ML20006D144
| ML20006D144 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1989 |
| From: | Kirkwood R NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| To: | |
| References | |
| REF-GTECI-070, REF-GTECI-NI, TASK-070, TASK-70, TASK-OR NUREG-1316, NUDOCS 9002120148 | |
| Download: ML20006D144 (26) | |
Text
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i NUREG-1316 Technical Findings and Regulatory Analysis Related to Generic Issue 70 E
i i
Evaluation of Power-Operated Relief Valve and Block Valve Reliability in PWR Nuclear Power Plants U.S. Nuclear Regulatory Cornmission i
L Office of Nuclear Regulatory Research R. Kirkwood l -
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LI, AVAILADILITY NOTICE Availability of Reference Materials Cited in NRC Publications Most doCumonts Cited in NRC publications will be available from one of the following sourCOs:
1.
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i
NUREG-1316 Technical Findings and Regulatory Analysis Related to Generic Issue 70 Evaluation of Power-Operated Relief Valve and Block Valve Reliability in PWR Nucicar Power Plants Manuscript Completed: September 1989 Date Published: December 1989
' R. Kirkwood -
Division of Safety Issue Resolution Omce of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission
. Washington,' DC 20555 g.a..y,
ABSTRACT
'Ihis repon summarizcs work performed by the Nuclear power plants. The report identifies those safety related Regulatory Commission staff to resolve Generic issue 70, functions that may be performed by PORVs and describes 3~,
" Power Operated Relief Valve and liksk Valve Reliabil-ways in which PORVs and block valves may be improved,
- ity." "Ihc report evaluates the reliability of PORVs and
'this report also presents the regulatory analysis for bksk valves and their safety significance in PWR nuclear Generic issue 70.
1 iii NUREG-1316
!s CONTENTS 6
fuge
. AbstraCl..........,...........................*****.**********'*****************
555
{
Acronyms a nd I nit ialism s....................................................................
Vii Ac knowl ed g m e n t s.............................................................................
ix Execu tive S u m ma ry..........................................................................
xi
!l
} I n t rod u Cl io n..............................................................................
I
?
2 ' I f a C h g ro u n d..............................................................................
I 2.1 Safety Functions of PORVs and Block Valves.............................................
6 2.2 Description of PORV Safety Functions...................................................
6 2.2.1 S t eam G ene rator Tube R u pt ure..................................................
6 2.2.2 low. Temperature Overpressure Protection.........................................
6 2.2.3 Plant Cooldown in Compliance with Ilranch Technical Position RSil 5-1................
7 j
2.3 R eactor Coolant Syst em Venting........................................................
7 2.4 nre e Mile Island Unit 2 PORV........................................................
7 2.5 NRC Information Notice 89-32. Surveillance Testing of low. Temperature Overpressu re Prot ection Syst ems........................................................
7 3 Con t ract o r R e po rt s....,...................................................................
7-3.1 N U R E G /C R -4 69 2....................................................................
8 3.2 N U R EG /CR -4 999....................................................................
9 4 Construction of PORVs and Block Valves.....................................................
10 4.1 Cod es a nd S t a n dards...................................................................
10 4.2 S cism ic D esig n...,...................................................................
10 4.3 Q uali ty Assu ra nce.....................................................................
10 4.4 Co nt rol Syste ms......................................................................
10 4.5 M ot o r Ope ra t ors......................................................................
10 4.6 Safety.Related and Non. Safety.Related PORVs and Block Valves............................
11 4.7 Operating Plant Mainte nance...........................................................
11 5 R egu la tory Analysis......................................................................
11 5.1 - A' t e r n a t ives..........................................................................
11 5.2 Potential Improvements to PORVs and 111ock Valves.......................................
12 5.3 Safety B e n e fi t s.......................................................................
13 5.4 O u tage Avoida nce Cost................................................................
14 15 5.5 Cost /Denefit Comparison
. 6 Findings................................................................................
15 "cierences.................................................................................
16 Figures 1.1 Pilot. ope rat ed relie f valve..................................................................
2
- 1.2 Air. operated (spring. loaded) relief valve......................................................
3 1.3 Pressu rizer safety and relie f valves...........................................................
4
)
y NUREG-1316 u
Tables 3.1. Fail u rc seve rity...........................................................................
8 3.2 Failure and degra 3ation modes-PORV mechanical............................................
8 3.3 Failure and degra dation modes-PORV controls.............................................
8 3.4 Failure and degra dation modes-block valves.................................................
9 3.5 PORV mechanical failures and degradation...................................................
P 5.1 Potential improve:ments to PORVs and bkwk valves............................................
12 5.2 Present value of improvements to PORVs and block valves......................................
13 5.3 Core melt probability with and without feed and bleed from case studies in USl A-45 (internal events only) with recovery..........................................................
14 NUREG-1316 vi
i ACRONYMS AND INITIALISMS APS Auxiliary pressurizer spray
.ASME American Society of Mechanical Engineers ATWS Anticipated transient without scram ENL Brookhaven National Laboratory BT1' Branch Technical Position H&W Babcock and Wilcox
.CE Combustion Engineering l
CESSAR Combustion Engineering Standard Safety Analysis Report CL Capacity loss CP Construction permit ECCS Emergency core cooling system EPRI Electric Power Research Institute FSAR Final safety analysis report 01-70 Generic issue 70 LER Licensee event report L'IDP 1.ow-temperature overpressure protection OBE Operating basis carthquake ORNL Oak Ridge National lisboratory PORV Power operated relief valve PRA Probabilistic risk assessment PWR Pressurized water nactor RCPB Reactor coolant pressure boundary RCS Reactor coolant system RPC Replaecment power cost RSB Reactor Systems Branch SGTR Steam generator tube rupture SSE Safe shutdown earthquake TMI-2 Three Mile Island Unit 2 USI Unresolved safety issue WNP-1 Washington Nuclear Project Unit 1
' 10 CFR 9 50.2 Title 10 Code of Federal Regulations Part 50,9 50.2-Definitions 10 CFR 9 50.55a Title 10 Code of Federal Regulations Part 50,9 50.55a-Codes and Standards 10 CFR Part 50, Appendix B Title 10 Code of Federal Regulations Part 50, Appendix B-Ouality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants 10 CFR Part 100, Appendix A Title 10 Code of Federal Regulations Part 100, Appendix A-Seismic and Geologic Siting Criteria for Nuclear Power Plants vii NUREG-1316 l.-
a
4wF*E,,4N>' % &.p-t 4-
F ACKNOWLEDGMENTS Several NRC staff members from the Office of Nuclear W.11. liardin, Advanced Reactors and Generic Issues Regulatory Research contributed to the evaluation docu-Ilranch NRC Division of Regulatory Applications, RiiS.
mented in'this report. Comments received from several NRC staff members of the Office of Nuclear Reac:or R. R. Riggs, Advanced Reactors and Generic issues Regulation are greatly appreciated.To support the staff's liranch, NRC Division of Regulatory Applications, RiiS.
overtl evaluation, consultants from the Oak Ridge i
National 12iboratory and the llrookhaven National laitx*
- 11. W. Sheron. NRC Division of Reactor and Plant ratory performed independent studies in the review of Systems, RLiS.
c nuclear power plant operating events invol,ing failures of power operated relief valves (pORVs)and the associated C. Liang, Reactor Systems liranch, NRC Division of bhick valves and in the areas of probabilistic nsk assess-lingineering and Systems Technology, NRR.
rnent, respectively Thc nuthors or contributors to this re-E"""'#
L Gallagher, Technical liditing R. I., llaer, I!ngineering Issues llranch, NRC Division of Safety issue Resolution, RiiS.
R. Rirkwocxl of the lingineering issues lininch (tilll),
NRC Division of Safety issue Resolution, RIIS, as Task
- 17. C. Cherny, lingineering Issues liranch, NRC Division Manager coordinated overall review of the G1-70 issue of Safety issue Resolution, R118.
and preparation of this report, under the direction of F. C. Cherny, Section Leader, tilli, and R. L llacr. Chief, R. Rirkwood, lingineering issues liranch, NRC Division 13111. M. L llevan of the I!lll provided much appreciated of Safety issue Resolution, R11S.
word processing support.
ix NURl!G-1316
r EXECUTIVE
SUMMARY
Examining Generic issue 70 (GI-70),
- Power-Operated 1.
hiitigation of a steam generator tube rupture acci-Relief Valve and Block Valve Reliability," involves the
- dent, evaluation of the reliability of power-operated relief valves (PORVs) and block valves and their safety signifi-2.
I ow-temperature overpressure protcetion of the re.
cance in pressurized water reactor (PWR) nuclear power actor vessel during startup and shutdown, or plants. Traditionally, the PORV and its block valve are 3.
Plant cooldown in compliance with Isranch Techni-provided for plant operational flexibility and for limiting cal Position RSil 5-1.
the number of challenges to the pressurizer safety valves, The block valve is installed upstream of the PORV be-cause of the potential for the PORV toleak or stick open.
