ML20206U733
| ML20206U733 | |
| Person / Time | |
|---|---|
| Issue date: | 12/21/1994 |
| From: | Holahan G Office of Nuclear Reactor Regulation |
| To: | Thadani A Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20206U672 | List: |
| References | |
| FOIA-99-82 TAC-M85648, NUDOCS 9902170325 | |
| Download: ML20206U733 (250) | |
Text
_
p UNf783 STATES 3
)W NUCLEAR REOULATORY coa: MISSION waemmorow. o.o. seenween j
- ,= c e m a r 21 1994 MEMORANDUM TO
Ashok C. Thadant, Associate Director for. Inspection and Technical Assessment df fRON:
Cary M. Holahan, Director
/. f g Division of Systems Safety and Arfa' lysis
SUBJECT:
REVIEW OF EXISTING ENVIROPMENTAL QUALIFICATION PROGRAM REQUIREMENTS (EQ-TAP ACTION ITEM 3.d) (TAC. M85648) i As discussed in the staff's Environmental Qualification Task Action Plan (EQ-TAP) of June 16, 1993, we are performing a programmatte review of EQ for electrical equipment. Our efforts in this regard are spectfically defined i
under Action Ites 3 of the EQ-TAP, which includes the following elements:
1 3.a Review License Renewal Background Information 3.b Review Fire Protection Reassessment Report I
3.c Elicit opinions from Others (Regions, EQ Experts) d 3.d Review Existing EQ Program Requirements 3.e Review NRC Audit / Inspection Practices 3.f Review Licensee laplementation Practices i
j 3.g Finalize Review Results I
Our objective in completing items 3.a through 3.f (above) is to identify potential EQ 1ssues and concerns that may deserve further staff consideration.
i j
lt is laportant to recognite that this part of our programmatic review is not intended to resolve or to otherwise adetss any of the EQ issues that are identified, After items 3.a through 3.f of the EQ-TAP have been completed, all of the Eri issues will be consolidated and specifically addressed in the staff's final report under item 3.g, " Finalize Review Results," which will l
include recommendations as appropriate. We hope to issue our final report by December 30, 1994.
We have now completed our review associated with Action Item 3.d of the i
EQ-TAP, " Review of Existing EQ Program Requirements," and our evaluation is i
enclosed for your infomation. The staff's efforts consisted of a fairly i
comprehensive review of information concerning EQ requiremer.ts, focusing R
heavily on research and development aspects.
As a result of our review, many issues were identified for further consideration within the overall context of 1
the EQ Task Action Plan. Each issue falls into one of the following six categories:
(a) EQ Methodology and Practices - General Issues; (b) EQ Methodology and Practices - Specific Issues; (c) EQ Program laplementation g
l 9902170325 990211 i
i PDR FOIA g- /, /7. iC N KO,HN99-82 PDR
,y a
P 4ldb
- i.j e !.} f",
.e i
t A. *nadant 7
With regard to the 'superheat effects" of a MSLB, the staff's resolution of TAP A-21 failed to consider single failure const(erations.
It is not clear to what extent single failure need not be considered such thet
" local effects
- can be excluded.
As stated in Regulatory Guide 1.89, the purpose of environmental qualification is to avoid "cosmon-cause* failures. Given this, it is not clear why it is necessary to qualify equipment to protect against single fallores.
The SRP suggests that NUREG-0588, RG 1.89, and IEEE 323 may be applicable for qualification of mechanical equipment, but specific guidance has not been provided in this regard.
These issues, along with any other potential issues that are identified during the course of our programmatic review relative to EQ (EQ-TAP Action Items 3.a through 3.f), will be assembled and addressed in our final report (EQ-TAP Action item 3.g).
Please contact me if you should have any questions regarding the enclosed evaluation or the staff's ongoing review efforts relative to EQ.
Attachment:
Review of Existing EQ Program Requirements (EQ-TAP Action item 3.d)
DISTRIBUTION:
Central Flie JWerniel SPL8 EQ File
$Newberry WRussell PShenanski FMiraglia AEl-Bassioni MVirgilio L01shan CMcCracken EButcher GHubbard LShao, RES l
JTatum JCraig, RES 1
CGratton MVagins, RES j
ADusper JYora, RES j
CBerlinger SAggarwal, RES l
l Cat (c'urrences:
.\\
f)
SPLB:DSSA DO:DSSA A
)
JTatum MVirgilio olahan d
9// 9/94 9/(g/94 g/94
[G:\\EQ\\ COMPLETE \\ TAP 3D\\*.*]
i
f A. Thadant Issues; (d) General Technical Issues; (e) Specific Technical Issues; and (f) Other Program / Policy issues.
The specific issues that were identified during this review are listed below for your information. However, it should be emphasized that these are " potential" issues, some of which may be dismissed under closer scrutiny.
[0 Methodoloov and Practices - General Issues:
Industry representatives have presented rational and compelling arguments strongly suggesting that the methodology and practices for estabitshing and maintaining EQ should be reassessed.
Offferent EQ standards were imposed (i.e., D0R Guidelines, NUREG-0588 Category I, and NUREG-0588 Category II) without supporting technical justification as to: (a) why more rigorous standards were warranted, i
and (b) why " progressively less strict standards" were adequate for the older plants.
Given the Regulatory Requirements Review Cosmittee and the NRC staff view that backfitting the IEEE 323-74 requirements would provide "...a small, unquantifiable inerease in the level of assurance that equipment is qualified as compared to the significant costs that would be i
involved...," IEEE 323-74 may not be warranted or sufficiently justified as a necessary qualification standard for power reactors, regardless of when the Construction Permit Safety Evaluation Report was issued.
The current version of IEEE 323 may be better suited for demonstrating EQ than the 1974 version since much more infomation and experience are available now than there was when IEEE 323-74 was endorsed by the staff.
i Current " state of the art capabilities" may not be sufficiently i
developed to support existing EQ requirements, such as detemination of
{
a " qualified life."
l
, The SRP (Section 3.11) is very much out of date and needs to be made current (e.g., the Environmental Qualification Branch is Itsted as the lead review group; there is no reference to the EQ rule; a " central file" is referred to contrary to what was ultimately required by 10 CFR 50.49; and RG 1.89 and IEEE 323-74 are not recognized as the 1
appropriate staff guidance documents for satisfying EQ requirements).
IEEE Standards 381, 535, 627, 649, and 650 (and perhaps others) pertaining to EQ have not been endorsed by the NRC.
PRA studies indicate that EQ Master Lists may need to be updated to include additional equipment.
1 E0 Methods. Guidance and Review Criteria - Specific Isutu:
The need for rigorous qualification of equipment located outside containment may not be warr&nted.
The need and/or ability to establish post-accident qualification beyond a two to four week period is questionable.
i
~
A:Ihi$lRI
- }:
i in lieu of attempting to define a 'quallfled life," It say be more e
appropriate to develop methods for addressing and/or monitoring in-service degradation.
Given more realistic assumptions for the release fractions, the timing of the release, the chemical form of the release, and accident mitigation effects resulting from equipment response, an tausediate and large source term (TID) say be overly conservative and inappropriate.
A general exemption for radiation qualification testing of equipment exposed to low-level radiation may be well suited for fQ purposes under certain defined circumstances.
Better definition is needed for which instruments are required to be qualified.
'kes0HE to th0 CQntrary' for not Ypgrading ftplacement equipeent to the requirements stated by 10 CFR 50.49 appear to be without merit and should be justified.
i Purchase specification requirements for replacement parts have not been addressed relative to EQ.
Periodic maintenance and surveillance requirements necessary to maintain EQ have not been defined.
Methods have not been established to ensure that installation and maintenance practices do not jeopardize equipment qualification.
Margin requirements for demonstrating EQ (e.g., one hour minimum operating time, thermal aging, et:.) may be too severe and without sufficient justification; overall margin requirements need to be better l
defined with supporting technical justification.
It may be appropriate to perform " aging
- of equipmert in the same functional state as it is used in the plant (i.e., energized or de-l l
energized).
The " double peak' requirement (i.e., exposure to two cycles of maximum temperature and pressure) is not representative of design basis conditions and say be too severe.
The need to consider oxygen diffusion effects on aging during normal plant operation may not be warranteo given realistic assumptions.
Humidity itsting capabilities may not be sufficient to assure equipment qualification.
In areas outside containment that are exposed to long-term recirculation, equipment was not required to be qualified for chemical sprayt. However, thia equipment may be sub. ject to actuation of fire suppression systems.
__ ~
- - - - - ~ ~ - -'
i A. Thadant,
1 The post-mandrel bend test may be too severe and perhaps unwarranted for EQ purposes.
Extrapolation of data, if adequately justified, may be well suited for j
EQ applications.
It may not always be appropriate to require test data in order to establish EQ, given the other qualification methods allowed by IEEE 323-74 (i.e., operating experience, analysis, and combination of methods).
i Given the complexities and uncertainties involved, laboratory accreditation may be necessary to assure that EQ testing is properly performed.
The generic temperature profile that was allowed by the DOR Guidelines i
and NUREG-0588 for equipment qualification (i.e., T, for PWRs and u
1 T, + 20*F for BWRs) was not fully justified.
u Comparison of the calculated MSLB surface temperature to the LOCA bulk i
j temperature may not assure that the equipment will survive the MSLB j
environment.
4 E0 Procram Imolementation Ishues:
i It is not clear to what extent the various clarifications and staff i*
positions that were stated in Generic Letters, IE Bulletins, Appendix 8 of NUREG-0737, etc., were fully implemented, and which ones are j
currently appitcable since they are not specifically referred to by 10 CFR 50.49.
f l
It is not clear y ;1ch parts of the DOR Guidelines and NUREG-0588 the staff considered to be " optional," and consequently, the minimum standards that were found to be acceptable to the staff are not well l
defined.
It is not clear to what extent all test failures (for all attempted tests) were required to be documented, evaluated, and saved as part of the equipment qualification record (i.e., to what extent was EQ determined based on
- selective" information).
It is not clear to what extent the limiting undervoltage and underfrequency conditions were assumed for qualification testing f.or all plants (i.e., seismic vs. other hostile environmental conditions),
i especially for plants subject to the DOR Guidelines.
Emergency shutdown systems "...used to bring the plant to a cold shutdown condition following accidents wisich do not result in a breach of the reactor coolant pressure boundary together with a rapid depressurization of the reactor coolant system" were required to be i
qualified by plants subject to the DOR Guidelines.
It is not clear, however, that this was the case for NUREG-0588 plants. Also, since
A. Thadant i h
qualification of cold shutdown equipment was not required by the EQ rule, it is not clear to what extent this requirement is currently valid.
Non-safety-related instruments that could impact the operability of safety-related instruments were not initially included in the scope of equipment that was required to be qualified, and implementation of this requirement may not be uniform among all plants.
RG 1.97 instruments were not addressed in the initial qualification 4
requirements and it is not clear to what extent (and to what criteria) instruments were required to be qualified.
General Technical Issues:
Qualification of equipment other than electrical, and equipment qualification for other conditions (e.g., mechanical and flow-induced vibration, seismic effects, dynamic effects, etc.) have not been i
reviewed within the EQ framework.
Research is only just beginning to assess the adequacy of EQ for RG 1.97 i
functions. The areas of primary concern include reactor coolant level instrumentation, core exit thermocouples, containment area radiation monitors, halogen and particulate sampling, and coolant activity measurements.
j Very little EQ research has been conducted on pressure switches, RTDs,
=
i pressure transmitters, and valve operators.
i Failure of other electrical components such as electrical penetrations and connector assemblies may be more important than the failure of electrical cables.
4 Specific measures may be warranted to minimize the impact of dust on l
equipment qualification.
Soecific Technical Issues:
Environmental conditions for accidents other than for LOCA (such as for MSI,B) were not defined for at least 65 power reactors. The staff failed to recognize this factor in its resolution of Task Action Plan
[
Item A-21.
The staff's " final position" regarding the velocity profile in containment during blow down was pending completion of Task A-21.
j However, the staff's resolution of Task A-21 was incomplete and this issue may need to be revisited.
Qualification of equipment seals and vapor barriers on plants, especially those that are subject to the D0R Guidelines and NUREG-0588, may not be sufficient.
Hydrogen burn scenarios may result in conditions that exceed the EQ t
envelope.
mis i u.. -.. _
l A. Thadant,
Temperature and radiation stratification may result in conditions that exceed the EQ envelope.
Terminal blocks for use inside containment may not be sufficiently qualified.
Qualification of solenoid valves for some applications may not be sufficient.
Bonded jackets on electric power and/or control cables may not be sufficiently qualif ted.
ihe color of insulation material may influence the rate of degradation of the insulation material.
i 1
Kapton vulnerabilitics may need to be addressed for nuclear applications.
Epoxy compound used for potting electrical penetrations may not be qualifted to the temperature conditions that are experienced post-LOCA
]!
and/or during a MSLB.
Fire retardant coatings (and other fire protection features) may not be accounted for within the existing EQ framework.
Other procram/ Policy issues:
Given the state of the art that was in existence at the time IEEE 323-74 was developed and the limitations that existed, it would seem that EQ program requirements may have been misdirected (especially with regard to the required determination of " qualified life" and the absence of surveillance requirements for obtaining advance warning of significant i
l degradation).
A lot of research has been completed and much more experience has been obtained in the area of EQ since 10 CFR 50.49 was issued, but focused NRC programs and initiatives apparently do not exist to continually monitor progress in this area and to make use of this information for restructuring, directing, a u roving EQ program reautrements.
NRC research activities h.. not been entirely successful in resolving the " age-old" EQ issues that were initially identified i
of NUREG/CR-4301 in Appendix L for specific examples). (see the summary I
Additional reporting requirements for EQ problems that occur i.e.,
problems that occur during qualification testing as well as pr(oblems that occur during plant operation) may be warranted given the uncertainties associated with establishing and maintaining equipment qualification.
It is not clear why it was not necessary to impose the single failure criteria for qualification of cold shutdown equipment (i.e., IEB 79-01, Supplement 3, only required one train of cold shutdown equipment to be qualified).
_I____------------------------~-~
~ - - ' ~ ~ ~ ~
~
--~ ~ ~ ^
~ '-~
~
A. Thadant.
With regard to the "superheat effects" of a MSLB, the staff's resolution of TAP A-21 failed to consider single failure considerations.
It is not clear to what extent single failure need not be considered such that
' local effects" can be excluded.
l As stated in Regulatory Guide 1.89, the purpose of environmental qualification is to avoid " common-cause" failures. Given this, it is j
not clear why it is necessary to qualify equipment to protect against single failures.
i The SRP suggests that NUREG-0588, RG 1.89, and IEEE 323 may be applicable for qualification of mechanical equipment, but specific guidance has not been provided in this regard.
These issues, along with any other potential issues that are identified during
)
the course of our programmatic review relative to EQ (EQ-TAP Action Items 3.a through 3.f), will be assembled and addressed in our final report (EQ-TAP Action item 3 9).
Please contact me if you should have any questions i
regarding the enclosed evaluation or the staff's ongoing review efforts i
relative to EQ.
Attachment:
Review of Existing EQ Program Requirements (EQ-TAP Action Item 3.d) l i
l l
1 i
j t
i f
s A. Thadant With regard to the "superheat ef.%:ts" of a MSLB, the staff's resolution of TAP A-21 failed to consider single failure considerations.
It is not clear to what extent single failure need not be considered such that
- 1ocal effects" can be excluded.
As stated in Regulatory Guide 1.89, the purpose of environmental qualification is to avoid "comon-cause" failures. Given this, it is not clear why it is necessary to qualify equipment to protect against single failures.
The SRP suggests that NUREG-0588, RG 1.89, and IEEE 323 may be applicable for qualification of mechanical equipment, but specific guidance has not been provided in this regard.
These issues, along with any other potential issues that are identified during the course of our programatic review relative to EQ (EQ-TAP Action Items 3.a through 3.f), will be assembled and addressed in our final report (EQ-TAP Action Item 3.g).
