time to complete startup to make oral statements should notd) gg testing and resultant design changes the ACRS staff member named below as before conducting the engineering fit in advance as is practicable so that Section 50.71(e)(3)(i) of to CFR Part 50 review associated with the preparation cppropriate arrangements can be mode. requires the licensees of riuclear power of the UFSAR.
During the mitial portion of the reactors to submit an Updated Final . The NRC staff considered safety meeting. the Subcommittee along with Safety Analysis Report (UFSAR) within 24 months of either July 22.1980, or the aspects of the requested extension to the cny ofits consultants who n.ay be pres:nt, may exchange prehmma 3 date ofissuance of the operating license. UFSAR submittal date.During the long virws regardmg matters to be whichever is later. The above regulation al oflow power and power ascension testing to date. Mississippi nsi ered durmg the balance of the 8-g have qu red u mitta of de
)ge 16. Power & Light Company (MP&L) has submitted four FSAR amendments and He Subcommittee will then hear presentatiens b) and hold dis:ussions 1984' By letter dated February 6.1984 other licensing documents providing with representatives of the NRC Staf' licensees requested an exemption to 10 information regarding changes in plant its c:nsu!! ants. and to other intereste$1 CFR 50.71(c) which would have, deferred design plant procedures, and safety p;rsons regardmg this review submittal of the UFSAR until12 months analyses. Mpal will continue to provide after Unit 2 was licensed on the basis information and analyses needed for Further information regardmg topics 12 be discussed whether the meeting that the FSAR is written for both Unit 1 accurate and timely evaluation of and Unit 2 of CGNS. By letter dated matters of safety significance. pending the submittal of the UFSAR. MP&L bas also b:s.irman's Ch been cancelled ruling on requests for ort rescheduled'he Jue 26.1984. the NRC staff denied that request because Unit 2 is scheduled to implemented a system to make cpportunity to present oral statements be completed after 1990, and the staff controlled copies of principal design cnd the time allotted therefor can be (bt ined by a prepaid telephone call to requested the licer. sees to provide a drawings available to reactor control the cognizant ACRS staff member.Mr'
- d'fied exempti n request.Byletter room operators and emergency response Richard Major [ telephone 20;;/63M414) dated December 31.1984. the hcensees facility staff.Thus, the granting of the requested an exemption to defer requested extension will have no between Pers:ns planning 8:15 a m.tl5is to attend and metir 5.00 p submittal m" e s oft'.g the UFSAR for Unit 1 of significant impact on plant safety.
tre urged to contact the above named GGNS until December 1.1985. on the %e public interest will be served by individual one or two days before th- basis that bcensees engineering granting the exemption since licensees scheduled meeting to be adsis d of any personnel. who will be involved in the can continue to use engineering ergmeermg rmw of the UFSAR. are ch:nges in schedule. etc.. w hich mas personnel to support startup testing.
hsve occurred neededt support startup testing of thus. assuring completien of startup Grand Gulf Unit I now in progress. testing and start of commgrcial D:ted: Februa 3 27.1985 The NRC staff has reviewed the operation sooner then would be the case Morton W. libarkin. licensees' request for an extensien of the if engineering personnel were diverted Assistant. Ea ccum e D. rector for rruject Grand Gulf Unit 1 UFSAR subzrittal to update the FSAR.
Remm date to December 1.1985.The extension Based on its review, the staff lFR Doc 8%5t9a Fded b1-e.5 rm a .; is,needed for Grand Gulf Unit 1 becaust concludes that iss :ance of this mumo coot nw., e '-' lans micrval between issuance of exemption will have ne sienificant effect
.- w paer itcense and ertherization
. pc ..c rperation. Fct m:st p ants. on plant safety. Furthe:. thf s action is in
,, the pubhe intuest ar.c reed cause has tm.e a evailable aw completion been sl owr. te serport the exemption.
Ci:sts s.;r,' s ,, , . ; U;;v Co. et et : 8 Therefore. e : H r-ar.th csernption from T;mpcc :- es.p+ From
.'.)e 24 [ont. in e . al e i the date of cc -- ..: :+ n acceptable.
,,n .St!. However, for G:and Compi.. Pursuant t C ' 51.32. the n er r.scension test:ng started t- the - dW- .hcr .19&4. about M. months after Corntnission ;r.. c - rmined that the C. >fid. Era , 'v.c.
e-J o a ..1[:- u cf me low powerlicense. issuimce of,tb t -*ption will have no h;
- re: m: P.. ,o -
to t.r-is expected h be ::gn;f cant tr - . . ti,e environment it -d C . car 5- % sed nr s'.pnl1985. It is desirsble
[ IT.583 $
g ; e !:.e power and powcr g:,
..'. ? . ;ns before upd.itmp the
' cssisc ?cwcr t . C. .3 so : . . de's:gn modnin ations :ordinF';. tr.t Commission has M .ile P ".nern. i .. n . n ra .
.y by testmg can be ( . .:nined t': .. ; arsuant to lo CFR M . ism 4i V ::n m . n d so that lict nst 's' . .. an c- -; !: authorized by law
( .' Lico. tr - :*
..t; scnnel who are }..avily J will nc L . . pr life or property or .