PORVs also provide safety related functions for events For overpressure protection of the reactor coolant pres-beyond the design basis such as for reactor coolant system sure boundary (RCPII)at normal operating temperature venting, feed and bleed cooling, and anticipated transient Cnd pressure, the operation of PORVs has not been ex-without scram (NIWS) mitigation. All events beyond the plicitly considered as a safety related function, Also, an design basis, including the above three events, are not inadvertent opening of a PORV or safety valve has been strictly speaking within the scope of GI-70. Ilowever,im-analyzed in the final safety analysis reports as an antici-provements in the reliability and availability of the pated operational occurrence with acceptable conse-PORVs would improve the ability of some plants to pro-quences. For these reasons, most PWRs, particularly vide venting of noncondensabic pases from the reactor those licensed prior to 1979, do not have safety related coolant system and to mitigate an NIWS event, and PORVs. The valve operators and their electrical control would provide additional assurance of feed and bleed ca.
systems are normally designed to non safety related stan-pability for those plants that include this technique in dards. Ilowever, the pressure retaining elements of their emergency procedures. In addition, further risk im-PORVs and block valves are within the RCPil and are plications from low-temperaturc overpressure protection constructed to the same codes and standards as those re-were studied, as reported in NUREG-1326 (Ref.1), as a quired for similar safety related RCPil components, separate action, Generic issue 94, " Additional low.
Temperature Overpressure Protection for Light Water Reactors." Generic Issue 84,"CE PORVs,"is separately The Three hiite Island Unit 2 ('thil-2) accident focused evaluating the need to upgrade or install PORVs to im-attention on the reliability of PORVs and block valves prove the reliability of the decay heat removal function.
since the malfunction of the PORV at Thil-2 contributed In support of the resolution of GI-70, the Oak Ridge to the severity of the accident. On other occasions National laboratory (ORNL) performed a study of pORVs have stuck open,when called upon to function.
PORV and block vahc operating experience. A report.
Also, there are PORVs in many operating plants that prepared by the ORNL Nuclear Operations Analysis have leakage problems so that the plants must be oper.
Center, was issued as NUREG/CR-4692 (Ref. 2). 'this
,ated with the upstream block valves in the closed position.
work was sIionsored I ' the NRC Office of Nuclear R#8u-D Ihe technical specifications governmg PORVs on most 1
opemting PWRs that deal with closing the block valve and latory Resyarch as a part of the Nuclear Plant Aging Research I rogram, removing power were developed to prevent excessive leakage through the valves and were not developed to en-Ilrookhaven National Laboratory (IINL) also performed sure the operability of the PORVs. Following the'lhil-2 a study that estimated the risk reduction from improved accident, the staff began to examine transient and acci' PORVand block valve reliability, llNL prepared a report dent events in more detail, particularly with respect to re-that was issued as NUREO/CR-4999 (Ref. 3). This study quired operator actions and equipment availability and showed only a small pot ential decrease in core melt prob-performance. As a result, the staff initiated an evaluation ability due to increased PORV and bkick valve reliability.
of the role of PORVs in accident management and miti-This was in part because by staff direction the study did pation.
not include consideration of feed and bleed capability.
'lhe classification of PORVs and thck valves should be Over a period of time, the role of PORVs has changed consistent with the system used forclassifying other com.
such that PORVs are now relied upon by many West-ponents of the RCPil and those other systems that per-inghouse, llabcock and Wilcox (ll&W), and Combustion form a sMety-related function as defined in Regulatory Engineering (CE) designed power plants with PORVs to Guide L26 (Ref. 4). While PORVs were originally pro-perform one or more of the following design basis safety-vided on PWRs for plant operational flexibility and for related functions:
limiting the number of challenges to the pressurizer j
xi NUREG-1316
)
p p
I safety valves, the operation of PORVs for overpressure protection. Stroke testing of the PORVs should not protection of the RCPU was not considered to be a safety-be performed during power operation. The staff has related function.
concluded that stroke testing during power opera-tion,in the words of the AShiE Code,"is not practi-However, since PORVs are relied upon in many PWR cal" (reference ASME Section XI, Paragraph
. plants to mitigate the design basis accidents identified IWV-3412) because of the potential for a PORV to
. above or to perform any other safety related function that stick open during the stroke test. In addition, PORV L
may be identified, the staff finds that it is appropriate to block valves should be specifically included in the reconsider the safety classification of PORVs and the as-scope of safety related motor-operated valves p
sociated block valves.
(MOVs) addressed in the resolution of Generic F
Issue ll.E.6.1, "In Situ Testing of Valves," in NRC
- For future PWR plants when PORVs and the associated Generic Letter 89-10 (Ref. 5).
block valves are used for any of the safety related func-tions discussed above, these components should be classi-3.
For operating PWR plants, modify the limiting con-ficd as safety related and a minimum of two PORVs and ditions of operation of PORVs and block valves in block valves installed. Certain recently licensed plants the technical specifications for Modes 1,2, and 3 to and plants currently under active construction that have incorporate the staff position adopted in recent li-solenoid pilot-operated PORVs* such as Vogtle 1 and 2.
censing actions. 'Ihat is, ensure that plants that run Millstone 3, Callaway, and Wolf Creek rnect these re-w th the block valves closed (e.g., due to leaking quirements.
PORVs) maintain electrical power to the block valves so they can be readily opened from the con-For operating PWR plants and construction permit (CP) trol room upon demand. Additionally, plant opera-holders, therc arc a number of potential improvements to tion in Modes 1,2, and 3 with PORVs and block PORVs and block valves (short of upgrading to fully valves inoperable for reasons other than seat leak-safety-grade hardware) that can increase the reliability of age is not permitted for periods of more than 72 these components and provide assurance that they will hours.
function as required when called upon to perform a safety related function.11 is anticipated that the reliabil-4.
Use, to the extent possible, more reliable PORV ity of PORVs and block valves can be increased by imple-menting the following improvements:
and PORV block valve designs that are resistant to failure. 'the NRC recognizes that licensees may 1,
include PORVs and block valves within the scope of choose to replace existing PORV and PORV block an operational quality assurance program that is in valves with more reliable designs as they are made compliance with 10 CFR Part 50, Appendix H.This available by valve manufacturers in the future. The program should include the following elements:
use of more reliable valves should result in less fre-quent corrective maintenance and can result in a.
The addition of PORVs and block valves to the longer inservice testing intervals as delineated in plant operational Quality Assuranec List.
Section XI of the ASME Boiler and Pressure Vessel
- Code, b.
Implementation of a maintenance / refurbish-ment pregram for PORVs and block valves that is based on the manufacturer's recommenda-For new construct. ion, the staff concludes that a minimum tions or guidelines and is implemented by of two PORVs and block valves and associated controls trained plant maintenance personnel, for these components should be provided.'these compo-nents should be identified as safety related if reqmred to 2,
include PORVs, valves in PORV control systems, perform any of the safety-related functions discussed and bhick valves within the scope of a program cov, above or to perform any other safety-related function that cred by Subsection IWV, " Inservice Testing of may be identified in the future and should therefore bc Valves in Nuclear Power Plants," of Section XI of constructed to safety grade standards, the ASME Boiler and Pressure Vessel Code. As permitted by the Code, stroke testing of PORVs This would include redundant and diverse control sys-should only be performed during Mode 3 (HOT tems, designed to Seismic Category I requirements and i
STANDHY) or Mode 4 (llOT SHtTI'DOWN) and environmentally qualified; increased technical specifica-in all cases prior to establishingconditions where the tion surveillance requirements; increased inservice test-PORVs are used for low-temperature overpressure ing requirements; and inclusion within the scope of a quality assurance program that is in compliance with 10 CFR Part 50, Appendix B.The safety related designation
' As of the date of this report.11cliefonte 1 and 2 and WNP-1 are not would include those improvements that were imposed i
mnsidered to te planis under active construcibn.
subsequent to the TMI-2 accident, such as requirements NUREO-1316 xii i
=
to be powered from Class 1E buses and to provide valve bleed only reduces the core melt frequency by a fraction position indication in the control room.
of the improvements indicated, it is a substantial reduc-tion in public risk.
1hc staff also concludes that item 4, identified above, should be implemented where possible in new construc.
The proposed improvements to PORVs and block valves tion and strongly encourages operating reactor owners to identified above enhances (but does not ensure) feed and evaluate the benefits of replacing existing PORVs and bleed because:
block valves with more reliable designs.
1.
Inclusion within an operational quality assurance program that is in compli:. ace with 10 CFR Part 50, For operating plants and CP holders, the staff concludes Appendix II, of an improved PORV maintenance /
it is not cost effective to upgrade (backfit) existing non-refurbishmcnt pro [lram and additional surveillance safety grade PORVs and bkick valves (and associated testing provide better assurance that PORVs will control systems) to full safety-grade qualification status open or close when called upon.
when they bave been determined to perform any of the safety related functions discussed above or to perform 2.
Currently, certain plants operate with the block any other safety related function that may be identified in valves closed. The techmcal specifications for these the future. Subsequent to the TMI-2 accident, a number plants require that power be racked out at a valve of improvements were required of PORVs, such as re-motor control center, making it unlikely that feed quirements to be powered from Class 1E buses and to and bleed could be initiated in a timely manner.The have valve position indication in the control room.There-proposed revised technical specifications require fore, additional improvements that would result from up-those plants that run with the block valves closed grading PORVs to fully safety-grade status are consid-(e.g., due to leaking PORVs) to maintain c!cctric cred to be of marginal benefit. For operating plants and po'ver to the block valves so they cm be readily CP holders, the greatest benefits can be derived from im-opened from the control room.
plementing items 1,2, and 3 identified above. 'the staff is proposing that these requirements be imposed to in-3.
Placing the block valves within the scope of Generic crease the reliability of PORVs and block valves to pro-Issue 11.11.6.1, as reported in Generic Letter 89-10, vide assurance that they will function as required. Items 1, would provide increased assurance that the block 2, and 3, which do not require hardware changes, can be valves would open against system differential pres-implemented within the scope of current licensing crite-sure to permit initiation of feed and bleed.
ria and coordinated with the Technical Specifications Im.
provement Program.