Please contact me if you should have any questions regarding the enclosed evaluation or the staff's ongoing review efforts relative to EQ.
Attachment:
Review of Existing EQ Program Requirements (EQ-TAP Action item 3.d)
DISTRIBUTION:
Central Flie JWermiel SPLB EQ File SNewberry WRussell PShemanski FHiraglia AEl-Bassioni MVirgilio L01shan CMcCracken EButcher G}lubbard LShao, RES JTatum JCraig, RES CGratton MVagins, RES ADumer JVora, RES CBerlinger SAggarval, RES CefIurrerices:
,,4
[ l SPLB:DSSA 00:DSSA A
JTatum MVirgilio olahan 9// 9/94 9/f\\/94
/L\\/94
[G:\\EQ\\ COMPLETE \\ TAP 30\\*.*]
w-
..c 4
e 1
Attachment REVIEW OF EXISTING ENVIRONMENTAL QUALIFICATION PROGRAM REQUIREMENTS j
{
1.0 INTRODUCTION
4 As discussed in the Environmental Qualification Task Action Plan (EQ-TAP) of June 16, 1993, the staff is performing a reassessment of the NRC environmental qualification (EQ) requirements for electrical equipment. Action Item 3 of 4
the EQ-TAP lists those actions that pertain to the programmatic review of EQ, which include:
3.a Review License Renewal Background Information 3.b Review Fire Protection Reassessment Report 3.c Elicit Opinions from Others (Regions, EQ Experts) 3.d Review Existing EQ Program Requirements 3.e Review NRC Audit / Inspection Practices I
3.f Review Licensee Implementation Practices 3.g Finalize Review Results i
i This particular evaluation is intended to address EQ-TAP Action Item 3.d,
" Review Existing EQ Program Requirements.* The specific objective of this review is to identify potential EQ issues and concerns by reviewing the existing environmental qualification program requirements and related information. Ultimately, all of the issues and concerns that are identified during the EQ progransatic review will be consolidated and discussed in the final report (EQ-TAP Action Item 3.g).
Therefore, this evaluation does not include specific recommendations for further staff actions.
2.0 REVIEW METH000 LOGY AND REPORT FORMAT In order to accomplish the objective of this evaluation (i.e., identify 4
potential issues and concerns associated with existing EQ requirements), a substantial amount of information was collected and reviewed. For example, existing EQ requirements were reviewed along with SECY Papers, research i
reports, generic communications, and related memoranda. The staff's review of background information was fairly comprehensive in order to fully appreciate how EQ requirements were developed over time and also to better understand what the relevant issues were and how they were addressed.
Although this review was both resource and time limited, the staff believes that sufficient information was reviewed in order to identify any programatic EQ issues that 4
may exist.
In completing this review, the staff attempted to maintain a
" balanced perspective" such that preconceived notions and biases were continually challenged throughout the review process and both regulatory and industry views were considered in an objective and equal manner.
contained in Section 3 through Section 7 of this eva$uation.The various ele Supporting information is contained in Appendix C through Appendix K, and potential issues (as well as some noteworthy coments) are highlighted in bold print z
within these appendices. Potential issues from the appendices are sumarized within the corresponding sections of this report, and Section 8 contains an overall sumary of the issues that were identified. Finally, Section 9 Ilsts the references that were used during this review effort and Appendix A is 'a l
listing of the abbreviations that are used throughout this report. The I
specific Sections and Appendices included in this report are listed here for reference purposes:
Section Inig 1.0 Introduction 2.0 Review Methodology and Report format 3.0 Background Information 4.0 NRC Requirements for Environmental Qualification of Electrical Equipment 5.0 NRC Review Criteria. Implementab an Guidance, and Industry Standards Related to EQ 6.0 Generic Cumunications 7.0 Research Studies Related to EQ 8.0 Sumary 9.0 References Accendix Iult A
Abbreviations 8
Chronological Information C
Background Information Related to EQ D
Information Concerning 10 CFR 50.49 E
Standard Review Plan Section 3.11 F
00R Guidelines H
Comparison of EQ Review Criteria X
Generic Comunications L
Research Activ'ities and Studies Related to EQ
3.0 BACKGROUND
INFORMATION
)
Ti.e first industry standard that focused strictly on environmental qualification of electrical equipment used in nuclear power reactors was IEEE 323, 'IEEE Trial Use Standard: General Guide for Qualifying Class !
Electric Equipment for Nuclear Power Generating Stations," which was issued in April 1971. Up until that point, the "high industrial quality
- of equipment was relied upon to provide assurance that electrical equipment would function during postulated accident conditions. Consecuently, 58 of the 70 nuclear power plants initially licensed to operate made no specific reference or comitment to IEEE 323-71. The standard was revised in February 1974 to address additional factors such as aging, margin, t.nd qualified life, and to provide more detailed guidance since the original standard was very brief and not very specific. At the time that IEEE 323-74 was issued the Regulatory Requirements Review Comittee (RRRC) recomended that the revised standard only be applied to future construction permit applications and expressed the 1 _ - _ _ _ _ _ - _ _ - _ -
{
4
'4 view that "...the incremental improvements (of IEEE 323-74] were not significant to safety..." as the reason for not backfitting the standard; no technical basis was cited fc,r this position.
Although the NRC had initiated a Qualification Evaluation Testing Program in 1975 as a preliminary step in addressing EQ for operating reactors, the staff's awareness and focused attention on EQ was accelerated to a large degree by a petition for Comission action that was submitted by the Union of Concerned Scientists (UCS) on Novecber 4, 1977. The UCS expressed concerns over the p alification of electric equipment at operating power reactors based on electrical connector failures that were reported by Sandia National Laboratories.
Sandia was performirg synergistic effects testing on electrical connectors uader the NRC's Qualification Testing Evaluation Program when the failures occurred.
In response to the initial UCS petition and other petitions that followed, such work was done by the NRC staff and by the nuclear industry to demonstrate that electrical equipment was qualified to function during postulated accident conditions.
Licensees of operating reactors were required to demonstrate qualification by responding to the Office of Inspection and Enforcement (IE)
Bulletin 79-01, " Environmental Qualification of Class IE Equipment," dated February 8, 1979.
The staff established criteria for reviewing the Itcensees' IE8 submittals, " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment for Nuclear Power Plants," (comonly referred to i
as the 00R Guidelines) which were applicable to plants that were licensed to operate as of May 23, 1980; plants that were licensed to operate after May 23, 1980, were required to conform to the criteria stated in NUREG-0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," dated December 1979. Appendix 8 is an abbreviated chronology of the staff's resolution of the concerns that were expressed by the UCS.
t As a result of the UCS petition and the work that was done by the NRC staff, the Commission issued a Memorandum ar.d Order (CLI-80-21) on May 23, 1980, expressing the view that IEEE 323-71, by itself, did not provide an adequate standard against which qualification could be judged (contrary to the view that had been expressed years earlier oy the RRRC). The Comission's 1
Memorandum and Order (among other things :
(a established that the D0R Guidelines and HUREG-0588 formed the requ)iremen)ts that must be met by operating reactor licensees and applicants (respectively) for environmental i
qualification of safety-related electric equipment; (b) directed the,taff to prepare additional technical specifications for all operating plants to require EQ documentation as stated in the DOC Guidelines; (c) established a
that, unless there were sound reasons to the contrary, the 1974 standard in NUREG-0588 (i.e., IEEE 323-74) would apply to replacement parts; and (d) i i
recognized that the D0R Culdelines and NUREG-0588 applied progressively less strict standards to the older plants and required that the staff provide i
justification, subject to public coment in the rule making process, for the less rigorous standards.
Environmental qualification requirements along the lines IEEE 323-74 were later codified as 10 CFR 50.49 (a.k.a. the EQ rule), "Environmantal Qualification of Electric Equipment Importcnt to Safety for Nuclear Power Plants." The EQ rule included a grandfather clause such that equipment that 1
J was previously required by the Comission to be qualified per either the DOR Guidelines or NUREG-0588 was not required to be requalified. However, the EQ rule did require al' licensees to qualify replacement equipment in accordance with the provisions of 10 CFR 50.49 unless there were sound reasons not to do so. Although the EQ rule did allow the use of less strict standards for the older plants, no technical basis was provided to justify the existence of separate standards, other than to recognize that substantial costs had already been incurred by the nuclear industry in satisfying the previous EQ upgrade i
requirements that hart tmen in@ sed by the NRC.
i Appendix C it a listing and descriptive sumary of the background information that was reviewed relative to the U".S petition that was filed with the Commission in November 1977, and on the subsequent staff actions that were taken to address the UCS concerns.
The following potential issues, taken from Appendix C, were identified for furtier staf f consideration:
Epoxy compound used for potting eiedrical penetrations may not be qualified to the temperature conditions that are experienced post-LOCA and/or during a MSLB.
Installation and maintenance practices may tend to invalidate EQ.
Exceptions may have been allowed to the DOR Guidelines and to NUREG-0588 during the staff's " screening and review" of operating reactors, making 4
it difficult to know just exactly what the minimum acceptable standard is/was.
Different EQ standards were required without supportina technical justification as to:
(a) why more rigorous studardt % *e warranted, and (b) why " progressively less strict standards" were~ adequate for the older plants.
Hydrogen burn scenarios may result in conditions that exceed the EQ envelope.
Temperature and radiation stratification may result in conditions that exceed the EQ envelope.
t 4.0 NRC REQUIREMENTS FOR ENVIRONMENTAL QUALIFICATION OF ELECTRICAL EQUIPMENT The specific requirements for environmental qualification of safety-related electric equipment are currently stated in 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," dated January 21,1983 (also known as the EQ Rule). Equipment covered by this regulation includes safety-related electric equipment, non-safety-related electric equipment whose failure could adversely impact the performance of safety-related equipment, and certain post-accident monitoring equipment. Licensees are required to maintain a list of the equipment covered by this regulation, including supporting documentation, that demonstrates the basis for equipment qualification.
In particular, licensees are required to address temperature and pressure effects, humidity, chemical effects, radiation, aging, submergence, synergistic effects, a s margins (to account for unquantified uncertainties) in their supporting documentation. The _..
- ~~ ~^ ~ ~
^~~
i k
1 i
regulation requires equipment to be quallfled based on testing of identical cr similar equipment with supporting analysis; experience with identical or similar equipment with supporting analysis; and/or by analysis in combination with type test data.
Qualification of equipment located in a mild environment was not included within the senpe of this regulation. Two important 4
provisions included in 10 CFR 30.49 are:
)
a)
Applicants and licensees were not required to requalify equipment to the requirements of 10 CFR 50.49 f f they were previously required to qualify equipment in accordance with the " Guidelines for Evaluating i
Environmental Qualification of Class IE Electrical Equipment in Operating Reactors,' November 1979 (DOR Guidelines), or NUREG-0588 (For Coment version), ' Interim Staff Position on Environmental Qualification
]
of Safety-Related Electrical Equipment," December 1979.
1 b)
Replacement equipment was required to be qualif ted to the raquirements stated by 10 CFR 50.49 unless there were sound reasons to do otherwise.
Before 10 CFR 50.49 was established, EQ requirements were stipulated by l
generic comunication Modification of 1.icense(see Appendix K) and by a Comission Order for dated October 24, 1980 (see Appendix C); and EQ requirements for protection systems were stated to some extent i
{
by 10 CFR 50.55a, " Codes and Standards," which required licensees (i.e.,
J construction permits issued after January 1,1971) to comply with IEEE Standard 279, " Criteria for Protection Systems for Nuclear Power Generating 4
Stations."
Section 4.4 of IEEE 279-71, pertaining to equipment qualification, i
j 1
required assurance based on test data, that protection system equipment "shall t
meet, on a continuing basis, the performance requirements determined to be necessary for achieving the system requirements." Also, the General Design Criteria (GDC), which were incorporated in the regulations 10 CFR 50, Appendix A) in May 1971, include the following provisions re(lated to equipment j
qualification:
GDC 2. "Desion Bases for Protection Aaainst Natural Phenomena" l
" Structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena such as earthquakes, tornadoes, hurricanes, floods, tsunami, and seiches without loss of capability to perform their safety functions...."
l GDC 4.
- Environmental and Dynamic Effects Desian Bases"
" Structures systems, and components important to safety shall be s
designed to accommodate the effects of and to be compatible with the environmental conditions associated with normal operation, maintenance, testing, and postulated accidents, including loss-of-coolant accidents.
These structures, systems, and components shall be appropriately l
protected against dynamic effects, including the effects of missiles, pipe whipping, and discharging fluids, that may result from equipment i
failures and from events and conditions outside the nuclear power unit.
i However, dynamic effects associated with postulated pipe ruptures in nuclsar power units may be excluded from the design basis when analyses reviewed and approved by the Comission demonstrate that the probability i.
a,
.t +
E-
of fluid system piping rupture is extremely low under conditions
]
consistent with the design basis for the piping."
Note:
It's interesting to note that the use of probability was allowed as a basis for excluding consideration of dynamic effects due to pipe rupture, and perhaps siellar uses of l
l probability may be appropriate in other applications as l
well.
l GDC 23. " Protection System Failure Modes' "The protection system shall be designed to fall into a safe state or into a state demonstrated to be acceptable on some other defined basis if conditions such as disconnection of the system, loss of energy (e.g.,
i i
electric power, instrument air), or postulated adverse environments I
(e.g., extreme heat or cold, fire, pressure, steam, water, and l
radiation) are experienced."
Development of the EQ Rule was very controversial and received a lot of attention and scrutiny from the Cossetssion, NRC staff, industry 1
l representatives, and others. By the time 10 CFR 50.49 was finalized, a tremendous amount of staff and industry resources had already been expended in addressing EQ, and the nuclear industry was very much opposed to a rule that would impose additional EQ backfit requirements.
Appendix D is a itsting and descriptive summary of information that was reviewed by the staff concerning the development of the EQ rule, and potential l
issues that were identified during this review are Itsted within the appendix in bold print. The following potential issues, taken from Appendix 0, were identified for further staff consideration:
Given the complexities and uncertainties involved, laboratory accreditation may be necessary to assure that EQ testing is properly -
performed.
Current " state of the art capabilities' may not be sufficiently developed to support existing EQ requirements.
The " double peak" requirement (i.e., exposure to two cycles of maximum temperature and pressure) during EQ testing may not be warranted.
The more current version IEEE 323 may be more suited for demonstrating EQ than the 1974 version since much more information and experience are available now than there was when IEEE 323-74 was endorsed.
Technical justification was not provided for why uniform critarla for EQ were not applied to all nuclear power plants.
i It may be appropriate to perform aging of equipment in the same e
functional state as it is used in the plant (i.e., energized or de-energized).
The adequacy of EQ for equipment other than electrical, and for other
.s.
m e
l I
j-conditions referred to b effects is questionable.y the GDC such as selseic and dynamic loading t
Reporting requirements for EQ-related problems (i.e., problems j
experienced during qualification testing as well as problems that occur during plant operation) may not be adequate, given the uncertainties j
inherent in the qualification process.
j EQ requirements for cold shutdown equipment are questionable.
EQ requirements for replacement parts should be reassessed.
r It may not be possible to establish a well-defined qualified life given the current state of the art.
I 5.0 NRC REVIEW CRITERIA, IMPLEMENTATION GUIDANCE, AND INDUSTRY STANDARDS j
RELATED TO EQ t
1 t
NRC requirements and review criteria for environmental qualification of 4
electric equipment have evolved as the staff has become more aware of the i
problems associated with EQ and as industry standards in this area have been 5
i developed.