F : tty C . :q i c. f u. - ~ .S .dmt. suppnrt of sicrtu;. testin: t? ' common c.Jcnn and security andis ineed R < ! 1.1I
. '* '4....t ;. 2nt modificationt can bc othe wise ir. the pblicinterest. ,
ccthorire> - , owe gtsor *.
p nr and reviewing the Therefore. ti Ccr.missior. hereby -
Grand GE .Jearinu.p.. Uma t & 4 *:-/.F. '. cus, for Crand Gulf approves th i#.owing temporary isciht)) A scerseced beense. N1 , . d cam has been shown for exemption frof comphance with the f 'i.4 :-h } &-
ATTACHMENT 5 REVISION 1
' RELIABILITY ASSURANCE (VALVES) SUBCOMMITTEE MARCH 19,1985 l 1717 H Street N.W., WASHINGTON, DC 20555 ROOM 1046 10 min 8:30 a.m. 1. Chairman's Introduction (C. Michelson)
- a. Purpose
- b. Goals 4
1 hr 40 min 8:40 a .m. 2. NRC-RES Discussion of Programs Relevant to Valve Operability and Reliability Assurance
- a. An overview J 17~ # K& -
b.QualificationandOperebility(G.Weidenhamer)
NRC-RES
- c. Aging & Service Wear; Detection (D. Eissenberg)
. of Defects and Characterization 10 min 10:20 a.m.
- . BREAK 2 hrs 10:30 a .m.'
- 3. Presentation by Movats, Inc. (A. Charbonneau)
- a. Description of motor operated valve analysis and test system
- b. Past experience with the prediction of future valve performance c.' Statistical Analysis of the System
- d. Questions'and Answers
- 1 hr 12:30 p.m. *** LUNCH (E. Brown) 1 hr
.1 : 30 p . m. 4. AE00-Current Work on Valve Operability and Reliability "
- a. Update of 8-25-82 HATCH Event - further efforts '
- b. Follow-up on the survey of valve .
> operator-related events n i f e-
- c. Other act'ivities con'cerning valve reliabili y A 7'l AC tl M CfII
1 I
t ,
l 2
i REVISION 1 (NRR) 30 min 2:30 p.m. 6. Discussions with NRR I
- a. Discussion of 8-25-82 Hatch Event
, description
. follow-up actions
. current status
- b. Browns Ferry: Significant Failure 30 min 3:00 p.m.
of an RCIC Steam Line Isolation Valve to Open Against Operating Reactor Pressure, March 21, 1984
. description
. follow-up actions
. current status 10 min 3:30 p.m. *** BREAK 30 min 3:40 p.m. c. Discussion of Isolation of Breaks Outside Primary Containment 1)Isvalvequalificationsufficient to close against LOCA forces ano environment 2)Specificexamplestobeaddressed:
HPCI, RCIC, RWCU 3)Valvespecifications
- 4) Testing: valve performance validated
- 7. IE Bulletins & Notices over the past few 35 min 4: 10 p.m.
years dealing with valve issues (R.Baer,IE)
- 8. Discussion of future Meeting: April 23, 1985, 15 min
- 4:45 p.rn.
Washington, DC a.' Topics
- b. Timetable
- iv 5100 p.m. ADJOURN ,
R. MAJOR ATTACHMENT C RELIABILITY ASSUR_ANCE _
ACRS SUBCOMMITTEE MEETING ON 4 ,
.LOCATIO:1:
ROOM 1046,1717 H St., NW., Washington, D.C.
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,.x5 SUBCOMMITTEE MEETING ON ROOM 1046,1717 H St., NW., Washington, D.C.
f0cATIOll:
Manh 19,1985 DATE:
ATTENDANCE LIST PLEASE PRINT:
BADGE NO. AFFILIATION NAME
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ATTACHNENT D HANDOUTS RELIABILITYASSURANCESUBCOMMITTEEMEETING(VALVES)
March 19. 1985 Washington, D. C.
- 1. Valve Problems with Generic Implications - R. Baer, NRC, ISE
- 2. Hatch 2 event, August 25, 1985 - G. Rivenbark, NRC
- 3. AEOD/C403 " Case study report for the Edwin I. Hatch Unit No. 2 Plant Systems Interaction Event on August 25, 1982, - E. Brown, AE0D
- 4. Overview of NPAR Activities Relevant to Valve Operability and Reliability - D. Eissenberg, ORNL
- 5. AE00 Evaluations of Valve Operability - E. Brown, AEOD
- 6. Chart Defining Aging - D. Eissenberg, ORNL
- 7. Aging Characterization and Detection of Defects in Valves - J.
Vora, NRC, RES
- 8. RES Programs Related to Valve Operability and Reliability - J.
Vora, NRC, RES
- 9. Figure 1.1, Explored diagram of valve operator internal assembly showing position of limit switches - source unknown
- 10. RES programs related to valve operability and reliability - G.
Weidenhamer, NRC
- 11. Brochure on MOVATS (motor operated valve analysis and test system -
A. Charbonneau, President M0 VATS O
e a g 1