Ilased on the above, the staff concludes that the proposed actions to PORVs and bhick valves provide a substantial As noted above, the llNL study performed specifically in increase in the overall protection of the public health and support of GI-70 did not include consideration of feed
- safety, and bleed capability. In the course of the resolution of Urresolved Safe'y issue (USI) A-4$ as reported in The staff estimates that the outage avoidance costs, based NUREG/CR-5230 (Ref. 6), the use of feed and bleed on industry data reportec' by EPRI, would far exceed the cooling on the primary system as an alternative measure cost of implementing items 1,2, and 3. Specifically, the to remove decay heat from the reactor core was explored present value associated with the improvements to in some detail.These studies in general support the con-PORVs and block valves for items 1,2, and 3 identified cept of feed and bleed.The effect of feed and bleed upon above are estimated to be $127,200 for a plant with two the probability of core melt was examined, and the report PORVs and two bhick valves. The present value of the indicates that this capability reduces the estimated core outage avoidance cost is estimated to be $2,541,000.The melt probability for internal events by a significant overall cost benefit is estimated to result in a savings of amount on the order of 25 to 90 percent. Even if feed and
$2,413,800 per reactor.
xiii NURl!G-1316
t 1 INTRODUCTION non-sarciy-grade PORvs to mitigate an SOTR design ba-sis accident was raised in Reference 8.
Ilxamining Oeneric Issue 70, " Power-Operated Relief As discussed in Section 2.2 of this report, PORVs are also Valve and Illock \\ ahe Reliability, myolves the evalu-relied upon to serve other functions such as low-ation of the reliability of power-operated relief valves temperature overpressure protection (LTOP) during (I ORVs) and block valves and their safety signtricance in plant cooldown in compliance with IITP RSil 5-2 in Stan-I WR plants. I raditionally, the PORV and its block valve dard Review Plan 5.2.2. In addition, PORVs provide are provided for plant operational flexibility and for safety benefits in events beyond the design basis such as limitmg the number of challenges to the pressurizer reactor coolant system (RCS) venting, feed and bleed safety valves. Ihc block valve ns installed upstream of the cooling, and ATWS mi'igation*
PORV because of the potential for the PORV to leak or stick open. Figures 1.1 and 1.2 show two typical styles of Considering the potential safety related functionsassoci.
4 i ORVs currently m general use m i W R plants.
ated with the PORVs and block valves, in addition to the Iigure 1.1 is representative of a pilot operated relief original PORV loss-of-coolant accident concern, it is ob-valve, and Figure 1.2 is representative of an air-operated v ous that there is a need to reassess the PORV and block (spring loaded) relief sahe. *Ihese two general types of valve with respec: to the safety related requirements in PORVs are discussed in greater detail in Reference 2.
order to determine if improvements are necessary to I,igure 1.3 shows tl c installation of PORVs and block valves on a typical IlWR plant. For overpressure protec-plants with non-safety related PORVs and block valves to ensure reliable PORV and block valve operation.
tion of the reactor coolant pressure boundary (RCPil)at normal operating temperature and pressure, the opera.
tion of PORVs has not been explicitly considered as a 2 BACKGROUND safety-related function. Also, an inadvertent opening of a PORV or safety valve has beer. analyzed in final safety The Three Mile Island Unit 2 (TMI-2) accident focused analysis reports (FSARs) as an anticipated operational attention on the reliability of PORVs and block valves since the malfunction of the PORVat TMI-2 contributed occurrence with acceptable consequences. For these rea.
sons, most PWRs, particularly those licensed prior to to the severity of Ihe accident. On numerous occasions, as 1979, do not have safety related PORVs. The valve reported in Reference 2 PORVs have stuck open when these valves were called upon to function in operating operators and their electrical control systems are nor, plants. Also, there are PORVs in many operating plants mally designed to non safety rclated standards.110 wever, the pressure-retaining elements of PORVs and bh>ck with leakage problems so that the plants must be oper-valves are within the RCPil and are constructed Io the ated with the upstream block valves in the closed position.
same codes and standards as those required for similar safety-related RCPil components. Some plants licensed The technical specifications governing PORVs on most prior to 1979 do maintain PORVs and bk)ck valves in an operating PWRs that deal with closing the block valve and operational state to protect against challenges to the removing power were developed to prevent excessive pressurizer safety valves, although other piants do not.
leakage through the valves and were not developed to en-sure the operability of the PORVs.
Prior to the Gin'na SGTR event in January 1982, the 1here have been no specific safety requirements ad-thermal hydraulic performance of SGTR events was not dressed in existing regulatory guides or rtandard review explicitly evaluated in licensing reviews. Instead, the re-plans (Ref 7)for the PORVs and block valves except for view of the SGTR event emphasized the radiologicalcon-compliance with Ilranch Technical Position (llTP) RSil sequences, and very general, unverified assumptions 5-1 in Standard Review Plan 5.4.7.1-lowever, as discussed were made regarding the system performance. Following in Sections 2.1 and 2.2 of this report, the staff haslearned the TMI-2 accident, the staff began to examine transient that,with the exception of recently designed Combustion and accident events in more detail, particularly with re-Engineering (CE) nuclear power plants without PORVs, spect to required operator actions and equipment avail.
PORVs are relied upon in Westinghouse and llabcock &
ability and performance. In addition, a reactor coolant Wilcox (H&W) designed plants to mitigate a design basis pump seat leak occurred at li.B. Robinson Unit 2 on steam generator tube rupture (SGTR) accident and, as November 30,1981, in which recovery was aggravated by such, the staff has conside red PORVs to perform a safety-malfunctioning pressurizer relief and block valves. As a related function. Owners of some recently licensed PWR result, the staff initiated an evaluation of the role of plants, in responding to staff questions regarding this reli-PORVs in accident management and mitigation. Finally, ance on PORVs, have made the PORVs on their plants the staff specifically reviewed the role of PORVs in safety related and designed them to safety grade stan-SGTR management and mitigation following the Ginna datds. For older plants, the acceptability of relying on SGTR event.
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\\ -k
.i i
i A
f 't 0~
-Q n
t p
~
~~
'/
<1
//s i
Figure 1.2 Air operated (spring-kuded) relief valve (Courtesy of Copes-Vulcan).
[
i, s
Nm\\
Reutr 2/"
BlDct YRUS 4;:A N
Ntorn Yans A
russumzI:n a
m f
.7-::...--As
'~
g.4.., g 7,.,,7,.,,_3,,,
)
coot 4x7 TAhT thoPs
/
J Figure 1.3 Pressurizer safety and relief valves.
NUIEG-1316 4
)
Ilased on the above cf forts, the staff concluded that PWR Review Plan 5.2.2, are to be single failure proof, testable, plants rely on a rapid primary system depressurization ea-designed to Section lit of the ASMi!iloiler and Pressure pability in order to limit the primary to+ccondary leakage Vessel Code and powered from essential buses. IrrP RSit (and thus limit the radiological release to the secondary 5-2 also notes that 11!!:11-279' should be used as guid-Systems) assumed in the SGTil licensing analyses. In ante in the design of IJOP systems and further specifies SGTil scenarios where the reactor coolant pump flow is that ! 'IOP systems should be designed to function during lost (i.e., loss of offsite power for compliance with Gen-an operating basis carthquake and not during a safe shut-Gral Design Criterion 17), Westinghouse and likW plants down carthquake. The LTOP system requirements were rely on the pressurizer PORVs, which are, in most plants, implemented as Multi Plant Action item 11-04. As noted designed to non safety related requirements.
in Section 2.3 of this report, when PORVs are used for high point vents in some plants, in accordance with item
!!.11.1 of NURiiG-0737 (Ref. 9), both PORVs and block in certain new plants without PORVs, designed by Cli, valves are required to be seismically and environmentally such as San Onofre Units 2 and 3 and Waterford Unit 3, this depressurization function is accomplished by a safety-qualified.
related auxiliary pressuriier spray (APS) systern. Ilow.
For PWRs licensed before 1982, there are no technical ever, m the Cll-designed CliSSAR System 80 plants, such specification requirements that these components be op-as I alo Verde Units 1-3, the Al S system (includmg its crational when the plant is at power, Continued opera-water supply)is not fully designed to safety related re-tion at power with inoperable PORVs is permitted by the quirements and is identified by the NitC staff in Supple-technical specifications if the block valve is closed and ment 3 of the CliSSAR safety evaluation report as an un-power to the block valve (s)is removed. Many plants now resolved item subject to resolution, it is not clear to the operate with the PORVs blocked, staff whether Cli plants with PORVs rely on their PORVs or APS system for mitigation of an SGTR event, West nghouse PWRs licensed since 1982 have upgraded nor is it known whether these systems are designed to technical specifications that permit plant operation only safety related standards, f lowever, this matter will be re-for periods up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with PORVs or block valves solved as part of Generic issue 84,"Cl! PORVs," which is considered inoperable for any reason other than excessive separately evaluating the need to upgrade or install seat leakage.
PORVs to improve the reliability of the decay heat re-moval function.