Specific NRC requirements and in nstry standards fcr (Q did not exist for the older nuclear power plants (i.e., those reviewed by the staff prior to about 1967), and these plants were Itcensed simply on the basis that equipment was of "htgh industrial quality." Plants that were reviewed by the i
i staff during the period from around 1967 through 1974 were assessed to varying degrees against the Institute of Electrical and Electronic Engineers (IEEE)
Standard 279-68, " Criteria for Protection Systems for Nuclear Power Generating Stations," and IEEE Standard 323-71, "!EEE Trial Use Standard:
General Guide for Qualifying Class ! Electric Equipment for Nuclear Power Generating Stations " as these standards became available.
i the staff after July 1974 were evaluated against the more comprehensivePlants t criteria of IEEE 323-74, 'IEEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations," and related ancillary standards as they became available. Regulatory Guide 1.89 " Environmental Qualification of Electric Equipment for Nuclear Power Plants," was issued in November 1974 to 9
endorse IEEE 323-74, and Section 3.11 of the Standard Review Plan (SRP; l
NUREG-0800), " Environmental Qualification of Mechanical and Electrical l
Equipment," was issued in September 1975.
In addressing concerns that were expressed by the Union of Concerned i
Scientists (see Section 3.0 of this evaluation), additional review criteria were established in November 1979 by the Division of Operating Reactors in the i
Office of Nuclear Reactor Regulation, " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors,"
4 commonly. referred to as the D0R Guidelines, to ensure that electrical equipment installed in reactor plants licensed to operate on or before May 23, 1980, was adequately qualifled for postulated accident conditions, j
NUREG-0588, "Interie Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," was issued in December 1979 (and revised in 1
July 1981) to establish review criteria for plants that were licensed after i
May 23, 1980.
Of these more recent plants, those with construction permit safety evaluation reports issued before July 1,1974, were subject to the i
i i
] ;
1
- m r.---.
i l
t criteria of NUREG-0588, Category II, while the other plants were subject to i
the NUREG-0588 Category I criteria. Section 3.!! of the SRP was revised in July 1981 to reflect the criteria that was established by NUREG-0588, and i
Regulatory Guide 1.89 was siellarly revised in June 1984. Thus, depending on when the construction permit and operating license were issued, reactor plants were reviewed to the criteria that were stated in either the 00R Guidelines or in NUREG-0588. Based strictly on the CP-SER issue dates (without regard to special exceptions that may have been allowed), 60 of the 109 operating power reactors (55%) were reviewed to the criteria in the DOR Guidelines; 24 of the operating power reactors (22% were reviewed to the criteria in NUREG-0588, Category !!; and 25 of the ope) rating power reactors (23%) were reviewed to the criteria in NUREG-0588, Category I.
The staff's EQ review criteria for new plant construction presently consists of SRP Section 3.11, RG 1.89, and IEEE 323-74 (along with ancillary standards related to IEEE 323-74 and l
associated Regulatory Guides).
The staff's evaluation of EQ program requirements included a review of the NRC 4
review criteria, implementation guidance, and industry standards referred to above. However, it should be recognized that there are many IEEE standards that are ancillary to IEEE 323 which were not included within the scope of this review. These other standards typically provide specific guidance for component qualification (e.g., penetrations, motors, valve operators, etc.)
based on the criteria stated in IEEE 323. By reviewing IEEE 323, most issues that are of a programmatic nature should be identified while any issues specific to a particular ancillary standard or component type would most likely be missed.
Information that was reviewed by the staff during this effort is summarized in Appendices E through I, and Appendix J is a fairly 1
comprehensive comparison of NUREG-0588 and the DDR Guidelines. The following i
potential issues, which are listed in bold print within these appendices, were j
identified for further staff consideration:
l SRP Section 3.11 (Accendix 0 t
l The SRP is very much out of date and needs to be made current (e.g., the Environmental Qualification Branch is listed as the lead review group; i
there is no reference to the EQ rule; a " central flie" is referred to i
contrary to what was ultimately required by 10 CFR 50.49; and RG 1.89 and IEEE 323-74 are not recognized as the appropriate staff guidance i
documents for satisfying EQ requirements).
Periodic maintenance and surveillance requirements necessary to maintain EQ have not been defined.
IEEE Standards 381, 535, 627, 649, and 650 (and perhaps others) '
i pertaining to EQ have not been endorsed by the NRC.
The SRP suggests that NUREG-0588, RG 1.89, and IEEE 323 may be applicable for qualification of mechanical equipment, but specific guidance has not been provided in this regard.
5 Non-safety-related instruments that could impact the operability of 4
safety-related instruments were not initially included in the scope of equipment that was required to be qualified.
RG 1.97 instruments were not addressed in the initial qualification requirements and it is not clear to what extent (and to what criteria) instruments were required to be qualified.
" Reasons to the contrary" for not upgrading replacement equipment to NUREG-0588 Cat. I requirements appear to be without merit and should be justified.
Purchase specification requirements for replacement parts are not addressed.
~
Use of an " appropriate margin" instead of the one hour margin that is currently required for equipment operability may be better suited to the realities of EQ.
In lieu of attempting to define a "qualifted life," it may be more appropriate to develop methods for addressing and/or monitoring in-service degradation.
Reculatory Guide 1.89 (Accendix F)
The stated purpose of environmental qualification is to avoid " common-cause" failures and (given this) it is not clear why it is necessary to qualify equipment to protect against single failures.
Given the uncertainties inherent in the qualification process, more emphasis may be needed on surveillance, testing and maintenance to maintain the qualification status.
The postulated source ters for EQ purposes has not been corrected to reflect the THI-2 experience.
During a case study, AE00 identified instances where qualified equipment failed due to humidity or moisture intrusion. Qualification of equipment seals and vapor barriers on plants, especially those that'are subject to the DOR Guidelines and NUREG-0588, may not be sufficient.
i
" Reasons to the contrary" for not upgrading replacement equipment to the requirements stated by 10 CFR 50.49 appear to be without merit and should be justified.
Extrapolation of data for EQ purposes, if adequately justified, may be j
well suited for EQ applications.
EQ may not be assured due to Ilmitations that exist relative to humidity testing capabilities.
Use of the " double hump" that is required in the accident proffie for EQ testing (i.e.,
sequential exposure of the test specimen to limiting temperature and pressure conditions) is not representative of design basis conditions and may be too severe.
The release assumptions that were imposed for EQ purposes are not realistic and may not be appropriate.
.g-t
A general exemption for radiation qualification testin.g of equipment exposed to low-level radiation may be well suited for EQ purposes under certain defined circumstances.
Ultimately, justification was not established for the generic temperature profile (i.e., T for PWRs and T
+ 20*F for BWRs) that wasallowedbytheDORGuideUnesandNUREG-0$,8.
Dose rate effects are sometimes surprising and say be unpredictable.
DOR Guidelines (Accendir G)
Emergency shutdown systems "...used to bring the plant to a cold shutdown condition following accidents which do not result in a breach of the reactor coolant pressure boundary together with a rapid depressurization of the reactor coolant system
- were required to be qua11(led by plants subject to the DOR Guidelines.
It is not clear, however, that this was the case for NUREG-0588 plants. Also, since qualification of cold shutdown equipment was not required by the EQ rule, it is not clear to what extent this requirement is currently valid.
Given the industry operating experience (see Appendix F), aging of equipment seals and moisture barriers may need to be better accounted for on plants subject to the DOR Guidelines.
With regard to service conditions in areas outside containment that are exposed to long-ters recirculation, equipment was not required to be qualified for cheatcal sprays. However, this equipment may be subject to actuatio'n of fire suppression systems..
l It is not clear to what extent all test failures (for all attempted l
tests) were documented, evaluated and saved as part of the equipment qualification record (i.e., to what extent was EQ determined based on l
" selective" information).
Environmental conditions for accidents other than for LDCA (such as i
for MSLB) were not defined for at least 65 power reactors. The staff i
failed to recognize this factor in its resolution of Task Action Plan l
Ites A-21.
With regard to the "superheat effects
- of a MSLB, the staff's resolution 1
of TAP A-21 failed to consider single failure considerations.
It is not clear to what extent single failure need not be considered such that l
" local effects" can be excluded.
NUREG-0588 (Accendix H)
It is not clear that the limiting undervoltage and underfrequency conditions were assumed for qualification testing for all plants (i.e.,
seismic vs. other hostile environmental conditions), especially for plants subject to the 00R Guidelines.
l l f
\\
m v
.-m.-
l'-
}
Specific measures may be warranted to sintalze the tapact of dust on j
equipment qualification.
j Overall margin requirements for demonstrating EQ (e.g., one hour minimum i
operating time, thermal aging, etc.) have not been justified in a rigorous and technical manner.
The staff's ' final position
- regardinq the velocity profile in containment during 110w down was pend ng completion of Task A-21.
j However, the staft's resolution of Task A-21 is questionable (see Appendix G information) and this issue say need to IN, revisited.
i More detailed qualification requirements may be needed for mechanical j
equipment.
i Justification for different qualification standards should be j
established.
Comparison of the calculated MSLB surface temperature to the LOCA bulk i
temperature say not assure that the equipment will survive the PtSLB
{
environment.
Civen more realistle assumptions for release fractions, the timing of i
the release, chealcal form of the release, and accident mitigation effects resulting from equipment response, an lamediate and large (TID) source term may be overly conservative and inappropriate.
i i
It may not always be appropriate to require test data to establish EQ.
given the other qualification methods allowed by IEEE 323-74 (i.e.,
l operating experience, analysis, and comblastion of methods).
4 IEEE Standard 323 fAenendix I) i Given the NRC staff view that backfitting the IEEE 323-74 requirements j
would provide "...a small, unquantifiable increase in the level of assurance that equipment is qualified as compared to the significant i
costs that would be involved...." IEEE 323-74 may not be warranted or i
sufficiently justified as a necessary qualification standard for power l
reactors, regardless of plant vintage, j
Comnarison of E0 Review Criteria (Accendtg_n
't l
Technical justification has not been established for the differences that exist in the EQ review criteria stated by the DDR Guidelines, NUREG-0588 (Category I), and NUREG-0588 (Category !!).
}
The qualification of electrical equipment for main steaa line break conditions inside containment was pending resolution of Task Action Plan Item A-21, but resolution of A-21 failed to adequately address this i
issue (see Appendix G).
i 6.0 GENERIC C0tt1UNICATIONS Following the electrical connector test failures that were experienced l
c t
n. w,,
48 by Sandla during EQ testing that was being conducted, and perhaps in response to concerns that were erpressed by the Union of Concerned Scientists, the NRC issued If Bulletin (IEB)icensees to address this issue.
77-05 to alert Ilconsees of the probles and to request specific actions by l Following the pubile,ation of this IE8 in November 1977, many other generic-cosaunications h?ve been issued by the NRC staff to address specific equipment qualification issues as they were identified. Qualification issues that have been address 9d by the staff in this regard have included for example, problems with electrical penetration assemblies, terstnal blocks, electrical cable spilcas, and 11 alt switches. Other generic correspondence that was issued by the staff provided specific guldance as to how EQ requirements pertaining to electrical equipment were to be satisfied, effectively-establishing the preliminary prograssatic requirements for EQ before speciffe' requirements were codified by 10 CFR 50.49.
The staff reviewed such of the generic consunication that was issued pertaining to EQ, focusing primarily on prograssatic requirements that were estabitshed. Appendix K provides a summary of the IE Bulletins CIE8s),
Generic Letters (GLs), etc., that were reviewed by the staff durnng this evaluation and, in particular, the following information is noteworthy:
The DOR Guidelines and WRIG-0588 were transsitted to power reactor licensees by IE8 79-018 in January 1980.
IES79-018, Supplement 2, specified that safety-related electrical equipment subject to EQ requirements are those required to bring the plant to a cold shutdown condition and to mitigate the consequences of an accident. However, the scope of IES79-018 did not include non-safety-related electrical equipment that could tapact safety-related electrical equipment as later stipulated by 10 CFR 50.49.
IE8 79-018, supplement 2, indicated that the staff considered certain l
(unspecified) parts of the DDR Guidelines and WREG-0588 to be optional.
IE8 79-018, Supplement 2, specified that when a determination was made l
that reasonable assurance did not exist that a Class IE electrical component or equipment could perfore its safety-related function, that j
condition should be reported.
i IE8 79-018, Supplement 2, provided significant clartfication regarding l
EQ documentatien requirements (especially for the plants subject to the i
DDR Guidelines and WREG-0588, Category I), essentially laposing the j
guidance contained in WREG-0588 and IEEE 323-74.
IE8 79-018, Supplement 2, speelffed that aging should be included (to some extent) in the testing programs of plants subject to the DOR Guidelines and WREG-0588, Category !!.
l l
The considerations stated by the staff in IES79-018, Supplement 2, for l
deteretning the qualification status of existing equipment purchased i
from a vendor, where a QA program did not exist, do not seem to be very l
quantitative or rigorous; and tend to relax the criteria contained in the DOR Guidelines and WREG-0588.
i 1,
I D
l IE8 79-01, Supplement 3, clarlfled that only one path of cold shutdown equipment needed to be qualifted by licensees.
CL 81-05 indicated that the qualification requirements for WR[G-0737 equipment were described in Appendix 8 of the WRIG (i.e., EQ i
requirements that are not spectfically included in the EQ rule).
CL 86-15 indicated that Itcensees may be able to make a findtag of operabtitty using analysis and partial test data to provide reasonable assurance that equipment will perform its safety function when called upon to attigate the accidents for whfch 1t Is needed. This position tends to relax the criteria that are contained in W REG-0548 for Category I plants.
Potential issues that were identified during this review are highlighted in Appendix K in bold print.
However, one addlttoral issue beyond those Itsted in the appendix was identified based on a broad overview of the information that was studied. Specifica11y:
It is not clear to what extent the various clartftcations and staff positions, including Appendix 8 of WREG-0737, were fully taplemented and which ones are currently appitcable, since they are not specifically referred to by 10 CFR 50.49.
The following additional issues were taken from Appendix K:
I It is not clear which parts of the DDR Guidelines and WRIG-0548 the i
staff considered to be ' optional," and consequently, the minimus i
standards that were found to be acceptable to the staff are not well i
def f ned.
Specific reporting requirements for EQ problems that occur slallar to what was required in IE8 79-018, Supplement 2) say be warran(ted given the uncertainties associated with estabitshing and maintaining l
qualtitcation.
i It is not clear how an " acceptable confidence level' is estabitshed for electrical equipment that must operate "long-ters' which is commensurate with the 1-hour minimum test time requirement that was established for 3
'short-ters' electrical equipment.
i i
EQ requirements with regard to mechanical and flow-induced vibrations, i
setssic effects, and other GDC considerations, were not addressed under i
the EQ rule.
i It is not clear why it was not necessary to tapose the single failure i
criteria for qualification of cold shutdown equipment (i.e., IE8 79-01, Supplement 3, only required one train of cold shutdown equipment to be qualified).
4 i
J
)
7.0 RESEARCH STUDIES RELATED TO EQ A substantial amount of EQ research has been conducted by the NRC and also by
j the nuclear industry in order to address questions and concerns that have been
{
raised in this area. Much of this work was done in order to address concerns that were expressed by the Union of Concerned Scientists in a Petitles for Commission Action which was submitted in November 1971; and more recently, additional research has been conducted partly in support of the license renewai Initiative, but also to address plant aging concerns.
In general, [Q research studies that have been completed provide additional information that needs to be considered Ard further developed. A lot of work has been completed, and is continuing, to assess condition monitoring techniques; and some of the age-old issues (e.g., quallf ted life, synergises, j
required operating times, etc.) have been further assessed. Some interesting PAA studies have also been completed which provide tesightful Information that may be useful in better focusing (Q requirements in the future. Appendix L is a listing and descriptive sunnary of the research studies that were reviewed by the staff during this effort, and potential issues that were identified during this review are Itsted within the appendix in bold print. However, the following additional issues beyond those itsted in Appendia L were identified based on a broad overview of the research informatton:
j NRC research activities have not been very successful in resolving the
' age-old" [Q tssues that were Inttlally identified.
i A lot of research has been completed and much more expertence has been obtained in the area of EQ since 10 CFR 50.4g was issued, but focused j
i NRC programs and initiatives apparently do not exist to continually monitor progress in this area and to make use of this information for restructuring, directing, and improving EQ program requirements.
The following potential issues, taken from Appendia L, were also identified
]
for further staff consideration:
l Deteretnation of a "qualtfled life" may not be possible given the current state of the art.
The degree of conservattse butit into source term requirements may not be appropriate given more realistic assumptions.
i i
EQ relative to fire retardant coating (and other fire protection j -
features) say not be adequate.
I Given the state of the art that was in existence at the time IEEE 323-74 was developed and the Italtations that existed, it would seem that EQ program requirements may have been misdirected (especially with regard i
to the required deterstnation of "qualtfled life" and the absence of i
surveillance requirements for obtaining advance warning of significant l
degradation).
I The one hour operating time requirement continues to be a controverstal a
l 1ssue.
The use of " margin on top of margin
- continues to be controverstal.
i 14 -
-... --- - - ~_.
,m-=
Use of in-containment terminal blocks in safety-related circuits may not be appropriate.
i Very little EQ research has been conducte1s on pressure switches, RTDs, j
pressure transaltters, and valve operators.
/
Research is only just beginning to assess the adequacy of EQ for AG 1.97 functions. The areas of primary concern include reactor coolant level instrumentation, core exit thermocouples, containment area radiation monitors, halogen and particulate sampling, and coolant activity measurements.
)
PRA studies indicate that EQ Master Lists may need to be updated to include additional equipment.
Instrumentation requirements for EQ purposes may not be well focused.
i j
EQ relative to hydrogen burn scenarios may not be adequate.
1 The post-mandrel bend test say be too severe for EQ purposes.
The need for rigorous qualification of equipment located outside containment may not be warranted.
l The need and/or ability to establish qualification beyond a two to four l
week period is questionable.
1 i
The need to consider oxygen diffusion effects on aging during normal j
plant operation may not be warranted given realistic assumptions.
EQ of safety-related equipment other than electrical is questionable.
Kapton vulnerabilities may need to be addressed for nuclear i
appilcations, i
e' The color of insulation material may influence the rate of degradation of the insulatten material.
j The qualification of bonded jackets is questionable.
i i
Industry representatives have presented reasonable and compelling i
arguments strongly suggesting that the methodology and practices for l
estabitshing and maintaining EQ should be reassessed.
Failure of other electrical components such as electrical penetrations i
and elecconnector assemblies may be more lorortant than the failure of trical cables, ii Qualification of solenoid valves for some applications is questionable.
I 8.0 SUPMARY s
l The staff has completed a fairly comprehensive review of information f
15 -
1 y
,n
o i
l concerning EQ requirements, focusing heavily on research and development aspects. As a result of this review, many issues were identified for further 4
consideration within the overall context of the [Q Task Action Plan.
In order i
i to arrive at a conservative end complete listing of potential issues, ne attempt was made to further judge or screen any of the issues after they were identified. Therefore, it should be emphaslied that these are potential issues, some of which say be dismissed under closer scrutiny.
l This section of the evaluation is a suarary of the various issues that were i
identified in Sections 3 through 7.
In order to faellitate this effort, any Issue dupilcation from the various sections was ellainated and the remaining Issues were each placed in one of the following six cate (a) EQ j
Methods, Culdance, and Review Criteria - General Issues:gories:
(b) EQ Methods, GWidenit d) Ceners' fechnical issues:and Rfylpw (filffle e Ipfflflf lilWil, (f f [0 Pffgram implementillen issues: (
(e)Speelffe"echnica Issues; and (f) Other Program / Policy Issues. The specific issues that were identified
]
during this review are listed below and will be addressed in the staff's final i
report (i.e., EQ-TAP Action Ites 3 9):
E0 Methodoloor and Practices - Ceneral issues:
i Industry representatives have presented rational and compelling i
arguments strongly suggesting that the methodology and practices for j
establishing and maintaining EQ should be reassessed.
Different EQ standards were irposed (i.e., 00R Guidelines, NUREG-0588 Category I, and NUREG-0588 Category !!) without supporting technical justification as to:
(a) why more rigorous standards were warranted, i
and (b) why " progressively less strict standards" were adequate for the older plants.
1 Given the Regulatory Requirements Review Comalttee and the NRC staff i
view that backf;tting the IEEE 323-74 requirements would provide "...a i
small, unquantifiable increase in the level of assurance that eq9tpoent
~
is qualified as compared to the significant costs that would be l
Involved... " IEEE 323-74 say not be warranted or sufficiently justified i
as a necessary qualification standard for power reactors, regardless of l
when the Construction Permit Safety Evaluation Report was issued.
The current version of IEEE 323 may be better suited for demonstrating i
EQ than the 1974 version since much more informatica and experience are j
available now than there was when IEEE 323-74 was endorsed by the staff.
Current " state of the art capabilities' may not be sufflctently i
developed to support existing EQ requirements, such as deterstnation of a "qualifted life."
The SRP (Section 3.11) is very such out of date and needs to be made j
current (e.g., the Environmental Qualification Branch is listed as the lead review group; there is no reference to the EQ rule; a " central flie" is referred to contrary to what was ultimately required by 10 CfR 50.49; and RG 1.89 and IEEE 323-74 are not recognized as the epropriate staff guidance documents for satisfying EQ requirements).
IEEE Sta',dards 381, 535, 627, 649, and 650 (and perhaps others) pertaining to [Q have not been endorsed by the NRC.
PAA studies indicate that EQ Master Lists say need to be updated to include additional equipment.
ED Methods. Guidance and Review criteria - soecifte issues:
The need for rigorous qualtffcation of equipment located outside containment may not be warranted.
The need and/or ability to establish post-accident qualification beyond a two to four week period is questionable.
In lieu of attempting to define a "qualif ted life," it may be more appropriate to develop methods for addressing and/or acnf toring in-service degradation.
L Given more realistic assumptions for the release fractions, the timing of the release, the cheetcal form of the release, and accident mitigation effects resulting from equipment response, an immediate and large source ters (TID) may be overly conservative and inappropriate.
A general exemption for radiation qualification tetting of equipment exposed to low-level radiation may be well suited l'or EQ purposes under certain defined circumstances.
Better definition is needed for which instruments are riquired to be qualified.
L
" Reasons to the contrary" for not upgrading replacement easipment to the requirements stated by 10 CFR 50.49 appear to be without serit and should be justified.
Purchase specification requirements for rep',acement parts have not been addressed relative to EQ.
Periodic maintenance and surveillance requirements necessary to maintain EQ have not been defined.
Methods have not been estabitshed to ensure that installation and maintenance practices do not jeopardize equipment qualification.
Margin requirements for demonstrating EQ (e.g., one hour minimum operating time, thersal aging, vtc.) may be tuo severe and without sufficient justification; overall sargin requirements need to be better defined with supporting technical justification.
It may be appropriate to perfors " aging
- of equipment in the same functional state as it is used in the plant (i.e., energized or de-energized).
The " double peak" requirement (i.e., exposure to two cycles of maximus
I I
l i
1
)
temperature and pressure) is not representative of deste basis
]
conditions and say be too severe.
i Theneedtoconsideroxygendiffusioneffectsonagingduringnormal plant operation may not be warranted given realistic assergt.ons.
Hualdity testing capabilities say not be sufficient to assure equipment j
qualification.
5 i
In areas outside containment that are exposed to long-tera j
recirculation, equipment was not required to be quallfled for chemical j
sprays. However, this equipment may be subject to actuation of fire suppression systems.
j The post-sandrel bend test may be too severe and perhaps unwarranted for
]
EQ purposes.
I Extrapolation of data, if adequately justified, may be well suited for EQ appitcattens.
]
It may not always be appropriate to require test data in order to establish EQ, given the other qualification methods allowed by IEEE 323-74 (i.e., operating experience, analysis, and combination of methods).
Given the complexities and uncertainties involved, labe atory i
accreditation may be necessary to assure that EQ testing is properly performed.
The generic temperature profile that was allowed by the 00R Guidelines l
and NUREG-0588 for equipment qualification (i.e., T, for PWRs and T, + 20*F for BWRs) was not fully justified.
u l
Comparison of the calculated MSLB surface temperature to f.be LOCA bulk temperature may not assure that the equipment will survive the MSLB l
environment.
E0 Procram imolementation Issues:
It is not clear to what extent the various clarifications and staff l
positions that were stated in Generic Letters, IE Bulletins, Appendix B l
of NUREG-0737, etc., were fully implemented, and which ones are l
currently appilcable since they are not specifically referred to by 10 CFR 50.49.
It is not clear which parts of the 00R Guidelines and NUREG-0588 the i
i staff considered to be ' optional," and consequently, the minimua l
standards that were found to be acceptable to the staff are not well defined.
It is not clear to what extent all test failures (for all attempted tests) were required to be documented, evaluated, and saved as part of.
~ - - - - ~ - - - - - ~ -
~~~"]
the equipment qualification record (i.e., to what extent was EQ determined based on " selective" informaticn).
It is not clear to what extent the limiting undervoltage and underfrequency conditions were assumed for qualification testing for all plants (i.e., seisaic vs. other hostile environmental conditions),
especially for plants subject to the 00R Guitfelines.
Emergency shutdown systems "...used to bring the plant to a cold shutdown condition following accidents which do not result in a creach of the reactor coolant pressure boundary together with a rapid depressurization of the reactor coolant system" were required to be qualif ted by plants subjec to the DDR Culdelines.
It is not clear, however, that this was the case for NUREG-0588 plents. Also, since qualification of cold shutdown equipment was not required by the EQ rule, it is not clear to v5at extent this requirement is currently (lj valid.
Mon-safety-related instruments that could impact the operability of safety-related instruments were not initially included in the scope of equipment that was regrired to be qualified, and implementation of this requireme d may not be uniform among all plants.
RG 1.97 instruments were not addressed in the initial qualification requirements and it is not clear to what extent (and to what criteria) g>'
instruments were required to be qualif ted.
j General Technical Issues:
Qualification of equipment other thac electrical, and equipment qualification for other conditions (e.g., mechanical and flow-induced vibration, seismic effects, dynamic effects, etc.) Have not been j
reviewed within the EQ framewc,rL
)
i j
Research is only just beginning to assess the adequacy of EQ for RG 1.97 functions. The araas of primary concern include reactor coolant level i
instrumentation, core exit therwocouples, containment area radiation monitors, halogen and particulate sampling, and coolant activity measurements, i
Yery little EQ research has been conducted on pressure switches, RTDs, l
pressure transmitters, and valve operators.
Failure of other electrical components such as electrical penetrations and connector assemblies may be more important than the failure of electrical cables.
i Specific measures may be warranted to minimize the impact of dust on equipment quellfication.
l Specific Technical Issues:
Environmental conditions for accidents other than for LOCA (such as I
i j,
t
P i'
i for MSLB) were not defined for at least 65 power reacto,s. The s.aff c
failed to recognize this factor in its resolution of Task Action Plan Itea A-21.
The staff's "ffnal position
- regarding the velocity profile in containment during blow down was pending completion of Task A-21.
However, the staff's resolution of Task A-21 was incomplete and this issue may need to be revisited, Qualification of equipment seals and vapor barriers on plants, i
i especially those that are subject to the DDR Guidelines and NUREG-0588, i
may not be sufficient.
j Hydrogen burn scenarios may result in conditions that exceed the EQ envelope.
Temperature and radiation stratification may result in conditions that l
exceed the EQ envelope.
\\ '
i Terminal blocks for use inside containment may not be sufficiently a
l quallfled.
.i Qualification of solenoid valves for some app 11 cations may not be sufficient.
U t
i Bonded jackets on electric power and/or control cables may not be sufficiently qualifted.
1
]
The color of insulation material may influence the rate of degradation of the insulation material.
i j
Kapton vulnerabilities may need to be addressed for nuclear applications.
Epoxy compound used for potting electrical penetrations may not be i
qualified to the temperature conditions that are experienced post-LOCA l
and/or during a MSLB.
fire retardant coatings (and other fire protection features) may not be a
acccunted for within tna existing EQ framework.
Other Procram/ Policy Issues:
l i
Given the state of the art that was in existence at the time IEEE 323-74 was developed and the limitations that existed, it would seem that EQ program requirements may have been misdirected (especially with regard to the required determination of 'qualifted life" and the absence of surveillance requirements for obtaining advance warning of significant i
degradation).
A lot of research has been completed and much more experience has been i
obtained in the area of EQ since 10 CFR 50.49 was issued, but focused i
NRC programs and initiatives apparently do not exist to continually
.l i
h,
5 monitor progress in this area and to make use of this information for restructuring, directing, and improving EQ program requireesnts.
HRC research activities have not been entirely successful in resolving the " age-old" EQ issues that were initially identified of HUREG/CR-4301 in Appendix L for specific examples). (see the summary Additional reporting requirements for EQ problems that occur (i.e.,
problems that occur during qualification testing as well as problems that occur during plant operation) may be warranted given the uncertainties associated with estabitshing and maintaining equipment qualification.
It is not clear why it was not necessary to impose the single failure criteria for qualification of cold shutdown equipment (i.e., IES 79-01, Supplement 3, only required one train of cold shutdown equipment to be qualifled).
With regard to the "superheat effects" of a MSLB, the staff's resolution of TAP A-21 failed to consider single failure considerations.
It is not clear to what extent single failure need not be considered such that
" local effects" can be excluded.
As stated in Regulatory Guide 1.89, the purpose of environmental qualification is to avoid " common-cause" failures. Given this, It is not clear why it is necessary to qualify equipment to protect against single failures.
The SRP suggests that HUREG-0588, RG 1.89, and IEEE 323 may be applicable for qualification of mechanical equipment, but specific guidance has not been provided in this regard.
l Principle Contributor: James Tatum (SPLB)
,1 ll 1
3.
, emme esp e * -
)
l o
)
l
.i 4
l 1
I i
1 4
i
~
I 4
1 4
t 1
i i
i d
]
l, I
i T
l THIS PAGE INTENTIONALLY BLANK i
I s
l i
G
,1 I
f s
I e
O
9.0 REFERENCES
Aroonne National Laboratory (ANL)
ANL Drt,ft Report re:
" Risk Impact of EQ Requirements for Operating FIN-A2336. Task Reactors," November 9, 1993 Assignment No. 5 Atomic Industrial Forum Letter From R. Eckert to H. Denton re: industry position.
of elements of EQ, January 4,1982 Letter from R. Eckert to W. Dircks re: industry position of elements of EQ, February 2,1982 Letter From R. Eckert to R. Vollmer re: industry position of elements of EQ, August 24, 1982 Letter from R. Eckert to R. Vollmer re: industry position of elements of EQ, December 28, 1982 Electric Power Rittty_ch Institute (FPRI)
EPRI EL/NP/CS-5914-SR
" Proceedings: 1993 EPRI Workshop on Power Plant Cable Condition Monitoring," July 1988 EPRI NP-1558 "A Review of Equipment Aging Theory and Technology," September 1980 i
1 EPRI TR-102399
- Proceedings: 1993 EPRI Workshop on Powe'r Plant Cable Condition Monitoring," June 1993 I
Franklin Research Center If.Efd FRC Lecture L-A5300/EPRI Review of Aging Theory and Technology,"
September 17, 1980 Gesellschaft fur Reaktorsicherheit - GRS Letter From S. to:sner to S. Aggarwal regarding the 1983 j
Symposium on Nuclear Power Systems, October 5, 1983 Technical Paper
" German Philo:ophy and Practice of Aging within Qualification of Electrical Equipment for Safety Systems of Nuclear Power Plants" _
1
__A_m.__-_.
-_____m____m_____m___.____.__.__..__.__..___.____.__-_____________.,____..____-m-.__-_-_--m___
_w
)
institute of Electrical and Electronics Encinggn flEEE)
IEEE 279-71
" Criteria for Protection Systems for Nuclear Power Generating Stations" IEEE 323-71
"!EEE Trial Use Standard:' General Guide for Qualifying Class ! Electric Equipment for Nuclear Power Generating Stations," April 1971.