IITP RSil 5-1 in Standard Review Plan 5.4.7 requires that plants licensed afier Jan uary 1978 be capable of cooldown Although Westinghouse and some !!&W plants have an to cold shutdown conditions using only safety related auxiliary pressurizer spray system, it is not designed to equipment (IrrP RSil 5-1 allows some relief from this safety-related requirements and is not designed for use position for plants whose construction permit was dock.
ehen engineered safety features are actuated. The staff eted before January 1978), it may be necessary to use has notified ail licensing boards associated with PWRs de.
safety grade PORVs for plants without safety related signed by Westinghouse and likW of the staff findings re.
auxiliary pressu rizer spray syst ems in order to comply with garding reliance on PORVs for SOTR mitigation, this staff position, item II.D.1 of Reference 9 requires all plants to demon.
In most plants, the LTOP system is designed to use the strate the ability ofthe PORVs and bhick valves to func.
PORVs. For this mode of operation, the valves are typi-tion under all flow conditions expected during transient cally set to open at approximately 500 psig rather than the and accident conditions. It also requires that the block high pressure (approximately 2300 psig) setpoint used at valves be capable of closing to ensure isolation of a stuck-power, Westinghouse and some Cli designed plants use open PORV. In response to this requirement, PORVs redundant PORVs for Ilf0P concerns.
were tested extensively by the Electric Power Research Institute (IIPRI) and the results reported in Reference These plants are brought to a water solid condition dur.
- 10. Limited bhick valve testing was also performed as a ing shutdown. In contrast, ll&W owners use a single part of the !!PRI test program.
PORV, and the gas (steam or nitrogen) space in the pres-surizer functions as the primary IXOP system. The item II.D.3 of Reference 9 requires direct indication of PORY and associated actuation circuitry function as a PORV position. Item 11.0.1 of Reference 9 requires backup should the operator fail to terminate a low.
emergency power for PORVs and bhick valves, temperature overpressure challenge prior to compres-sion of the gas space. In the new CII plants without PORVs, low temperature overpressure protection is pro-6 d b im si.w braio3wch is
,linstanerd mis bn n rek.153,"[riteria for Iwer,Instrutnen-vided by relief valves on the shutdown coolmg system, endorsed by Regulatory Guide 1l1 OP systems, as specified in Ir! P RSil 5-2 in Standard nation. and Control l'ornons of safety syuenc 5
NURl!G-1316
l 2.1 Safmy Fur.ctions of PORVs and safety valves will then Itft, allowing the leaked reactor H lot.k V(tlyes coolant to escape directly to the environment. To prevent this situnu,on from occurring, the primary pressure must in PWR plants, PC RVs and associated block valves were be rapidly decreased to stop the primary to-secondary originally provided 10. plant operational flexibility and for leakage. 'Ihis depressurization can be accomplished in a limiting the number of challenges to the pressuriier variety of ways, meluding (1) the use of the normal safety valves. For overpressure protection of the RCPil at pressurizer spray that is available only when the reactor normal operating temperature and pressure, the opera.
C00lADI pumps are running: (2) the use of the auxiliary tion of PORVs had not explicitly been considered as a pressurtzer spray, which does not require the reactor safety related function because of the availability of the coolant pumps but rather derives its flow from the charg-safety-related pressurizer safety valves. Therefore, these mg pumps; or (3) opening the PORV and discharging components were designated as non-safety-related be, steam from the pressurizer steam space. The Westin-cause they were required neither to safely shut down the ghouse, ll&W, and CI! plants with PORVs rely on the plant nor to mitigate the consequences of accidents.
pressurizer PORV to accomplish this depressurization whenever the reactor coolant pumps are not operating, llowever, over a period of time the role of PORVs has llowever, current Cli plants without PORVs apparently changed such that PORVs are now relied upon by many rely on the auxiliary pressurizer spray system to keep the Westinghouse, B&W, and C11 plants with PORVs to offsiteradiological consequences within regulatory lim-perform one or more of the following design basis safety-its. The ability of these CII-designed plants without related functions:
PORVs to meet regulatory requirements is discussed in NURf!G-1044 (Ref.11). In addition, Generic issue 84, 1.
Mitigation of a steam generator tube rupture "Cl3 PORVs,"is separately evaluating the nced to install
- accident, PORVs to improve the reliability of the decay heat re-moval function.
2.
Iow temperatureoverpressureprotectionof there-actor vessel during startup and shutdown, or 2.2.2 Low Tetuperature Overpressure Protection 3.
Plant cooldown in compliance with IITP RSil 5-1.
When the PWR reactor coolant system is in a cold :. hut-in addition to these design basis safety related functions, down condition, the maximum allowable pressure in the PORVs also provide safety-related functions for events reactor vessel is low because of vessel irradiation and em-beyond the design basis such as for reactor coolant system brittlement. The inadvertent starting of a high. pressure venting, feed and biced cooling, and KlWS mitigation.
safety injection pump can result in an overpressure tran-sient. To ensure that in these situations the maximum All events beyond the design basis, including the above three events, are not strictly speaking within the scope of pressure remains below the limits specified in the license Generic issue (GI) 70. Ilowever, improvements in the re-technical specifications, a low-temperature overpressure protection system (!!!' OPS) must be available, irrP RSil liability and availability of the PORVs would improve the 5-2 in Standard Review Plan 5.2.2 states the functional ability of some plants to provide venting of nonconden-requirements for this system, but does not specify a par-sable gases from the RCS, as discussed in Section 2.3 of ticular mitigation technique, in addition, Generic issue this report, to mitigate an ATWS event, and would, as dis-i cussed in Section 5.3 of this report, provide additional as-94," Additional low-Temperature Overpressure Protec-tion for Light Water Reactors," as a separate action surance of feed and bleed capability for those plants that evaluated further risk implications from IlrOP due to the include this technique in their emergency proccuures, continuing occunence of overpressure transient events after the completion of Generic Issue A-26, " low Tem-2.2 Description of PORY Safety perature Over Pressure Protection," that was resolved by Functions Multi-Plant Action 11-04. As a part of the resolution of Generic Issue 94, the staff prepared a regulatory analysis This section provides a description of the safety-related based on work performed by llattelle Pacific Northwest functions that may be performed by PORVs on PWR Iaboratories; this analysis is reported in NURI!G-1326 plants.
(Ref.1).
Most PWR designs use PORVs as a means of mitigating 2.2.1 Stcaut Gencrator Tube Rupture low-tempyaturme[ pressure transients. In these plants, the PORN setpomt is manually lowered to around 500 in the event of an SGTR, leakage of reactor coolant from psig at low RCS temperatures, and, should the RCS pres-the primary system to the secondary system will eventu-sure reach this value, the PORY opens to limit system ally pressurize the secondary system. The secondary pressure. Westinghouse and some of the Cli plants with NURIIG-1316 6
pressure. Westingho' use and some of the CE plants with vivability of various electrical components and cables PORVs use redundant PORVs for low temperature associated with the PORV. There was considerable un-overpressure transients, whereas B&W plants use the certainty as to whether the PORV in its post accident pressurizer gas space as a primary means of controlling condition, that is, severely corroded from years of expo-overpressure and the single PORV as a backup. In the sure in a high humidity and a high radiation enviror tr.ent, current CE-designed plants without PORVs, low-would provide any conclusive evidence as to the cause of temperatu.e overpressure protection is provided by relief its failure to close.
valves on the shutdown cooling system.
Because of funding and scheduling difficulties, it was de-cided that the effort to identify the failure mode of the 2.2.3 Plant Cooldown in Compliance with TMI-2 PO RV was not justified on a cost / benefit basis as a Branch Technical Position RSB 5-1 part of the work performed under GI-70 and the effort Branch Technica! Position (IrTP) RSB 5-1 in Standard was therefore not undertaken.
Review Plan 5.4.7 8tates that current PWRs should have safety grade systems capabic of maintaining the RCS in 2.5 NRC Information Notice 89-32, the hot standby condition for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, followed by a cool.
down to the cold shutdown condition. In Westinghouse, Surveillance Testing of low-B&W, and CE plants with PORVs, depressurization of Temperature Overpressure the RCS is accomplished by using a combination of either Protection Systems RCS fluid contraction caused by the cooldown and heat losses from the pressurizer to ambient or by a safety-The staff in information Noticc 89-32 (Ref.12) expressed related PORV, llowever, the new CE plants without concern with respect to potential plant operability prob -
PORVs apparently rely on the auxiliary pressurizer spray lems due to lack of inservice testing of PORVs in their system as if it were a safety-related system.
LTOP mode. The staff noted that in three cases (identi-fied in the information notice) valve opening times that were in analyses of the licensee's LTOP systems were not 2.3 Reactor Coolant System Venting being transferred into inservice testing requirements and eventually into plant surveillance test procedures. A Following the TMI-2 accidcat, a number of additional significant increase in valve opening time could result in requirements were imposed on reactor plant applicants exceeding the limits of Appendix G to 10 CFR Part 50.
and licensees. One of these requirements, item II.B.1 of Reference 9, requires that high point vents be installed in Paragraph IWV-3400 of Section XI of the ASME Boiler PWRs for the purpose of venting from the reactor coolant and Pressure Vessel Code requires valves to be exercised system noncondensable pses that may inhibit natural cir-to the position required to fulfill their safety function, culation and adversely affect core cooling during loss of Therefore, nct testing the PORVs in the open direction is offsite power events. When PORVs are used for high not in accordance with the Code, point vents, both the PORVs and block valves are re-quired to be seismically and environmentally qualified As noted in Section 5.2 of this report, it is anticipated that and included within the scope of an inservice testing pro
- the reliability of PORVs and block valves can be in-gram that is in confarmance with Section XI of the ASME creased by implementing the improvements prescnted in Boiler and Pressure Vessel Code.
Table 5.1. Specifically, item 2 of Table 5.1 in part recom-mends that stroke time testing of PORVs should be per-I ""ed in aa rd nc whh Subsection IWV of ASME 2.4 Three Mile Island Unit 2 PORY Section XI of the ASME Boi!cr and Pressure Vessel Code. When establishing stroke time acceptance criteria -
The failure of the PORV to reclose during the initial f r PORVs used in LTOP systems, licensees should take TMI-2 transient initiated the accident and contributed to int account the PORV opemng stroke time used in the its severity.The open or partially open PORV therefore setpoint analysis for the LTOI system.
provided a pathway in the RCPB for the release of accident-generated hydrogen, steam, and fission products directly from the primary system to the containment 3 CONTRACTOR REPORTS building.
l In support of the resolution of GI-70, the Oak Ridge Na-Consideration was given to removal of the PORV and a tional Laboratory (ORNL) performed a study of PORV section of the downstream piping with subsequent efforts and block valve operating experience. A report, prepared
. to examine then components.The intent of the examin-by the ORNL Nuclear Operations Analysis Center, was ation was to determine, if possible, the actual failure issued as NUREG/CR-4692 (Ref. 2). This work was mechanism of the PORV and also to determine the sur-sponsored by the NRC Office of Nuclear Regulatory 7
r I
t Research as a part of the Nuclear Plant Aging Research A summary of identified PORV mechanical failure and
-progrm degradation modes by reactor vendor is shown in Table 3.2.
Bro-w.n National Iahoratory (HNI.) also performed I
a study that estimated the risk reduction from improved Table 3.2 Failure a~l degradation modes-PORV/bhick valve reliability. BNL prepared a report PORY m echanical.
_ that wr.s issued as NUREG/CR-4999 (Ref. 3).
Il& W W
CE Total 3.1 NUREG/CR-4692 Ixakage-internal 19 33 10 62 Leakage-external 3
3 NUREG/CR-4692 (ORNUNOAC-233) contains a re.
Failure to open 3
8 1
12 view of nuclear power plant operating events involving Failure to close 4
3 7
failure of PORVs and associated block valves.The report Other 6
6 5
17 reviewed events reported from 1971 to mid-1986. Each Total 32 53
.16 101 1
PORV and bk>ck valve event was judged as to the severity of the failure, and the terms chosen to identify the degree of failure were as follows:
A review of the PORV control failure modes indicates the most common control failure or degradation mechanism 1.
" Degraded,, (but operable), the component oper-for PORVs (57%) involved problems with the air or elec.
ated at less than its specified performance level, and trical actuation controls that would have prevented op-cration of the PORV ifit had been required. Twelve per-cent cf events where the PORV unintentionally opened 2.
" Failed," the component was completely unable to resulted mostly from inadvertent or accidental actuation perform its function, by human error.
A compilation of PORV and block valve events with re-A summary f identified failure and degradation modes f r PORV controls by reactor vendor is shown m Table O
spect to failure severity is shown in Table 3.1.
Table 3.1 Failure severity.
Table 3.3 Failure and degradation modes-PORY controls.
Degraded Failed Total H&W W
CE Total PORV mechanical 77 24 101 Failure to open 3
2 1
6 PORV control 30 61 91 Failure to close i
1 2
PORV design 6
0 6
Spurious opening 1
4 6
11 -
Bhick valve events 17 15 32 Control degmded 1
49' 2
52 Total 130 100 230 Other 4
12 4
20 Total 10 68 13 91 b
- 'lwenty five events involved recurring pmblems with nitrogen con-Thus 23 percent of the PORV mechanical events and 67 tml systems at North Anna 1 and 2.
percent of the PORV control events were failures. Forty-seven percent of the bhxk valve events were also failures.
For bhick valves,37 percent of the events involved exter-nal leakage, and 37 percent involved failure of the block A review of the PORV mechanictd failure modes indi-valve to close on demand. Such a failure can pose a threat cates the most common mechanical degradation or failure to safety if it occurs in coincidence with a stuck open mechanism for PORVs appears to be deterioration of the PORV For this reason, the ability to close is the most im-seat / disc interface or other internal parts by high-portant function for PORV bk)ck valves. As noted above,
_ pressure steam and/or water.This results in internal leak-there are a number of PORV internal leakage events age through the valve seat into the valve outlet tailpipe (62%), and many plants operate with the block valve and was the most common failure mode apparent from closed when the unit is at power. Therefore, under these l
the study (61%)while failurc of the PORV to open was 12 circumstances it is also important that the bk>ck valve be
~
percent and failure to close was 7 percent, able to open reliably as well as close.
y p-L A summary of identified failure and degradation modes ing a more reliable PORV design. 'lhe reliability of exist-
= for block valves is shown in Tab!c 3.4.
ing PORVs and block valves would be enhanced by improved surveillance testing, advanced diagnostic techniques where applicable, and maintenance applied to Table 3.4 Failure and degradation modes-PORVs and block valves, particularly the block valve mo-block valves.
tor operator.
Leakage-external 12 ORNL also interviewed four PORV manufacturers in or-
$[t]
der to obtain their views related to manufacturing, instal.
lation, testing, maintenance, and operation of these Spurious opening 3
valves and any feedback from utilities or problems en-Other 3
countered during operation of PORVs.
Total 32 A review of events collected for NUREG/CR-4692 indi-cates that Dresser and Crosby pilot valve designs ac-In NUREG/CR-4999 (ilNlrNUREG-52101), an analy-counted for 40 percent of the PORV mechanical failures.
sis was performed to explore the risk reduction potential
'Ihese designs were involved in failures that occurred at of improving the PORV and block valve reliability for two all nine ll&W plants. (Most CE units have blocked off representative PWR plants, Indian Point 3 and Oconec 3.
their PORVs or do not employ them in the design.)
Existing probabilistic risk assessments (PR As), pertinent event trees, fault trees, and the equipment reliability data -
Table 3.5 presents a compilation of PORV mechanical presented in the Indian Point probabilistic safety study failurcs and degradation listed by PORV manufacturer. It and in the Oconce PR A were used to quantify the bene-should be noted that the Dresser and Copes Vulcan de-fits of improved PORV and bhick valve reliability in t erms signs have been in use for a number of years, hence the of potential reduction in core melt frequencies, llecause relatively high total number of events, of their importance, attention was focused upon those safety related functions identified in Section 2.1, namely, an SGTR accident, the use of PORVs in reactor vessel Table 3.5 PORY mechanical failures and degradation.
LTOP events, plant cooldown in compliance with IrrP y
RSil 5-1, and the feasibility of using PORVs as high point wns supp m n ned ns reactor usM PORY Manufacturer Fal 1
)e ed Total head vent system, in addition to the above, a stuck-open
' PORV was also studied.
Crosby (p)*
2 5
7 s
! Dresser (p) 8 25 33 The core melt frequencies attributable to PORV or block 5
5 valve failures were found by UNL to be relatively insig-Oarrett (p) 3 25 28 nificant and to represent only a very small fraction of the Copes-Vulcan (a)"
' Masoncilan (a) 3 7
10 total core melt frequency attributable to internal plant Control components (a) 2 2
events. Specifically, llNL results show a potential reduc-
. Unknown 8
8 16 tion in core melt frequency of about I to 3E-7 for the Total 24 77 101 SGTR and stuck-open PORV events. For LTOP cvents, the llNL study showed a potential reduction of core melt frequency of about 2E-6. It should be noted that BNL
- (p) PiloFoperated.
"(a) Air operated (spring elase).
used information for the plant system reliabilities pro-vided in the utilities' PRA documents.
As noted in Appendix D to NUREG/CR-4692, PORV control systems for PWRs are not provided by the valve The staff believes that the UNL results, which were
. manufacturer. These control systems are usually pro-largely based on the Indian Point 3 (Ref.13) and Oconce vided by the nuclear steam supply system supplier, (Ref.14) PRAs, underestimate the safety benefit that architect-engineer, or utility.There is, therefore, no com.
would be achieved by improving PORV and block valve parable table to Table 3.5 for PORV control failures and reliability for the following reasons:
degradation by PORV manufacturer.
1.
The Indian Point 3 and Oconce PRAs that were the An assessment of the need to upgrade PORVs and bk)ck basis of the HNL study used PORV failure rates that valves to safety related status concludes that such action were as much as two orders of magnitude lower than l
would improve PORV and bhick valve reliability. ORNL the failure rates determined by ORNL in NUREO/
believes the greatest improvement would result from us-CR 4692.
t 2.
The llNL study did not consider that older PWRs the same construction as other motor-operated gate are permitted to operate at power indefinitely with valves in PWRs and are constructed to the same codes the block valves closed and power removed. In this and standards as those identified above for PORVs.
operating - configuration, the PORVs are not
- available to perform the safety functions listed in Section 2.1.
4.2 Seism. Des.ign ic sm a ca n f PORVsand blockvalvesisan 3.
Ily staff direction, the BNL study did not include consideration of feed and bleed capability' (See Sec-
@ pre appears to k no unHonn appkadon area tion 5.3.)
of scismic design requirements. Since 1972, Regulatory G uide 1.29 (formerly Safety Guidc 29)(R ef.15) has speci.