IEEE 323-74
"!EEE Standard for Qualifying Class IE Equipment for Nuclear Power Generating Stations," February 1974 International Atomic Enercy Acency flAEA)
IAEA Technical Comittee "U.S. Practices for Review, Upgrading and Presentation Maintaining Equipment Qualification," by Philip M. Holzman of Strategic Technology and Resources, Inc., June 1993 Nuclear Reculatory Commission (NRC)
" Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," January 21, 1983 10 CFR 50.55a
" Codes and Standards," June 12, 1971 l
" General Design Criteria for Nuclear Power Plants," February 20, 1971 l
" Design Bases for Protection Against Natural Phenomena" GDC 4
" Environmental and Dynamic Effects Design Bases" GDC 23
" Protection System Failure Modes" i
00R Guidelines
- Guidelines for Evaluating Environmental i
Qualification of Class IE Electrical Equipment in Operating Reactors," November 1979 Generic Letter 81-05 "Information Regarding the Program for Environmental Qualification of Safety-Related l
Electrical Equipment," January 19, 1981 Generic Letter 82-09
" Environmental Qualification of Safety-Related Electrical Equipment," April 20, 1982 m-m.m
. m m
ua-m m
e
.d-
^*
maamma i
Nuclear Reaulatory Commission (NRC) 1 (cont.)
Generic Letter 85-15 "Information Relating to the Dead 11aes for
{
Compilance with 10 CFR 50.49, ' Environmental
.l Qualification of Electric Equipment important to
)
Safety for Nuclear Power Plants,'" August 6, 1985 Generic Letter 86-15 "Information Relating to Compliance with 10 CFR 50.49, ' Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants,'" September 22, 1986 Generic Letter 88-07
" Modified Enforcement Policy Relating to 10 CFR 50.49, ' Environmental Qualification of Electric Equipment important to Safety for Nuclear Power Plants,'" April 7, 1988 IE Bulletin 77-05
" Electrical Connector Assemblies," November 8, 1977 IE Bulletin 77-05A
" Electrical Connector Assemblies," November 15, 1977 IE Bulletin 77-06
" Potential Problems with Containment Electrical Penetration Assemblies," November 22, 1977 IE Bulletin 79-01
" Environmental Qualification of Class IE Equipment," February 8, 1979 IE Bulletin 79-018
" Environmental Qualification of Class IE Equipment," January 10, 1980 IE Bulletin 79-01B
" Environmental Qualification of Class IE (Supplement)
Equipment," February 29, 1980 IE Bulletin 79-01B
- Environmental Qualification of Class IE (Supplement No. 2)
Equipment," September 29, 1980 IE Bulletin 79-01B.
" Environmental Qualification of Class IE (Supplement No. 3)
Equipment " October 24, 1980 IE Circular 78-08
" Environmental Qualification of Safety-Related Electrical Equipment at Nuclear Power Plants,"
May 31, 1978 Letter From D. Ross to all construction permit and operating license appilcants, "Issuanced of NUREG-0588, ' Interim Staff Position on Equipment Qualification of Safety-Related Electrical Equipment
'" February 5, 1980 ~~~
5 Nuclear Reaulatory Commission (NRC)
(cont.)
Letter From D. Ross to operating license appilcants,
' Qualification of Safety-Related Electrical Equipment," February 21, 1980 Memorandum From R. Fett to L. Tong. *Trar,smittal of Trip Report," August 5, 1977 Memorandum from S. Chilk to L. Gossick, " Staff Requirements
- Discussion on Union of Concerned Scientists Petition for Emergency and Remedial Action,"
November 8,1977 Memorandum From E. Case to the Comission, " Union nf I
Concerned Scientists' Petition of November 4, l
1977, for Emergency and Remedial Action,"
November 9, 1977 Memorandum From E. Case to the Commission, " Union of Concerned Scientists' Petition of November 4, 1977, for Emergency and Remedial Action,"
November 18, 1977 Memorandum From E. Case to the Commission, " Union of l
Concerned Scientists' Petition of November 4
- 1 1977, for Emergency and Remedial Action,"
November 22, 1977.
Memorandum From E. Case to the Comission, "Use of
)
Electrical Connectors in Safety Related Systems,"
l November 25, 1977 i
Memorandum From E. Case to the Comission, " Union of j
Concerned Scientists' Petition of November 4, 1977 for Emergency and Remedial Action,"
December 6, 1977 4
Memorandum From E. Case to the Comission, " Union of I
l Concerned Scientists Petition," December 15, 1977
)
Enclosure "NRC Staff Report on Union of Concerned Scientists' Petition for Emergency and Remedial Action."
i Memorandum from E. Case to the Comission, " Union of.
Concerned Scientists' Petition - the Use of l
Electrical Connectors in Safety Systems,"
January 6, 1978 i
Memorandum From E. Case, " Union of Concerned Scientists'
. Petition," January 13, 1978 I
i
=
hciearRecu1 story Comission (NRC)
(cont.)
Memorandum From E. Case to the Comission, " Union of Concerned Scientists' Petition," January 20, 1978 Memorandum From E. Case to the Comission, " Union of Concerned Scientists' Petition," February 10, 1978 Memorandum from E. Case and E. Volgenau to the Comission,
" Union of Concerned Scientists' Petition,"
March 23, 1978 Memorandum From E. Case to the Comission, " Union of Concerned Scientists' Petition," March 30, 1978 Memorandum from Comissioner Bradford to 1.. Gossick, " Staff Actions in Response to Deficiencies in Connector Spilces," April 26, 1978 Memorandum From E. Case to the Comission, "Comissior's Memorandum and Order [of. April 13,1978): UCS' Petition for Emergency and Remedial Action,"
May 4, 1978 Memorandum From E. Case to the Comission, " Union of Concerned Scientists' Petition," May 12, 1978 Memorandum From S. Chilk to i.. Gossick, " Union of Concerned Scientists (UCS) Petition for Reconsideration, l
May 2, 1978," June 21, 1978 Memorandum From E. Case to the Comission, " Union of i
Concerned Scientists' Petition for
{
Reconsideration Dated May 2, 1978," July 6, 1978 l
Memorandum From H. Denton to the Comission, " Union of Concerned Scientists' Petition for Reconsideration Dated May 2, 1978," August 31, 1978 Memorandum from R. T. Kennedy to L. Gossick, "UCS Petition for Reconsideration," August 7, 1979 Memorandum From H. R. Denton to Comissioner Kennedy, "UCS Petition for Reconsideration," August 24, 1979 Memorandum From W. Otrks to Comissioner Kennedy, "01E Bulletin 79-018, ' Environmental Qualification of 1
Class IE Equipment,'" February 15, 1980
)
I i,
w
1
.1 Nuc1 ear Raou1storv Commission (NRC)
(cont.)
i Memorandum from H. Denton to R. Minogue " Standard for Qualification of Safety Related Equipment,"
September 2, 1980 t
Memorandum From R. Smith to W. Dircks, "Rulemaking on Environmental Qualification of Safety Grade Electric Equipment," January 6, 1981 i
Memorandum From W. Morrison to R. Minogue, N. Moseley,
- 11. Thornburg, and T. Murley, " Proposed Rulemaking and Associated Regulatory Guide 1.89,"
March 2, 1981 Memorandum From Z. Rosztoczy to G. Knighton, " Proposed i
Rulemaking and Associated Regu1 story Guide 1,89,*
March 23, 1981 Memorandum From S. Aggarwal to E. Wenzinger, " Proposed Rulemaking and Associated Regulatory Guide 1.89 l
]
on Environmental Qualification," April 7,1981 Memorandum From V. Stello to R. Minogue, " Proposed i
Rulemaking, ' Environmental and Seismic Qualification of Electrical Equipment for Nuclear 4
i Power Plants,'" September 4, 1981 Memorandum From D. Sullivan to G. Arlotto, " Background Information on Proposed Revision to Regulatory Guide 1.89, ' Environmental Qualification of Electric Equipment for Nuclear Power Plants,'"
January 21, 1982 Memorandum from R. Vollmer to D. Eisenhut, " Clarifications to Environmental Qualification Requirements,"
February 17, 1982 0
Memorandum from S. Trubatch to Files, " Final Rule on Environmental Qualification of Safety-Related
'i Electric Equipment for Nuclear Power Plants,"
May 28, 1982 Memorandum From S.Chilk to W. Dircks, "SECY-82-207/82-207A -
.l 2
Final Rule, ' Environmental Qualification of Safety-Related Electric Equipment for Nuclear Power Plants,'" June 25, 1982 Memorandum from L. Bickwit, Jr. to the Commission, "Draf t Final Rule ' Environmental Qualification of Electric Equipment important to Safety for i
Nuclear Power Plants,' SECY-82-207C/2070,"
September 2, 1992 1. - -
I-I l
1 Nuclear Reculatory Commission (NRC)
(cont.)
Memorandum From W. Di M to Commissioner Ahearne,
" Environmental Qualification Rule," October 26, 1982 Memorandum From Chairman Palladino to S. Chilk "SECY 207C - Final Rule, ' Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants,'" November 1,'1982
)
Memorandum from S. Chilk to W. Dircks, " Staff Requirements -
Affirmation / Discussion and Vote, 3:30 P.M.,
Thursday, January 6,1983, Comissioners Conference Room (Open to Public Attendance),"
3 January 7, 1983 Memorandum From T. Speis to H. Denton. "SECY-82-207C/D -
Final Rule on Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," February 4, 1983 Memorandum From R. Vollmer and R. Mattson to D. Eisenhut, I
" Guidance for Licensees and License Applicants to Demonstrate Compliance with 10 CFR 50.49,
' Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants,'" April 8, 1983 Memorandum From W. Olmstead to R. Vollmer, " Interpretation of 10 CFR 50.49," October 4,1983 Memorandum From K. V. Seyfrit to S. K. Aggarwal, " Regulatory Guide 1.89, Rev.1, ' Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants'," October 20, 1983 Memorandum From Chairman Palladino to the Commissioners,
" Environmental Qualification Meeting Scheduled for January 6,1984," January 3,1984 i
Nemorandum from H. Taylor to CRGR rtmbers, "Information Related to CRGR Meeting No. 54," January 17, 19P,4 Memorandum From A. Gody to E. Beckjord, " Periodic Review of low-Priority Generic Safety Issues," June 16, 1993 NUREG-0413
" Staff Report on the Environmental Qualification of Safety-Related Electrical Equipment,"
)
February, 1978.
e
l l
Nuclear Reaulat.gn Commission fNRC)
(cont.)
NUREG-0458 "Short Term Safety Assessment on the Environmental Qualification of Safety-Related Electrical Equipment of SEP Operating Reactors,"
May 1978 NUREG-0588 (For Comment
" Interim Staff Position on Environmental Version)
Qualification of Safety-Related Electrical Equipment," December 1979 NUREG-0588 Revision 1
" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment, including Staff Responses to PJblic Comments" l
NUREG-0800 Standard Review Plan Section 3.11
" Environmental Qualification of Mechanical and Electrical Equipment," September 1975 NUREG/CP-0129
" Workshop on Program for Eliminttion of l
Requirements Marginal to Safety," September 1993 NUREG/CR-4144 "Importance of Ranking 8ased on Aging l
Considerations of Components Included in Probabilistic Risk Assessments," April 1985 NUREG/CR-4301
" Status Report on Equipment Qualification Issues Research and Resolutien,' September 1986 l
" Equipment Qualification (EQ) - Risk Scoping Study," January 1989 1,
1 NUREG/CR-5772 (Vol. 1)
" Aging, Condition Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class IE Electric Cables," Crosslinked Polyolefin Insulation, I
August 1992 NUREG/CR-5772 (Vol. 2)
" Aging, Condition Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class IE Electric Cables," Ethylene Propylene Rubber Insulation, November 1992 NUREG/CR-5772 (Vol. 3)
" Aging, Condition Monitoring, and Loss-of-Coolant Accident (LOCA) Tests of Class IE Electric Cables," Miscellaneous Cable Insulaticn Types, November 1992 Regulatory Guide 1.89
" Environmental Qualification of Electric Equipment for Nuclear Power Plants," November 1974 '
~ '.'i 'J S 5 4 i.
~
' ~ ~ - ~ ~ ~
~ ^~
i i
Nuclear Reoulatory i
Commission (NRC) j (cont.)
" Environmental Qualification of Certain Electric (Revision 1)
Equipment Important to Safety for Nuclear Power Plants," June 1984 4
" Evaluation of Responsed to IE Bulletins 77-05 l
and 77-05A, Staff Actions Related to Bulletins, i
and Related Quality Assurance Matters," June 26, 1978 i
i SECY-79-112A i
" Union of Concerned Scientists' Petition for i
Reconsideration - Environmental Qualification of 2
Electrical Equipment," March 15, 1979 i
" Analysis of Alternatives for Conducting i
Independent Verification Testing of Envircnmentally Qualified Equipment," July 1, j
1980 j
" Order to Confirm the Requirements of Comission j
Memorandum and Order of May 23, 1980, Regarding Environmental Qualification and Fire Protection,"
l August 6, 1980
)
" Impact of TMI Lessons Learned on the Staff Requirements for Environmental Qualification of Safety Related Electrical Equipment,"
September 19, 1980 l
" Completion of Unresolved Safety Issue A-24,"
August 7, 1981 1
" Equipment Qualification Program Plan,"
August 20, 1981 SECY-81-603
" Proposed Rulemaking, ' Environmental Qualification of Electric Equipment for Nuclear i
Power Plants,'" October 20, 1981 l
SECY-81-603A
" Proposed Rulemaking, ' Environmental j
Qualification of Electric Equipment for Nuclear Power Plants,'" November 4, 1981 4
" Proposed Rulemaking, 'Er.vironmental Qualification of Electric Equipment for Nuclear Power Plants,'" November 16, 1981 T
" Staff Requirements - SECY-81-6038 - Proposed j
Rulemaking, ' Environmental Qualification of Electric Equipment for Nuclear Power Plants."
i february 4, 1982 1
1 1
i Nuclear Reaulatory i
Commission (NRC) j (cont.)
" Final Rule, ' Environmental Qualification of j
Safety-Related Electric Equipment for Nuclear Power Plants.'" May 24, 1982 i
i SECY-82-207A
" Final Rule, ' Environmental Qualification of Safety-Related Electric Equipment for Nuclear
]
Power Plants,'" June 9, 1982 SECY-82-207C
" Final Rule, ' Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants,'" July 27, 1982 SECY-82-2070
" Supplement to SECY-82-207C Final Rule, j
- Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants,'" August 26, 1982 i
" Decision in Union of Concerned Scientists v.
Nuclear Regulatory Commission and United States of America, U. S. Court of Appeals for the District of Columbia Circuit No. 82-2000,"
j July 5, 1983 l
SECY-83-457A "NRC Response to Court Decision Vacating Interim Rule on Environmental Qualification Deadline (SECY-83-457)," December 7,.1983 i
" Environmental Qualification Policy Statement -
SECY-83-457," February 21, 1984 Technical Paper
" Overview of Environmental Qualification of l
Electric Equipment Important to Safety at Nuclear l
Power Plants," authored by A. Masciantonio, l
R. LaGrange, and V. Noonan Unresolved Safety Issue
" Main Steam Line Break Inside Containment -
l A-21 Evaluation of Environmental Conditions for i
Equipment Qualification" i
Unresolved Safety Issue
" Qualification of Class IE Safety-Related j
A-24 Equipment
]
f{gglear Safety Analysis Center (NSAC)
NSAC-58 "A Guide to Qualification of Electrical Equipment j
for Nuclear Power Plants," September 1983 I
i l m
- - - - - -- --- -~ ~ - - - - --^^-
Nuclear Utility Grouo on l
Eautoment Oualification l
(NUGEO) i Letter From N. Reynolds to N. Palladino, "NRC Staff's j
Proposed Rule Regarding Environmental I
Qualification of Electrical Equipment,"
l September 10, 1982 I
Sandia National l
Laboratories (Sandia) j!