Although the staff believes that items 1 and 2 above result fied that the RCPB should be seismically designed to in a somewhat low estimate of the potential reduction in w thstand the effects of a safe shutdown earthquake core,rncit frequency by BN1 the BNL fault trees are (SSE). that is Seismic Category 1. Ilowever, unless spe-dominated by operator error considerations, particularly cifically requested by the customer, PORVs were not nor-4 for the SGTR event. Therefore, it does not appear that mally qualified to Seismic Category I requirements.
the results would have changed a great deal even if BNL had used higher PORV failures and considered that 4.3 Quality Assurance plants often operate with the block valves closed and power removed. llowever, as discussed in Section 5.3, the As noted in Appendix D to NUREG/CR-4692, PORVs consideration of feed and bleed indicates a much greater are generally constructed to a manufact"rcr's quality as-safety importance of PORVs and block valves, surance program that is in compliance with 10 CFR Part 50, Appendix B. The manufacturers' quality assurance programs have been in effect at least since the introduc-4 CONSTRUCTION OF PORVs AND tion of valves into the 1971 Edition of Section til of the BLOCK VALVES
^SME Boiler and Pressure Vessel Code. Prior to 1971, PORVs were constructed to each manufacturer's quality Although most PWRs licensed prior to 1979 did not have assurance program.
safety related PORVs and block valves,it was recognized that the pressure-retaining portions of these components were a part of the reactor coolant pressure boundary 4.4 Control Systems (RCPB) as defined in 10 CFR $ 50.2. At the time they The control systems for PORVs and block valves are not were not considered to perform a safety-related function supplied by the valve manufacturer but are designed and (other than retaining reactor coolant system pressure) be-supplied by the PWR nuclear steam supply system sup-cause they were not required to s~ ut down the reactorand plier or the architect-engineer. For PORVs the control n
maintain it in a safe shutdown condition or to prevent or mitigate the consequences of accidents that could result systems consist of an external power supply that is pneu-matic or electrical for valve operation. Valve operation is in potential offsite exposures comparable to the guideline typically controlled by an electrical signal resulting from exposures of 10 CFR Part 100, Appendix A. For these high system pressure or by manual actuation from the reasons PORVs and block valves were not considered to be safety related, control room. The block valve is actuated by the motor operator and is manually controlled by an electrical signal from the control room. Prior to the TMI-2 accident, 4.1 Codes and Standards PORV and block valve control systems were not qualified to standards, such as IEEE-382. Post-TMI-2 PORV and Since the pressure-retaining portions of PORVs and block valve cont rol systems are now generally qualified to block valves perform the same safety-related function as IEEE-382, IEEE-323, and IEEE-344.
other safety related pressure-retaining components of the RCPU, they are constructed to the same codes and 4.5 Motor Operators standards in conformance with 10 CFR 9 50.55a. As noted in Appendix D to NUREG/CR-4692 (Ref. 2),
With the exception of San Onofre Unit 1, in all U.S. reac-PORVs are currently constructed to Section HI, Class 1, tor designs to date, the PORV block valve is a gate valve of the ASME Boiler and Pressure Vessel Code. Prior to actuated by a motor operator. The basic design of the introduction of the 1971 Edition of Section IH of the block valve actuator is the same as for other actuators us-Code, PORVs were constructed to earlier codes and stan-ing an electric motor when used on gate valves in similar i
dards, such as the Draft ASME Code for pumps and applications of PWRs.The major manufacturers of elec-valves, manufacturer standards, USAS B31.1.0-1967, tric motor valve operators in the United States are and related standards, such as B16.5. Block valves are of Limitorque and Rotork.
- NUREG-1316 10
1.
Take no further action.
-406 Sarety-Related and Non-Safety.
Related PORVs and Block Valves 2.
Require operating plants, nii future PWRs, and those currently under construction (except for CE -
',Phere are several differences between PORVs and block plants without PORVs) to install safety-related
. valves that are classified as safety related and those PORVs and block valves and associated controls for PORVs and block valves that are classified as non safety-these components, related.
3.
Require (1) operating plants and those currently During construction, PORVs, valves in PORV control under construction to include those improvements systems, and block valves that are safety related are cur-presented in Table 5.1 of this report to existing non-rently constructed to a manufacturer's quality assurance safety grade PORVs and block valves, and (2) all
- program that is in compliance with 10 CFR Part 50, Ap-future PWRs (except for CE plants without PORVs)
- pendix 13, These components are designed to Seismic to install safety-related PORVs and block valves and Category I requirements and are environmentally quali-associated controls for these components.
. fled. The control systems of safety related PORVs and block valves are also constructed in accordance with the Alternative 1 to take no further action was rejected based quality assurance, seismic, and environmental require.
on an assessment of PORV failures as reported in ments identified above. Components that are non safety.
NUREG/CR-4692 (Ref. 2). If no action is taken, acci-related need not be constructed to any of the above qual.
dents that challenge PORVs and block valves may com-ity assurance, seismic, or environmental requirements.
promise plant safety. For example, on many PWRs the liowever, as noted in Section 4.1 of this report, the PORVs and block valves are not tested to verify their op-pressure-retaining portions of PORVs and block valves crational status and on other PWRs are operated with the are constructed to the same codes and standards as other PORVs in a degraded condition. Continued power opera-
' RCPil components, tion with inoperable PORVs is permitted by the technical specifications, and inany plants now operate in this condi-tion.The operational status of these PORVs is uncertain 4.7 Operating Plant Maintenance and they cannot be relied upon to perform a safety function.
For operating plants, prior to the TMI-2 accident, maintenance practices on PORVs and block valves and The part of Alternative 2 that requires operating plants associated control systems, including the block valve mo.
and those currently under construction to install safety-tor operators, varied widely from plant to plant.This wrs related PORVs and block valves or upgrade existing
' because of the perception that these components did not valves to safety grade status was rejected because the staff concluded that it is not the most cost effective perform a safety related function and were therefore not safety related.
method of achieving an acceptable level of safety. In the staff's judgment, an acceptable level of safety for existing PORVs and block valves can be achieved at less cost by PORVs that were degraded by excessive leaking were fre.
quently bkicked by closure of the associated block valve, other means.The installation of safety grade PORVs and Plant operation in this manner for plants licensed prior to bh>ck valves on operating plants could require redundant 1982 is permitted by the plant technical specifications.
and diverse control systems; components designed to The operational readiness of these degraded PORVs is Scismic Category I requirements and cavironmentally questionabic and cannot be ensured when they are called qualified; increased technical specification surveillance upon to open and close on demand, as happened in the requirements; mercased mservice testing requirements; case of TMI-2. In the post TMI-2 era, there appears to and inclusion within the scope of a quality assurance pro-be a greater awareness of the impact of malfunctioning gram that is in compliance with 10 CFR Part 50, Appen-
' PORVs and of the need for prompt operator action, dix 11. In addition, there may be rerouting of piping and
.However, as discussed in NUREG/CR-4692, it appears subsequent piping reanalysis. Impicmentation of this al-there is still a need for improvement in maintenance prac.
ternative would probably require additional plant outage tices of PORVs and block valves.
time beyond that normally required during refueling.The part of Alternative 2 that would require future PWRs to install safety-related PORVs and block valves and associ-5 REGULATORY ANALYSIS ated controls for these components is discussed as Part (1) of Alternative 3.
5.1 Alternatives Part (1) of Alternative 3 that requires operating plants and those currently under construction to include those The alternatives that were considered in the resolution of improvements presented in Table 5.1 of this report to ex-01-70 are as follows:
isting non. safety-grade PORVs and block valves was 11 NUREO-1316
F I
adopted because it was the most cost effective means of PORVs and bkick valves and associated :ontrol systems achieving an acceptable level of safety for these compo-for these components was adopted provided the licensee nents provided they perform one or more of the safety-performs one or more of the safety-related functions dis-related functions discussed in Section 2.1 of this report, cussed in Section 2.1 of this report. CE plants without The estimated annual utility costs required to implement PORVs are not considered to be within the scope of these i
those improvements identified in Table 5.1 are presented requirements. PORVs and block valves would therefore i
in Table 5.2. The overall cost benefit per reactor is pre-be classified in a manner that is consistent with other sented in Section 5.4 of this report. Part (2)of Alternative safety related componen':in PWR plants.
3 that requires all future PWRs to install safety-related I
i Table 5.1 Potentialimprovements to PORVs and block valves, 1,
include PORVs and block valves within the scope of secs' resp (mse to the expanded MOV test program an operational quality assumnce program that is in discussed in Generic Letter 89-10 (Ref. 5).
compliance with 10 CFR Part 50, Appendix II.This program should incluste the following elements:
3.
For operating PWR plants, modify the limiting con-ditions of operation of PORVs and block valves in a.
The addition of PORVs and block valves to the the technical specifications for Modes 1,2, and 3 to plant operat;onal Quality Assurance List.
incorporate the position adopted by the staff in re-cent licensing actions. The staff recognizes that b.
Implementation of a maintenance / refurbish-some recently licensed PWR plants already have ment program for PORVs and block valves that technical specifications in accordance with the staff is based on the manufacturer's recommenda-position. Such plants are already in compliance with tions or guidelines and is implemented by this position and need merely state that in their re-trained plant mamtepance personnel.
sponse. These recent technical specifications re-quire that plants that run wi,th the block valves closed (e.g., due to leaking PORVs) maintain elec-2.
Include PORVs, valves in PORV control systems, trical power to the block valves so they can be readily i
and block valves within the scope of a program cov-opened from the control room upon demand. Addi-cred by Subsection IWV,' " Inservice Testing of tionally, plant operation in Modes 1,2, and 3 with l
Valves in Nuclear Power Plants," of Section XI of PORVs and block valves inoperable for reasons the ASME Iloiler and Pressure Vessel Code. As other than seat leakage is not permitted for periods permitted by the Code, stroke testing of PORVs of more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
should only be performed during Mode 3 (HOT STANDilY) or Mode 4 (110T SHUTDOWN) and 4.
Use, to the extent possible, more reliable PORV in all cases prior to establishing conditions where the and PORV block valve designs that are resistant to i
PORVs are used for low-temperature overpressure failure. The NRC recognizes that licensees may protection. Stroke testing of the PORVs should not choose to replace existing PORV and PORV block be performed during power operation.The staff has valves with more reliable designs as they are made j
concluded that stroke testing during power opera-available by valve manufacturers in the future. The tion,in the words of the ASME Code,"is not practi-use of more reliable valves should result in less fre-cal" (reference ASME Section XI, Paragraph quent corrective maintenance and can result in IWV-3412) because of the potential for a PORV to longer inservice testing intervals as delineated in stick open during the stroke test. Additionally, the Section XIof the ASME Iloilerand Pressure Vessel PORV block valves should be included in the licen-Code.
'In the 1988 Addendum to ASME 5cetion XI. the content of Subsection !WV is replaced with a reference to Part 10 of AShtE/ ANSI OM-1987,
- Inservice Testing of Valves in Ught-Water Reactor Power Plants."
)
In addition. it is anticipated that the following proposed standard applicable to currently used PORVs will be included in OM-1987 and may ultimately be referenced in Subsection IWV of ASME Section XI OM-13. Requirements for Periodic Periormance Testing and Monitoring of Power Operated Relief Valve Assemblics I
k 5.2 Potential Improvements to PORVS ponents of the RCPil and systems that perform a safety-aHCl BIOck Valves related function as defined in Regulatory Ouide 1.26 (Ref. 4). While PORVs were originally provided in PWRs The classification of PORVs and block valves should be for plant operational flexibility and for limiting the num.
consistent with the system used for classifying other com-ber of challenges to the pressurizer safety valves, the NUREO-1316 12
i-operation of PORVs for overpressure protection of the such as Vogtle 1 and 2, Millstone 3 Callaway, and Wolf RCPil was not considered to be a safety-related function.
Creek, meet these requirements.
However, since PORVs are relied upon in many PWR plants to mitigate certain design basis accidents or to pet.
For operating PWR plants and construction permit hold.
form any other safety related function that may be identi.
ers, there are a number of potential improvements to fied, the staff finds that it is appropriate to reconsider the PORVs and block valves (short of upgrading to fully classification of PORVs and the associated block valves.
safety grade hardware) that can increase the reliability of these components and provide assurance that they will For future PWR plants when PORVs and the associated function as required when called upon to perform a block valves are used for any of the safety functions dis-safety-related function. It is anticipated that the reliability cussed in Section 2.1 of this report, these components of PORVs and block valves can be increased by imple-should be classified as safety related and a minimum of menting the improvements presented in Table 5.1.
two PORVs and block valves installed. Certain recently licensed plants and plants currently under active The estimated utility present value for implementation of construction that have solenoid pilot opentted PORVs,'
the improvements to PORVs and block valves for items 1, 2, and 3 of Table 5.1 are presented in Table 5.2.These present value estimates are in constant dollars using a
' As of the date of this report,llellefonte 1 and 2 and WNP-1 are not real discount rate of 5 pcreent for a period of 30 years and comidered to be plants under active construction.
are for a plant with two PORVs and two block valves.
Table 5.2 Present value ofimprovements to PORVs and block valves.
Utility Cost /
Item Description Reactor 1
Includo PORVs and block valves in an operational quality assurance program that is in compliance with 10 CFR Part 50, Appendix U (onctime cost of $8,000 plus a recurring cost of $200 per year for 30 years).
$11,100 2
Implement a maintenance / refurbishment program for PORVs and block valves (recurring cost of $5,300 per year for 30 years).
$81,600 3
Testing in accordance with Subsection IWV of Section XI of the ASME Code for PORVs and block valves (recurring cost of $1,200 per year for 30 years).
$18,500 4
Revision of technical specifications for PORVs and block valves (onctime cost of $16,000).
$16,000 5
Test block valves in accordance with NRC Generic Letter 89-10.
Total utility present v:ttue for implementation of items 1,2,3, and 4
$127,200
- See Value-impact Analysis as reported in NURt!G/CR-$140 (Ref.16) for resolution of Generie issue llE6.t. "In Situ Testing of Valves," as reported in NRC Genene letter 89-10 (Ref. 5).
5.3 Safety Benefits decay heat from the reactor core was explored in some de-tail. Studies performed under USI A-45 in general sup As discussed in Section 3.2, the BNL study performed P.0ft the concept of feed and bleed, but do point out that specifically in support of GI-70 showed only a small po-timmg is a critical parameter in establishing whether or tential decrease in core melt probability. This was in part not primary feed and blecd can successfully remove decay because by staff direction the study did not include consid-heat. However, discussions with personnel at various cration of feed and bleed capability.
PWR plants revealed that most PWR utilities claim that feed and bleed is a viable decay heat removal method and in the course of the resolution of Unresolved Safety Issue (USI) A-45, as reported in NUREG/CR-5230 (Ref. 6),
The effect of the feed and bleed process upor, the prob-the use of feed and bleed cooling on the primary system as ability of core melt p(cm) was examined in NUREG/
an alternative, essentially last resort, measure to remove CR-5230 (Ref. 6) as a sensitivity issue, and in constructing 13 NUREG-1316
L i,
accident sequence event trees credit was given for this ca-percent. The results are plant specific and show that feed pability. The core melt probabilities for internal events and blecd reduces core melt frequency by 4.8E-5 to only, with and without feed and bleed, were calculated, 1.15E-3. These resu!!s are presented in Table 5.3. Even if and the report indicates that feed and bleed capability re-feed and bleed only reduces the core melt frequency by a duces the estimated core melt probability for internal fraction of the improvements shown above, it is a substan-events try a significant amount-on the order of 25 to 90 tial reduction in public risk.
Table 5.3 Core melt probability with and without feed and bleed from case studies in USI A-45 (internal events only) with recovery, p(cm) pr Reactor Year plem) per Reactor Year p(cm)
Plant Without Feed and lilced With feed and Bleed per Reactor War A
1.87E-4 1.39E-4 4.8E-5 B
LOOU-4 7.1E-5 2.9E-5 C
4.8 E-5 1.4 E-5 3.4 E-5 D
1.23E-3 8.8E-5 1.14E-3 ~
The analyses performed in NUREG/CR-5230 clearly 5.4 Outage Avoidance Cost show that a feed and bleed capability can have a signifi-cant effect on the probability of core melt, however, it is As noted on page 3-32 of Reference 17, PWR capacity noted that decisions to feed and bleed must be made early losses, that is, outage time due to RCS relief valve prob-in an accident progression for it to be successful.
lems, have remained relatively unchanged over theyears.
The PIoP 8ed actions to PORVs and block valves that are As examps: m k yean N plant capady presented in Table 5.1 of th.is report enhance (but do not losses from the RCS safety / relief salves were 0.15 per.
ensure) feed and bleed because:
cent, and for the years 1980-1982 the RCS safety / relief valve capacity losses were 0,18 percent (page 4-57 of Ref.
18). Approximately 75-85 percent of the RCS safety /
1, inclusion within an operational quality assurance relic! valve capacity losses are attributed to PORVs and program that is in compliance with 10 CFR Part 50, block valves (Refs.17 and 18).
Appendix H, of an improved PORV maintenance /
refurbishment program and additional surveillance Based on the above EPRI historical data, PORVs and testing provide better assurance that PORVs will block valves have been responsible for, on an average, open or close when called upon.
PWR capacity losses (CL) of approximately:
2.
Currently, certain plants operate with the block valves closed. The technical specifications for these ne EPRI observation that the capacity losses attributed plants require that power be racked out at a valve to these valves have remained relatively constant over the motor control center, making it unlikely that feed years infers that little,if any, improvement has been made and bleed could be initiated in a timely manner.The in the maintenance and quality assurance procedures for proposed revised technical specifications require these valves. EPRI in t heir limiting factors study on valves those plants that run with the block valves closed to (page 5-4 of Ref.19) concluded that, during a plant life-maintain electric power to the block valves so they time, valve performances could be significantly improved can be readily opened from the control room.
by proper operation and maintenance. In this regard, EPRI stated: "Probably no other single recommendation 3.
Placing the block valves within the scope of NRC will impr ve valve availability as much as proper care and Generic Letter 89-10 (Ref. 5) would provide in-respect dunng plant lifetime.
creased assurance that the block valves would open against system differential pressure to permit init a-Considering the above and other EPRI recommendations tion of feed and bleed.
(see Sections 5.6 and 5.7 of Ref.19), it is estimated that placing PORVs and block valves on the quality assurance program list of critical valves, improvements in PORV Based on the above, the staff concludes that the proposed and bhick valve maintenance programs and quality assur-actions to PORVs and bhrk valves provide a substantial ance procedures, inservice testing in accordance with Sec-increase in the overall protection of the public health and tion XI of the ASME Code and additional testing for safety.
PORV block valves requested in NRC Generic Letter NUREG-1316 14 k.
89-10, and upgrading the technical specifications on
(,
FINDINGS these valves as discussed in Table 5.1 of this report could result in approximately a 75 percent reduction in plant ca-pacity losses attributed to PORVs and block valves.