Letter From L. Bonzon to W. Rutherford, "Coments on
{
NUREG-0588, Revision I, Draft 1 dated November 1980," January 13, 1981 t
Union of Conegnqtd i
Scientists (UCS)
]
j Letter Supplemental Affidavit of Robert E. Pollard, i
j
" Union of Concerned Scientists' Petition for Emergency and Remedial Action," November 17, 1977 I'
Letter From Legal Counsel for the Union of Concerned j'
l Scientists to the Comission, " Union of Concerned Scientists' Petition for Emergency and Remedial
{
Action," November 23, 1977
)
l
)
j Letter From the Union of Concerned Scientists to the
{
Comission " Union of Concerned Scientists' I
Petition for Emergency and Remedial Action,"
l January 9, 1978 l
Letter From the Union of Concerned Scientists to the'
)
Comission, " Union of Concerned Scientists' Petition for Emergency and Remedial Action,"
{
January 20, 1978 i,
j 33 -
1 t-m.
mm-.
m m
i APPDCIX A Abbreviations ACRS Advisory Committee on Reactor Safeguards ADS Automatic Depressurization Systes AEC Atomic Energy Commission AEDD Office for Analysis and Evaluation of Operational Data AHL Argonne National Laboratory ATR Attenuation Total Reflectance [ Spectrometry]
BlW Boston Insulated Wire BWR Bolling-Water Reactor
{
Cat.
Category CAT Computer Assisted Tomography CDF Core Damage frequency CEI Cleveland Electric illuminating Company
)
CFR Code of Federal Regulations CP Construction Permit CP-SER Construction Permit - Safety Evaluation Report CRGR Committee to Review Generic Requirements CSPE Chlorosulfonated Polyethylene CT Computed X-Ray Tomography DBA Design Basis Accident DBE Design Basis Event l
DOR Division of Operating Reactors DSC Differential Scanning Calorimeter ECAD Electronte Characterization and Diagnostic [ System) i ECCS Emergency Core Cooling Systes EDO Executive Director for Operations EMAP Electrical Monitoring and Analysis Program EPA Electrical Penetration Assembly EPR Ethylene Propylene Rubber EPRI Electric Power Research Institute EQ Equipment Qualification EQ Environmental Qualification EQB Equipment Qualif1 cat 1on Branch EQPP Equipment Qualification Program Plan i
EQ-TAP Environmental Qualification - Task Action Plan I
FRC Frknklin Research Center FSAR Final Safety Analysis Report FTIS Fourier Transforu Infrared Spectroscopy FW Feedwater i
GDC Ceneral Design Criteria GE General Electric i
GL Generic letter HELB High Energy Line Break HEP 8 High Energy Pipe Break HPCI High Pressure Coolant Injection HPCS High Pressure Core Spray w,,--
o f
~
IAEA International Atoele Energy Agency IE Office of Inspection and Enforcement IES IE Bulletin IEEE Institute of Electrical and Electronics Engineers LEPT Low-Energy Pulse Testing LER Licensee Event Report LOCA Loss of Coolant Accident MASC Maintenance and Analysis of Station Cables NOV Motor-0perated Valve MSIV Main Steam Isolation Valve MSL8 Main Steam Line Break NBS National Bureau of Standards i
NIR Near Infrared Reflectance NMAC Nuclear Maintenance Application Center NRC U.S. Nuclear Regulatory Cosnission NRR Office of Nuclear Reactor Regulation NSSS Nuclear Steam Supply System NTOL Near Tern Operating License NSAC Nuclear Safety Analysis Center
{
NUGEQ Nuclear Utility Group on Equipment Qualification OGC Office of the General Counsel OIT 0xidation Induction Time OR Operating Reactor i
ORNL Oak Ridge National Laboratory 050 Office of Standards Development i
1 PE Polyethylene PNPP Perry Nuclear Power Plant PORY Power-Operated Relief Valve PRA probabilistic risk assessment PSA Probabilistic Safety Analysis l
PWR Pressurized Water Reactor QA Quality Assurance QTE Qualification Testing Evaluation [ Program)
RCIC Reactor Core Isolation Cooling l
RES Office of Nuclear Regulatory Research i
RG Regulatory Guide RPS Reactor Protection System l
RRRC Regulatory Requirements Review Committee RTD Resistance Temperature Detector i
Sindia Sandia National Laboratories Sat Saturated i
SBR Styrene-Butadiene Rubber SCE Southern California Edison Company SECY Office of the Secretary of the Commission A-2
l
=
I l
SEP Systematic Evaluation Program SER Safety Evaluation Report
{
SNL Sandia National Laboratories SONGS San Onofre Nuclear Generating Station l
SOY Solenoid Dperated Yalve SPLB Plant Systems Branch SRM Staff Requirements Memorandum SRP Standard Review Plan SRV Safety-Relief Valve SST Steam Saturation Temperature
.z _
TDR Time-Domain Reflectometry TDS Time-Domain [ Dielectric) Spectrometry TGA Thermogravimetric Analysis TMI Three Mlle Island TM1-2 Three Mlle Island Unit 2 1
i UCS Union of Concerned Scientists i
USI Unresolved Safety Issue XtPE Cross-Linked Polyethylene XLP0 Crosslinked Polyolefin i
i
\\
k i
i i
1 a
i i
i i
I k
A-3
--o,
]-
1 1
APPDCIX 8 1
1 Chrenological Information.
f The following abbreviated chronological listing of correspondence tracks the development and resciution, through issuance of the 10 CFR 50.49, of the EQ 4
j concerns that were raised by the Union of Concerned Scientists:
{
l August 5, 1977 Staff report issued discussing results of the i
Qualification Testing Evaluation Program that was i
i being conducted by Sandia for the NRC. Problems were i
identified with the environmental qualification of electrical connectors.
i November 4, 1977 Petition filed by UCS re: fire equipment qualification issues. protection and i
The petition alleged i
l that information from recent Sandia tests, which was i
J withheld from Licensing Boards, indicated that safety equipment may fall to operate because (a) certain electrical connectors in safety-related systems cannot j
withstand a LOCA environment and (b) fires can destroy j
redundant electrical cables.
{
November 8, 1977 IE Bulletin 77-05, " Electrical Connector Assemblies.'
j was issued requesting that licensees and permit i
holders submit information to the NRC indicating the use of connectors installed inside containment i
affected by a LOCA and the documentation supporting their qualification.
November 9,1977 Coassission issued an order which directed the staff to evaluate the entire UCS petition and provide its views i
by November 25. The deadline was subsequently i
l extended and the staff's response was submitted on December 15.
l November 11, 1977 The Commission was briefed by the staff in an open meeting on the emergency aspects of the UC5 petition.
t Based on the staff's briefing, the Commission determined that no immediate remedial actions were j
required.
November 14, 1977 Supplemental IE Bulletin 77-05A was issued to all i
licensees and permit holders expanding the scope of IE i
Bulletin 77-05 to include all connectors in safety systems which could be affected by accidents other than LOCAs and to locations outside containment.
t November 22, 1977 IE Bulletin 77-06, " Potential Problems with j
Containment Electrical Penetration Assemblies' was issued requesting information; the staff issued the Bulletin in response to electrical shorts that were experienced in an electrical penetration at j
Millstone 2.
i
o i
i December 8,1977 The staff briefed the Consission in a second open j
meeting on the emergency aspects of the UC5 petition.
December 15, 1977 The staff subeltted its report on all of the issues i
raised in the UC5 petition. The staff explained the actions it had taken concerning the qualification of 4
electrical connectors, containment electrical penetrations, and other safety-related electrical.
equipment in response to the Sandla tests, recent operating experience, and the UC5 petition.
December 19, 1977 IE Bulletin 77-07 was issued requiring licensees to i
provide information concerning certain electrical 1
penetration asseabiles as a result of electrical j
shorts expertenced at Millstone 2.
l December 22, 1977 The staff briefed the Consission on its December 15 j
report in an open session.
January 30, 1978 The staff issued IE Bulletin 78-02 requiring all Itcensees to provide follow-up qualification t
i documentation for terutnal blocks.
l i
l March 23, 1978 The staff provided a suonary to the Comelssion of all actions taken to qualify electrical connectors, terminal blocks and penetrations.
April 13, 1978 The Commission issued a Memorandus and Order in response to the UC5 petition that was flied on l
November 7, 1977, with regard to environmental qualification and fire protection issues. The
}
Commission denied the actions requested by the UCS, i
but (with regard to equipment qualification) required that the staff take specific actions to further address the issue.
May 2, 1978 The UCS subaltted a petition for reconsideration re:
l EQ and fire protection issues.
I May 31, 1978 IE Circular 78-08, " Environmental Qualification of j
Safety-Related Electrical Equipment at Nuclear Power l
Plants
- was issued to highlight environmental I
qualification deficiencies and identify lessons l
learned as a result of the staff's review in response l
to concerns expressed by the UCS.
l June 21, 1978 Memorandue for L. Gossick (E00) from 5. Chilk, " Union of Concerned Scientists (UCS) Petition for Reconsideration, May 2, 1978," informed the staff that the Commission would consider the petitioner's request and requested specific staff actions.
February 8, 1979 IE Bulletin 79-01 was issued to upgrade IE Circular 78-08 to the level of a Bulletin, therefore requiring spectfic actions by the licensees.
B-2 1
I
'i March 15, 1979 SECY-79 !!!A was issued to inform the Commission of the staff's position on specific questions and concerns that were raised by the UC5 in its petition i
for reconsideration dated May 2, 1978.
1 Movember 13, 1979 Memorandum for V. Ste11o free H. Denton, forwarded the DDR Guidelines for IE to use in its review of Ilconsee j
responses to IE Bulletin 79-01.
I December 197)
NUREG-0588 was issued for comment.
January 14, 1980 IE Bulletin 79-018 was issued to expand the E4 requirements being leposed on the licensees of l
operating reactors.
February 5, 1980 Construction permit holders and appilcants for an j
operating Ilcense were notified of the staff's j
resolution of Generic Technical Activity A-24 deallag t
I with environmental qualification requirements: the addressees were informed that NUREG-0588 established the staff's position on EQ and provided specific guidance for implementation.
j January 19, 1981 Generic Letter 81-05 was issued to provide infonmation on EQ requirements that were teposed by Order in response to Itcensee requests. The specific topics that were addressed included cold shutdown, replacement parts, NUREG-0737 equipment, and the l
laplementation dead 11ne.
o
{
July 1981 NUREG-0588, Rev. 1, was' issued.
{
April 20, 1982 Generic Letter 82-09 was issued to provide 1
clarification " questions and answers" on environmental i
l qualification requirements.
l January 21, 1983 EQ requirements were codified in 10 CFR 50.49.
I i
1 i
l f
i i
I 1
1 4
8-3
i j
}
APPDSIX C l
Background Inforar. tion Related to EQ i
i i
A petition for Commission action regarding environmental qualification
}'
(EQ) of electric equipment was submitted by the Union of concemed i
Scientists (UC5) on November 4, 1977. The UCS expressed concern over the i
qualification of electric equipment at operating power reactors based on 4
electrical connector failures that were reported by Sandia National Laboratories. Sandla was performin synergistic effects testing on i
electrical connectors under the NRC s Qual'fication Testing Evaluation j
Program when the failures occurred. As a result of this initial UCS l
pet tion and other petitions that followed, such work was'done by the NRC i
staff and by the nuclear industry to demonstrate that electric equipment i
c:as qualified to function during postulated accident conditions. This appendix is a listing and descriptive summary of the information that was reviewed relative to the UCS petition and the subsequent staff actions that were taken. Potential EQ issues and notewcrthy comments are listed in bold print.
2' l
Memorandum for L. Tona from R. Fett. " Transmittal of Trio Renort."
Aucutt 5. 1977 l
Mr. Felt summartred the purpose and results of his trip to Saindia Laboratories in Albuquerque, N. M. during the period of July 27 and 28, 1977. The purpose of the trip was among other things) to discuss Evaluation Program. probless with re(gard to the Qualification
).
specific management In particular, the failures experienced in recent i
j sequential and simultaneous synergistic effects tests on connector assemblies was discussed and the failed assemblies were inspected. These tests were described in Quick-Look Reports dated January 21, March 4, and July 12, 1977. Tha failures were "...the result of exposure to the Loss-i of-Coolant-Accident (LOCA) environment and did not appear to be i
indicative of any synergistic effects." Since the connectors were
. supposed to be qualified by the supplier, Sandia also tested new i
connector assemblies supplied by the three suppliers of the faulty i
i connectors and uncovered significant design and fabrication problems with j
new connector assemblies as well.
" Union of Concerned Scientists' Petition for Fueroency and Rn.c; dial j
Action." November 4.1977 i
The Union of Concerned Scientists' petitioned the Commission to take the following actiens with regard to equipment qualification (actions pertaining to fire protection were also included in the petition).
{
i Direct the staff to accelerate a testing program for environmental
{-
qualification of connectors.
1
+
Direct the staff to independently verify the environmental qualifications of all safety-related systems, components, and 4
structures.
Notify all Licensing and Appeal Boards that no further construction u
l O
permits or operating licenses can in issued untti Applicants candemonstrate compliance with the regulations pertaining to environmental qualification.
Notify all holders of construction permits to cease all construction
=
activities involving the connectors identified as defective.
Order all operating reactors to shut down untti operators can demonstrate compliance with the environmental qualification requirements.
- l The petition was based on Sandia test results that indicated that certain electrical connectors in safety-related systems could not withstand the environmental c onditions of a LOCA. The Commission was specifically asked to take. : tion because the staff was aware of the connector problems as early as August 1977 and failed to make the infomation known to Licensing Boards, and licensees and holders of construction permits were not informed by the staff of the urgent need for corrective actions.
The petition alleged that the basic single failure criterion was not satisfied (i.e., given the failures resulting from the LOCA environmental conditions in conjunction with a postulated single failure, the safety function was not assured); qualification tests perfomed by equipment vendors failed to identify the electrical connector problems; and the staff failed to take substantive actions when it determined that the
)
qualification program of at least one reactor was unacceptable. The petition also made refersace to a letter dated July 21, 1971, from S.
Hanauer (Technical Advisor to the E00) to J. Forster (General Electric Co.) stating that the standard (i.e., IEEE Std. 322-71) that was used as a basis for judging qualification programs prior to issuance of Regulatory Guide 1.89. " Qualification of Class IE Equipment for Nuclear Power plants," was without a " single redeeming feature" and ' worthless."
i i
Note:
The message here is that using IEEE 323-71 as a qualification standard does not provide assurance of environmental qualification, which tends to conflict with the view expres' sed later by the RRAC (see the note on page C-11). Ultimately, in its Memorandte and Order of May 23, 1980, the Cossaission adopted the view that IEEE 323-71 did not provide adequate assurance of environmental qualification for safety-related electrical equipment and additional requirements were subsequently imposed on licensees via IE Bulletin 79-01/79-018 and through rulemaking.
Memorandum for L. Gossick from S. Chilk. ' Staff Reouirements - Discussion
)
on Union of Concerned Scientists Petition for @eroency and Remedial Action." November 8. 1977 In an open meeting, the Conssission discussed the " Petition for Emergency and Remedial Action" that was filed by the Union of Concerned Scientists on November 4, 1977. The Commission directed that the staff report on any matters of safety significance raised in the petition that required immediate action, and asked the staff to discuss the three specific requests for inmediate action that were raised in the petition. The C-2 i
4 staff was also requested to prepare and circulate a draft order that invited public coment on the issues raised by the petition. The Comission asked that the staff complete these actions no later than November 9, 1977.
l Memorandum for the Comissioners from E. Case. " Union of Concerned Scientists' Petition of November 4.1977. for Emercency and Remedial j
Action." November 9. 1977 j
In response to the Comission's request, the staff provided its position j
regarding the Union of Concerned Scientists' petition of November 4, 1977.
It was the staff's position, that immediate Comission action was not required to protect public health and safety. This view was based on j
the staff's perception that electrical connectors of the type tested at i
Sandia generally wcre not used in containments for safety systems that are required to function in environmental conditions which would be 4
present during a postulated LOCA.
IE Bulletin 77-05 was issued by the staff on November' 8,1977, to obtain more definitive information from operating reactors and plants under construction as to whether electrical connectors were used in safety systems within containment. Therefore, pending review of licensee responses to IE Bulletin 77-05, the staff concluded that there was no need for emergency action by the Cosmiission on this matter.
i UCS Letter Submittina a Sunclemental Affidavit of Robert D. Pollard in SuoDort of the Union of Concerned Scientists' Petition for Emercency and
{
Remedial Action." November 10. 1977 Mr. Pollard stated his position that nothing in the staff's report of l
November 9,1977, altered his conclusion that. operating reactors did not c
meet NRC safety regulations. Mr. Pollard made the following points:
The staff's view that electrical connectors of the type that failed i
generally are not used in safety systems that are required to function during a LOCA is incorrect. The staff does not know where these connectors are used in operating plants and determining where they are located would be difficult. Further complicating the issue, manufacturers assigned different model numbers to the connectors each time a new order was received.
i I
The staff is aware that three operating reactors use electrical connectors in systems that must operate in a LOCA environment but does not consider this to be a problem since the connectors were:
(a) produced by a different manufacturer than the one who produced the faulty connectors, and (b) the connectors were qualified for i.
LOCA environment. This view does not give due consideration to the fact that the faulty connectors were also " qualified" for a LOCA environment.
Since Sandia found serious deficiencies in the design and fabrication of new electrical connectors (the same as those that failed the Sandia LOCA testing) which is indicative of programatic vendor 1
C-3
/ $
~
Vii se.
4
\\
l quality assurance problems, the NRC should inspect each connector used in plant safety systems.
Other components that are siallar in design, materials, and function to electrical connectors are also suspect. For example, containment i
electrical penetration assemblies and other cable terminations (such as terminal boxes) may also be faulty.
Plants now operating were not evaluated against the current standards for environmental qualification, but against standards that are considered to be worthless by some members of the NRC staff. The l
Sandia test experience with electrical connectors show that some equipment manufactured to meet the current standards will fall when exposed to the LOCA environment. Therefore, there is reasonable assurance that equipment in operatirg plants will fall when subjected to the LOCA environment. Consequently, there is undue risk to the health and safety of the public if plants are allowed to operate.
Sunnlemental Affidavit of Robert E. Pollard. ' Union of Concerned Scientists' Petition for faeroency and Remedial Action.* November 17.
.)
121.1 Additional information and clarification was provided in the subject affidavit, strengthening the position taken by the Union of Concerned i
Scientists and further discounting the staff's view that the failed electrical connectors had no safety significance. The applicable j
regulations, a general lack of knowledge within the staff relative to specific equipment used in safety systems, and inaccurate informatic:.
L provided by the staff to the Commission on the use of electrical connectors by licensees (the staff's positionschanged from no connectors being used to connectors being used in at least 10 operating reactor plants) were discussed in the affidavit and the rationale for concluding h
that there was reasonable assurance that safety equipment would fall to j
operate during an accident condition was provided.
I Memorandum for the consission from E. Case.
- Union of concerned l
Scientists' Petition of November 4.1977. for faeroency and Remedial j
Action." November 18. 1977
)-
The staff provided additional information as directed by the Consission l
on the results of preliminary plant surveys and on specific legal i
requirements for Commission action to resolve EQ concerns. The staff completed a survey of all 65 operating units and found that eight nuclear power stations, representing 13 operating units, used electrical connectors of the type tested at Sandla ' n LOCA-related safety systses i
laside containment. The licensees involved were asked to provide i
justification as warranted to demonstrate qualification and compliance i
with the regulations.
In the case of one plant (O. C. Cook, Unit 1), the licensee took voluntary action to shutdown and remove unquallf ted connectors; approxteately 70 unqualified connectors were used inside containment in safety applications.
9 i
j C-4 i
E e
l t
Memorandum for the f-innf an fram f. Cane. *Dilan af fencerro 1
1cientists' Petition a ' November 4.197L for
'merner.cv and Remedial Action.* News ^ r 22.
977 The staff forwarded responses to the UC5 letter of November 10, 1977, and to questions raised in the November 10,1977, memorandum prepared by the Office of the General Counsel and the Office of pelley Evaluation (these enclosures were not attached to the letter and therefore were not reviewed.
on the u)e of electrical connectors.The staff aise updated the status of the pi s
Previously, the staff had identified 13 units using electrical connectors of the type tested at 4
Sandla in safety systems inside containment.
In this memorandue, the i
staff reported that additional operating plants were identified which
)
apparently used connectors inside conta'neent associated with Target Rock safety relief valves on certain SWR systems. ferther information on the 4
use and significance of these connectors was being collected by the staff.
4 Letter for the C-tsnion from Leaal Counsel for the Untan af faarer:d icientints. ' Union of Concerned Sc entlita' Petition far faa r:- v and Remedia' Action." Novemher 2L le7:
The UCS countered the legal position (point-by-point) that was stated by i
the staff in its November 18 letter en this subject. The UC5 argued that 1
the staff's legal position failed to consider the specific facts of the situation and dealt only with genera 11tles. The UCs maintained that (with regard to EQ) *...the UCS petition and affidavits of Robert pollard deal with the facts as well as the appilcable regulations,' and that the UCS han demonstrated that *... electrical connectors in safety systems inside containment are Itkely to fall in an accident.' The UC5 maintained that this was a violation of the regulations and *... clearly pose a threat to the pubile health and safety; they are not hypothetical cases.'
4 l
Meerandum for the C-instoners from f. Case. 'Use of Eigetrical Connectors in Safety Related Systent." Nov- :.ar 25. 1977 1
}
Information was provided to the Commission to supplement the information i
provided on November 18, 1977. Items discussed included use of electrical connectors in safety-related systees inside containment at Oyster Creek (nine connectors did not have qualification documentation);
4 the use of connectors inside containment on Target Rock safety relief valves used on some BWRs (information subeltted by CE indicated that j
these connectors did not need to be qualified to perfom their particular function); and the current status of D. C. Cook Unit I which was shutdown on November 18 b9cause unquellfled electrical connectors were used inside i
containment. Also on November 22, 1977, IE Sv11etin 77-06 was issued to address concerns re,garding electrical penetrations. The Bulletin was issued following reports of electrical shorts in penetration assemblies i
that were observed during normal operation at M111 stone 2.
l T
C-5 1
\\
l l
Memorandte for the C-innten from L f ate. " Union of Concerned i
scientists' Petition of November 4.1977 for fae-cency and RetPedlal i
Action.' Oecemher 4. 19 H I
Information was provided to supplement the staff reports of November 9, 1
November 18 22 and 25, 1977. Of the six operating plants at four sites that used electrical connectors inside containment D. C. Cook Unit I did j
not have sufficient qualtftcation documentation and additional j
confirmatory evalification tests were being conducted for connectors used i
i at the three Browns ferry units and at Nine Mlle point Unit 1.
The l
connectors at D. C. Cook Unit I were replaced with qualifted splices J
lactquellfled). subsequent testing demonstrated that the electrical connectors w 2
L In order to assess the extent of the eiectrical penetration probles, the
]
staff issued IE Bulletin 77-06 which required Itcensees to provide i
internation on their handling and assessment of this issue, and the staff
)
conducted a prellsinary telephone survey of the licensees regarding the information requested by the Sulletin. Of the 65 operattnq plant licensees contacted during a pre 11sinary phone survey of s' ectrical penetration assemblies, 62 indicated that their electrical penetrations j
i were qualified for the LOCA environ mnt based on their review of existing i
documentation. Licensees of the Yankee Rowe, Connecticut Yankee (Haddas j
Neck) and Dresden fact 11ttes were asked to demonstrate qualification of
)
their electrical penetrations within 10 days or justify any continued l
l operation beyond the 10-day period. Besides Millstone Unit 2 thatexperiencedpenetrationproblemswhichpromptedthestaff{theplant ssurvey),
j the only other plant to expertence electrical penetration shorting i
prcbless was Surry Unit Nos. I and 2 in 1973 (the penetrationt were replaced).
l The surveys that were conducted by the staff concerning electrical j
connectors and penetratton assemblies indicated that this equipment is geaerally of high quality and capable of performing its safety function l
l in post-accident environments. However, the staff also found that there l
were a number of operating plants for which complete documentation of the j
environmental qualtffcation tests of this equipment was not readily available.
It wat the staff's belief that sistlar results would be obtained for other safety-related electrical components located inside containment. Based on the results of the surveys that were conducted, it was the staff's belief that licensees have generally satisfied their commitments to qualify safety-related equipment and that there war, no need for the lamediate actions requested by the Union of Concerned Scientists. Scyond the question of immediate action, the staff was considering expanding the scope of its review to include (on a longer tern basts) the adequacy of environmental qualification of other safety-related electric equipment. The Systematic Evaluation Program for Operasing Reactors was being considered as a possible mechanism for expanding this effort.
The staff's evaluation of the second supplemental affidavit of Robert Pollard did not identify any other safety issues or other information bearthg on the need for immediate action.
i C-6 pumunqy r
,i2
. gm
- - - ~ ~ - - - --
~
Memorandun 'or the femalsnioners from L Cana. ' Union of concerned l
Scientists Petitton.' December 15. 1977 ii The staff's overall conclusions and recommendations concerning the UC5 i
petition of November 4,1977, were provided and specific details of the I
staff's review and supporting information was provided in an enclosure, i
- MAC Staff Report on Union of Concerned Scientists' Petition for faergency and Remedial Action.' The staff concluded that there was no lamediate safety probles with respect to electrical connectors and other j
safety-related electrical egulpeent in operating plants plants under construction, or in plants la the licensinq rav'ow proce,ss. Accordingly, j
the staff ret.oemended that the UC5 petition be denied.
With regard to qualification of electric components, on the basis of information and actions taken to date, the staff concluded that there was i
reasonable assurance that electrical connectors and containment l
clectrical penetrations would perform their required functions in the
{
LOCA environment. The staff discussed the chronology of actions that had i
been taken following receipt of the UC5 petition, provided detailed j
background information pertaining to past equipment qualification practices and the staff's application of the single failure criteria, and i
i provided the following information specifically addressing the concerns expressed by the UC5:
1 The Sandla tests (which prompted the UC5 petition) were performed to evaluate the adequacy of a testing methodolc3y and not to verify the i
qualifications of any particular electric coonent. The connector i
tests which resulted in failures were conducted to evaluate potential 4
synergistic effects. Also, contrary to what was originally intended, Sandla did not use quallfled electrical connectors in its testing j
j program and therefore, these connectors were not Ilkely to be found 4
in applications that required equipment to be qualified.
The staff had been investigating the development and evaluation of quellfication test procedures, test methods and test results since before 1968. The staff also established the Sandla Qualification j
Test Program to evaluate the adequacy of the qualification testing methodology, and a generic task action plan (A-24) was established to I
evaluate the adequary of environmental qualification of safety-l related electrical equipment.
i j
The staff was currently performing a comprehensive review of the qualification status of electrical connectors and penetrations. Of i
l the 65 operating plants, 19 were Idsntified that used connectors in safety-related systems inside containment. Qualification of the electrical connectors had been demonstrated for most of these plants, l
despite some early inadequacies in the qualification documentation, i
and qualtf tcation was demonstrated for all of the electrical l
It was the staff's belief that the results of surveys on the qualification status of electrical connectors and penetrations provided assurance that other safety-related electrical equipoent was most Ilkely C-7
k 1
)
qua11(led for their Intended functions, and the staff concluded that no i
immediate actions were warranted to address qua11fication concerns.
j However, the staff determined that, in addition to the research afforts l
that were already in progress, additional action was required to assess I
the current status of qualiffcation for other equipment (i.e., not just electrical connectors, terstnal blocks and penetrations). Therefore, it was the staff's intention to review the entire matter of electrical equipoent qualificat1on durIng the $ystosatIc Evaluation Progras j
init1ative.
i Memorandum for the C-Ission from K. Case. "Unfon of Concerned
$cientists' Petition - the use of f actrical connectors in safety
$vstems.* Janaary 6. 1978 4
i j
The staff provided an update on electrical connectors in use at Pilgrim i
l' nit 1.
The staff determined that additional formal documentation was j
required from the ifcensee to demonstrate qualification of 35 electrical 4
connectars vsed in safety-related systems inside containment. Since the plant was operating at reduced power (505), the connectors were housed in j
steel boxes, tests of stellar connectors indicated that they would remain j
operable during at least the inttf al phase of a LOCA, and sufficient safety and non-safety equipment existed to sitigate the consequences of a LOCA in the event that the electrical connectors were to fall, the staff agreed with the Itcensee's position that there was no need to require an lamediate plant shutdown. However, the licensee agreed to resolve any l
remaining staff concerns relating to the electrical connectors prior to returning to power operation foi owing a scheduled January 21, 1978 maintenance outage.
j Letter for the Commission from the Union of Concerned Scientists. " Union i
of Concerned Scientists' Petition for [mercener and Rennedial Action."
January 9. 1978 The Union of Concerned Scientists submitted a Draft Memorandum and Order for consideration by the Commission. The UCS was critical of the NRC staff in its response to the electrical connector issue, stating in the Draft Memorandum and Order that *...the facts are that 30% of the operating reactors have connectors required to survive a LOCA and main steam line break, and the Staff was unaware of this fact.
In addition, i
no licensee had available full documentation for the qualification j
testing, under the most adverse design conditions, for those l
cc,nnec to rs.... '
Memorandum for the Commission from E. Case. " Union of Concerned Scientists' Petition.' January 13. 1978 The Commission was informed of the results of the review that was j
performed by the Office of Inspection and Enforcement (IE) of licensee
)
respenses to IE Bulletins 77-05/05A and 77-06 concerning electrical j
connectors and electrical pen 6tration assemblies, respectively.
IE concluded that the responses either provided or referenced information
=
i which indicated that adequate qualification test results existed to j
support continued operation of all operating reactors, with the exception j
i C-8 n
t
=
of Pilgrim and Connecticut Yankee (19 reactors had reported the use of i
connectors in safety systems inside containment.
Connecticut Yankee chose to replace the four con)nectors with qualifled The licensee for terstnal blocks inside sealed junction boxes. As a result of unfavorable environmental qualification screening tests, the Itcensee for P11gris began an orderly shutdown on January 9,1978, to resolve the issue.
licensees for Browns Ferry Unit Nos. 1, 2, and 3 and Nine Mlle Point were The perforsing additional testing which was scheduled for completion by the end of February.
appitcable to Maine Yankee,The results of the Nine Mile Point tast would also be testing to be completed by the end of February. Oyster Creek was also conc'u The IE report on connectors also stated that Itcensees have not defined specific environmental conditions with regard to accidents other than LOCA; e.g., a MSL8.
the generic subject of equipment qualification as described in TaskIt w Action Plan A-24, " Qualification of Class IE Safety-Related Equipment."
I i
The IE report on electrical penetration assemblies concluded that the i
penetrations used in operating reactors were environmentally qualtfled
)
for the LOCA condition, based on IE review of Itcenses qualification test reports and comparattre design analysis.