For future PWR plants, the staff concludes that a mini-mum of two PORVs and blotk valves and associated con-Considering an average replacement power cost (RPC)of trols for these components should be provided. These
$500,000 per day, the above improvements in PORVs and components should be identified as safety related if re-block valves could yield a yearly per plant outage avoid-quired to perform any of the safety-related functions dis-ance cost (savmgs)of:
cussed in Section 2 1 of this report or to perform any other safety related function that may be identified in the fu-OAC - (0.75)(CL)(365)(RPC) -
ture and should therefore be constructed to safety grade
$165,000 per reactor year.
standards.'this would include redundant and diverse con-trol systems, designed to Seismic Category I requirements E
ic t on surve I e require ents inct s ns rvic d of 3 ye testing requirements; and inclusion within the scope of a quality assurance program that is in compliance with 10 CFR Part 50, Appendix IL The safety grade designation OAC (Present value) - $2,541,000 per reactor would include those improvements that were imposed subsequent to the TMI-2 accident, such as requirements to be powered from Class III buses and to provide valve 5.5 Cost /Henefit Comparison position indication in the control room.The staff also con-cludes that item 4 in Table 5.1 of this report should be im-plemented where possible in future PWR plants and
'lhe calculated numerical values used in this cost / benefit strongly encourages operating reactor owners and con-comparison are used only as an aid to the decisionmaking struction permit holders to evaluate the benefits of re-
. process and are not intended to be used as the final decis-placing existing PORVs and block valves with more reli-ionmaking criterion on this issue. The values are, there-able designs, fore, considered a supplementary tool to provide addi-tional insight in an overall evaluation of this issue (Ref. 20).
For operating plants and construction permit holders, the staff concludes it is not cost effective to replace (back-
,the present value associated with the.mipravements to f t) existing non-safety-grade PORVs and block valves PORVs and block valves for items 1,2,3, and 4 of Table (and associated control systems) with PORVs and block 5.2 of this report are estimated to be $127,200 for a plant valves th'it are safety grade for the sole purpose of making with two MRVs and two block valves, them safety grade when they have been determined to perform any of the safety related functions discussed in The overall cost / benefit (OC3) present value per reactor Section 2.1 of this report or to perform any other safety-related function that may be identified in the future. Sub-
[3, sequent to the TMI-2 accident, a number of improve-ments were required of PORVs, such as requirements to OC3(Present Value) -' Outage Avo. dance Cost be powered from Class 111 buses and to have valve posi-i (OAC) - Implementation Cost (le)of items 1, tion indication in the control room.Therefore, additional 2,3, and 4, Table 5.2 improvements that would result from upgrading PORVs -
to fully safety-grade status are considered to be of mar-ginal benefit. For operating P}VR plants and construction OC3 (Present Value) - OAC $2,541,000-Ic permit holders, the greatest tmmediate benefits can be
! $127,200 - $2,413,800 per reactor derived from implementing items 1 through 3 in Table 5.1 of this report. The staff is proposing that these require.
The projected costs to the NRC upon implementation of ments be imposed to increase the reliability of PORVs -
items 1,2, and 3 of Table 5.1 of this report are as follows.
and block valves to provide assurance they will function as Implementation of items 1(b) and 2 are recurring costs required. Items 1 through 3 in Table 5.1 of this report, that result from inspections and evaluations covered by which do not require hardware changes, can be imple-existing NRC monitoring programs and are not charge-mented within the scope of current licensing criteria and able to the operation cost of this issue. lmplementation of coordinated with the Technical Specifications improve-items 1(a) and 3 are one thac costs of $15,000 per plant.
ment Program.
15 NURF.G-1316
REFERENCES
- 10. Electric Power Research Institute (EPRI), "EPRI PWR Safety and Relief Valve Test Program-Safety 1,
E. D. Throm, " Regulatory Analysis for the Resolu-and Relief Valve Test Report," EPRI NP2628-SR, tion of Generic Issue 94, 'Arlditional low-December 1982.
Temperature Overpressure Protection for Light-Water Reactors'," NUREG-1326, December 1989.
- 11. L hlarsh and C Liang,"Evaluationof the Need for a Rapid Depreskrization Capability for CE Plants,"
2.
Oak Ridge National laboratory, " Operating Experi-NUREG-1044, December '1984.
ence Review of Failures of Power Operated Relief Valves and Illock Valves in Nuclear Power Plants,*
- 12. NRC Information Notice No. 89-32, " Surveillance NUREG/CR-4692, orb '/NOAC-233, October Testing of Imw Temperature Overpressure Protec-1987.
tion Systems," dated hiarch 23,1989.
3.
C. Hsu et al., " Estimation of Risk Reduction from
- 13. Power Authority of the State of New York, Consoll.
Improved PORV Reliability in PWRs," Brookhaven dated Edison Company of New York, Inc., "Indmn National laboratory, NUREG/CR-4999, llNL-Point Probabilistic Safety Study," Amendment 2, De-NUREG-52101, Final Report, hiarch 1988, cember 1983.
4.
NRC, Regulatory Guide 1.26, Revision 3, " Quality
Probabilistic Risk Assessment of Oconce Unit 3,"
Steam, and Radioactive-Waste-Containing Compo-NSAC-60, Vol.1, p. 3-40, June 1984.
nents of Nuclear Power Plants," For Comment, Feb-ruary $7&
- 15. NRC, Regulatory Guide 1.29, Revision 3,"Scismic Design Classification," September 1978.
5.
NRC Generic letter 89-10, " Safety Related hiotor Operated Valve Testmg and Surveillance," dated
- 16. J. C. Higgins et al., "Value impact Analysis for Ex-June 28,1989.
tension of NRC llulletin 85-03 to Cover All Safety-Related h10Vs," Brookhaven National Laboratory, NUREG/CR-5140, UNL-NUREG-52145, July 6.
D. hi. Ericson, Jr., et al., " Shutdown Decay Heat 1
1988' Removal Analysis-Plant Case Studies and Special Issues: Summary Report," Sandia National 12bora-tories, NUREG/CR-5230, SAND 88-2375, April
- 17. EPRI," Nuclear Unit Operating Experience-1977
- 1989, and 1979 Update," EPRI NP2092, October 1981.
- 18. EPRI, "Nucicar Unit Operating E.perience: 1980 7,
NRC, " Standard Review Plan for the Review of Through 1982 Update," EPRI NP3480, Aprii 1984.
Safety Analysis Reports for Nuclear Power Plants, LWR Edition," NUREG-0800, July 1981.
- 19. EPRI,"LimitingFactor Analysisof High Availability Nuclear Plants, Volume 3: Supplement Report, Lim-8.
NRC hiemorandum to W. hiinners from B. Sheron, iting Valves Study," EPRI NPI139, August 1979.
" Proposed Generic Issue on PORV and Block Valve Reliability," dated June 27,1983.
- 20. S. W. Heaberlin et al., "A Handbook for Value-Impact Assessment," llattelle Pacific Northwest 9.
NRC, " Clarification of Thil Action Plan Require-Laboratories, NUREG/CR-3568, PNL-4646, ments," NUREG-0737, November 1980.
December 1983.
o NUREG-1316 16
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NRC Fanu 335 U.S. NUCLE AR RE OULATORY COMMIS$10N
- 1. R E POR 1 NUMBE R A
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m.m BIBUOGRAPHIC DATA' SHEET neoinmi,com,<ib NUREG-1316 4
- 2. llTLE AND SUS 14T LE l Technical Findings-and Regulatory Analysis Related to Generic
,~ ' Issue 70: Evaluation of Power-0perated Relief Valve and Block E
3.
04Tc REPORT PUeti$sto
' Valve Reliability in PWR Nuclear Power Plants l
December 1989
- 6. AUTHOR (S) 6 TYPE OF REf' ORT R.-Kirkwood Tonical
- 7. PE R 400 COVE RE D tenc& sere 04:ess NA
- 8. P.E.A.FORMtN,n,,ORGANIZ ATiON - N AVE AND ADon ESS tor Nnc. prere coveen. orterr or anaan. v.s Necorer parasatory commeuen. ome nwkne emoren. tr e G
.~4 saena Division of Safety Issue Resolution Office'of Nuclear Regulatory Research U. S.- Nuclear Regulatory Commission Washington, D.C.20555
- 9. SPONSORING ORGANIZ ATION - N AME AND ADOf.ESS (st her, arer %=, m ean,e ; er teneractor.provede Nac 0,vesdon. Onere er Rev oa. d.1 Svedrer Assu4 tory Co**dadea.
- endensilsnt ad*esL)
<i Same as above-
- 10. SUPPLEMENTARY NOTES
.None 11, ABSTR ACT (200=orm os Arsst i
l'
.This report summarizes work performed by the Nuclear Regulatory Commission staff to resolve Generic Issue 70, " Technical Findings and Regulatory Analysis Related to Generic Issue'70 - Power-0perated Relief Valve and Block Valve Reliability." The l
report evaluates the reliability of PORVs and block valves and their safety significance in PWR. nuclear power plants.
The report identifies those safety-related functions i
-that may be performed by PORVs and describes ways in which PORVs and block valves may.
be improved.
This report also presents the regulatory analysis for Generic Issue 70.
j i
- 12. KE Y WOROS/DESCR!PTORS (ter were erparenes thes we# ass /ss seswcmers an tocaems the troort.J t3. AvAlLAasuiY 5I AILMENI Generic Issue 70, -Power-0perated Relief Valves, Block Valves, Unlimited Valve Reliability, Safety Functions of PORVs, Steam Gerarator Tube Rupture, Low-Temperature Overpressure Protection, Plant Cooldown in Compliance with BTP RSB 5-1, Events Beyond Design Basis, Technical Unclassi fied Findings ario "latory Analysis GI 70, PRA
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