The staff's review included the qualification tett re Amphenol Sams, Conax, ports on penetrations designed and manufactured by i
l Viking, and Westinghouse.Crouse Hinds, D. G. O'Brien, General Electric, The NRC staff considered temperature, i.
pressure, humidity parameters of conce,rn. leakage rates, and seismic conditions to be the by evaluation of materials used, while others had specific radiationR i
exposure testing data for the assembled unit.
Although NRR was still pursuing the question of pressurization, IE recognized that the test data j
verified that the penetrations were qualified with and without nitrogen 1
pressure.
I j
Memorandum for the Commission from E. Case. ' Union of Concernej j
Scientists' Petition." January 20. 1978 i
The staff provided its response to the Commission on the ' Draft l
Memorandum and Order" that was filed by the UCS on January 9,1978.
i was tl.a staff's view that no new facts were presented by the UCS in its It t
subetttal, and various shortcomings of the UCS perspective were discussert.
In particular, the staff pointed out that the UCS had not i
1:ade a connection between the Sandia test condittor.s and the cond j
that are assumed to exist during an accident, and contrar j
i were not qualf fled to IEEE 323 (with one possible exception).
{
by the staff concerning the use of electrical connectors ove It was 4
i due to the staff's need to promptly respond to the UCS petition and to keep the Comission informed, coupled with the evolving nature of the information being submitted by licensees as more in-depth, follow-on 1
1 reviews were completed.
the staff maintained that adequate qualification of connectors inBased o operating reactor plants had ultimately been demonstrated despite i
4 C-9
i inadequacies in the supporting qualification documentation.
The staff recommended that the Connission not adopt the Draft Memorandum and Order proposed by the UCS.
Letter for the Connission from the Union of Concerned Scien".ists. " Union) of Concerned Scientists' Petition for Emeroency and Remedial
]
January 20. 1978 Action."
The UCS urged the Coamiission to act on its petition of November 4,1977, 4
and that it would be a *... grave sistake for the Cosnission to believe i
that this problem has gone away by virtue of the Staff's actions...."
As a result of recent electrical connector test failures and the subsequent licensee decision to shutdown Pilgris Unit 1, the UCS concluded that 1
...the staff's filings with the Consission consistently overstated the extent to which known and verified facts supported its conclusion that, despite inadequate documentation, electrical connectors in operating plants are qualified te withstand accident conditions."
The UCS supplemented its conclusion with the following additional assertions and observations:
The staff has given the false impression that information submitted by licensees concerning the use and qualification of electrical j
8 has been reviewed and verified by the staff to be correct.
cou
}
i tW.
. owed a new and apparently ad har justification for 1
i e
i d Nw ; eration of a non-complying facility in that the use of i
Nt >
quipment was credited contrary to Standard Review Plan 9-i The Pilgrim incident suggests that the actual situation at other
{
fact 11 ties is equally uncertain (i.e., there is no reason to believe that the facts are any better for other facilities than they were for l
Pilgrim).
Although the staff has indicated that partial qualification testing 1
i has been completed for several plants, it would be more accurate to j
say that those connectors are unqualified because certain basic parameters such as aging and radiation were not included in the testing.
The staff does not have a ba;is for concluding that electrical connectors will survive a steam line break inside containment, and connector qualification for accidents outside containment has not 1
been adequately reviewed by the staff.
In the case of Quad Cities, the staff has allowed the ifcensee to credit failure modes and the existence of diverse protective features i
l for mitigating electrical c]nnector qualification problems.
This approach is contrary to the established requirements and the guidance of the Standard Review Plan.
The problems and test failures that were experienced with electrical i
l C-10 1
_,.,4
l equipment. connectors raises doubts as to the adequacy of quellfication of other No detailed basis has been offered by the staff to support its conclution that electrical penetrations are "...quallfled to function as (, oded during LOCA conditions...." and the staff. does not addrei the qualification of penetrations for accident conditions other than for LOCA.
I i
Enn though the epoxy compound used for potting the electrical penetrations on the Lacrosse plant would be exposed to post-LOCA temperatures greater than its rated temperature, the staff concluded i
that the penetrations would maintain their integrity (the higher 1
temperatures of a main steam line break were not even discussed).
Issue:
i Epoxy compound used for potting electrical penetrations l
may not be qualified to the temperature conditions that i
are experienced post-LOCA and/or during a MSLS.
There is a consistent pattern of tM staff's refusal to consider the i
broad impliertions of the test fat'. ares, of totally inadequate 1
quality assurance documentation and of qualification testing wh3ch (when performed conform to curre)nt standards. frequently omitted basic parameters and did not Memorandum for the Cggifslon from E. Case. ' Union of Concerned i
Scientists' Petition." February 10. 1972 The staff provided a summary status on the on being conducted by licensees for Browns ferry. going qualification programs Units 1, 2, and 3 Nine Mile Point, and Oyster Creek. Based on the documentation that had been provided, the staff felt that there was a high likelihood that the confimatory tests would be successful.
The staff also provided clarification on certain issues raised in the UCS letter of January 20, i
1978.
In particular, the staff asserted that:
j (1) the staff's actions with regard to Pilgria 1 had been sufficiently discussed in the previous staff filings, and (2) the Quad Cities temperature detectors were not i
required to be qualified for a steam line break in the HPCI roos,
}
contrary to the position taken by UCS.
i Memorandum for the Commission from E. Case and E. Voicenau. " U Concerned Scientists' Petition." March 23. 1978 1
Information was provided to the Commission on being performed by licensees on electrical conn (ec) tors, penetrationthe status 1
i assemblies, and terminal blocks; (2) the staff position on aging t
1 requirements; and associated with qua(3) The Office of Inspection and Enforcement activities i
lification of electrical connectors.
j Regulatory Guide 1.89 was developed by the staff to endorse IEEE 323-74 l
for environmental qualification reviews perfomed during the licensing process.
The significant changes in this standard relative to i
IEEE 323-71 concerned the establishment of a ' qualified life" and i
C-Il
~~ ~'
~
~
~
o.
- \\
radiation qualification.
The Regulatory Requirements Review Committee (RRRC) reconnended that the guide be applied only to future construction permit applications.
That decision reflected the position that '...the incremental taprovements were not significant to safety and that full taplementation of IEEE 323-74 required further development of other ancillary standards to provide guidance on specific safety-related electrical equipment and components."
Note:
This RAAC position tends to confilet with the view that was expressed earlier (see the note on page C-2).
j Based on current licensing experience, the staff also recognized that additional guidance was needed in the area of accelerated aging techniques used to establish a qualified life for electrical equipment
)
and assemblies.
Task Action Plan A-24 and the NRC research program were expected to develop this additional guidance. Applicationsfor j
construction permits flied since issuance of Reguiatory Guide 1.8g in i
1974 were made subject to the " qualified life' provisions of IEEE 323-74, j
but the staff recognized that there continued to be a need for development of test methods in this area (hence, the high priority of this subject area in the NRC research pr> gram at Sandia).
l Memorandum for the Ca==issioners from E. Case. ' Union of Concerned Scientists' Petition." March 30. 1978 j
j Additional information was provided to the Commission re: unqualified electrical connections insida containment in safety-related systems at D. C. Cook and Haddam Neck.
The problem involved terutnal blocks inside metal boxes.
At D. C. Cook, the terminal blocks were replaced with qualified splices.
Unquallfled Narathon teminal blocks at Haddam Neck l
were replaced with qualified Westinghouse teminal blocks.
[
Westinghouse terstnal blocks (also used at Yankee Rowe and Ginna) hsdHowever, failed recent testing at Franklin Institute. The failure mechanise appeared to be stress related screw and/or differential expan(excessive torque applied to one mounting sion between the aluminum enclosure and l
the terminal block).
1 l
Connission Memorandum and Order dated Anril 13. 1978 i
As a consequence cf the UCS petition, the staff took actions to specifically address environmental qualification of electrical connectors, terminal blocks, and penetrations. The Commission generally agreed with the staff's actions in this regard, but was critical of the staff's handling of P11gris.
In this case, the itcensee had initially i
provided erroneous information regarding the use of electrical connectors j
and obtaining complete information from the licensee was delayed well j
br m d the requested response date.
The licensee ultimately shutdown Pilgria in order to replace unqualified electrical connectors.
The Connaission's view was that the staff's actions in obtaining and reviewing the Pilgrim documentation was unsatisfactory. The Commission also noted l
that the NRC is dependent upon the licensees for accurate and timely information, and that some licensees failed to exhibit a detailed knowledge of the quality of installed plant equipment. The Commission i
C-12
,,,a, me,-u*
~ '
^~
~
~
o i'
l Indicated that Itcensees must have a "... detailed understanding of their own plants in order to satisfy their obligations for public safety by ensuring a sound basis for making assessments of plant safety....The licensees must be knowledgeable and vigilant and must take more initiative in ferreting out details of potential plant weaknasses.*
Consequently, the staff was asked to consider the need for further regulatory actions to include a possible NRC policy statement to re-q emphasize the important safety responsibilities of licensees.
In its decision, the Comission stated that the adequacy of environmental qualification of all Class IE electrical equipment connectors, penetrations, or terminal blocks), will(be confirmed for all not limited to operating plants as a first priority matter in the NRC Systematic 1
Evaluation Program (SEP).
actions with regard to equipment qualification:The staff was directed to take the fol 1
)
Perform additional testing of electrical connectors to obtain data for verifying the current methodology for environmental qualification i
of electrical components.
4 These tests should be performed with a i
representative sample of comercially available electrical connectors qualified in accordance with IEEE Std. 323-74 and in use in nuclear power reactor safety systems.
Review the procedures by which (1 offices, and Licensing Boards are) notified of research infonnationthe Co 1
which is of safety significance, and (2) follow-up actions are taken with licensees and applicants.
estimates 'of resource requirements and poter.tial benef i
conducting independent verification testing of environmentally qualified equipment which is required to operate in safety systems.
Conduct a comprehensive ' lessons learned" evaluation to include the 4
following:
(a) review all licensee responses attention to the Pilgrim case), to determine co(nformance withpaying particular applicable quality assurance documentation requirements, as well as the accuracy and timeliness of information provided (where justified, appropriate enforcement actions should be taken ; (b) review how electrical equipment, not fully qualified, came)to be installed in those plants where found; case so that similar delays (c) review staff actions in the Pligrim i
may be avoided in the future; and i
NRC policy statement to re-emphasize the important safet sible responsibilities of licensees.
Inform the Comission of the results of the staff review of further i
qualification testing by licensees for which fully documented test results are not yet available (Browns Ferry 3, Nine Mlle Point, and Haddam Neck).
Inform the Comission of the decision made on the question of whether i
{
C-13
i 4
i nitrogen gas will be required for those centainment penetrations which can accoanodate such pressurization.
}
Review the results of the first phase of the Systematic Evaluation
=
Program concentrating on the sty adequacy and environmental qualification of all Class V ectrical equipment. Provide recommendations on whether plants.
its review needs to be extended to other Memorandum for L. Gossick from Comissioner Bradford. " Staff Actions in
)
Resoonse to Deficiencies in Connecto-Solices." Anril 26. 1978 Comissioner Bradford made reference to PN-78-91 which reported that i
approximately 100 splices inside the Monticello containment did not have qualification documentation.
Comissioner Bradford questioned if this tas an industry-wide problem similar to what was experienced with electrical connectors and, if so, what course of action was the NRC 4
i taking to avoid the ' problem" that arose while trying to resolve the electrical connector qualification problem.
Memorandum for the Comission frou E. Case. ' Commission's horandum and Order fof_Aoril 13. 19781:
Action." May 4. 1978 UCS' Petition for Emeroency and R m dial The staff described the status and/or actions that were being taken for i
sach of the ten required staff actions listed in the Order, two of which j
were related to fire protection.
regarding the equipment qualification action items:The following information was pro i
j Perform Additional Dualification Testino of Electrical Connectors The staff planned to perform the tests requested by the Comission in two steps:
(a) electrical connectors qualif ted to IEEE 323-71 would be tested for LOCA conditions, thereby allowing the staff to assess the adequacy of electrical connectors currently in use without subjecting them to conditions for which they were not qualified; and (b) electrical connectors qualified to IEEE 323-74 would be tested, including main steam line break environmental conditions.
i i
Review Notification Procedures for Informino the Comission. Staff Offices, and Licensino Boards of Research Information. and for
{
Pursuino Follow-un Actions with Licensees and Aeolicants The procedures outlined in SECY 78-27 currently being considered by i
the Comission were proposed for notifying licensing boards of important research information and existing procedures were credited i
for notification of the Comission and appropriate staff offices.
Therefore, most of the staff's efforts would be placed on developing procedures for taking follow-up action with licensees and applicants.
Analysis of Alternatives for Conductino Indeoendent Verification Iestino of Environmentally Oualified Eouicment i
The staff had initiated action to develop a plan for completing the l
C-14 l
- q
~
required analysis. A task group under IE leadership, with accomplish and oversee this work. representation from other offices, Since IE was already evaluating the feasibility of increasing independent verification in all areas of the inspection program, this effort was viewed as an extension of the assessment that was already underway.
Conduct a Comorehensive lessons tearned Evaluation Much of the effort directed by the Commission was well underway as a consequence of actions taken b i
petition for remedial action. y the staff to respond to the UCS Action to initiate generic comunications reatnding licensees of their responsibilities for providing prompt and accurate information to the NRC was to be censidered as part of this effort.
The staff planned to issue a
- 1cssons learned" report to the Comission by June t
13, 1978.
Practices for WRC-Soonsored Confi matory Research Pro The Office of Research was assigned lead responsibility for j
completing this action, with assistance from other program offices.
documented assurances of the appropriate nuclear g those test items which will be related to actual counterparts in nuclear power plants."
}
Inform the Commission of the Results et j
Dualification Testino Beino Performed by Certain Licenseestne Staff Rey!ew of Fu i
the plants specifically referenced (i.e., Browns Ferry Point, and Maine Yankee).
k i
Jnfem the Commission on Whether Nitrocen Cas is Recutred for Containment Electrical Penetration Assemblies i
Based on the staff's review, no technical basis was found for requiring nitrogen pressurization of electrical penetration assemblies.
4 Dualiff eation of Electric Eouioment and Prov Program dealing with environmental qualification of all was in preparation. electrical equipment was essentially completed and the l
I Memorandum for the Commission from E. Case. " Union of Concerned Scientists' Petition." May 12. 1978 The staff provided the results of its short term assessment of the i
C-15
I, i
l environmental qualification of safety-related electrical equipment of the i
eleven operating reactor facilities in the Systematic Evaluation Program (SEP) (published as NUREG-0458 in May 1978). The staff's review was completed as directed by the Comission in its Memorandum and Order of April 13, 1978 (Appendix II, Item 10).
The staff did not identify any significant safety concerns that required imediate remedial action for these facilities.
Since the SEP facilities included eleven of the older i
of environmental qualification for safety-related electrical equ the staff felt that this conclusion could be generalized to all operating reactors.
However, due to the problems that were identified relative to electrical connectors, the staff felt that it would be prudent to highlight this experience in a Circular to licensees of all power reactor
{c 1
fact 11ttes (IE Circulst 78-08 was issued on May 31,1978).
The staff also planr.ed to initiate an augmented inspection effort as part i
of the normal NRC ar.tivities, concentrating on inspection of installed i
safety-related electric equipment and on an audit of the records for i
environmental qualification.
tts review and evaluation of safety-related electric equipment for theThe sta
}
eleven SEP facilities on a detailed plant-by-plant basis as part of the i
overall SEP effort, and indicated that any significant safety concerna j
integrated into NRC licensing decisions and actions. identified dur j
Memorandum for L. Gossick from S. Chilk. ' Union of Concerned Scientists ui j
J,UCS) Petition for Reconsideration. May 2. 1978.* June 21. 1978 j'
j At the May 31, 1978, Comission meeting, the Comission decided to i
reconsider its April 13 decision on the earlier UCS petition of j
November 4, 1977, as requested by the UCS.
The staff was asked to pruvider a complete and objective assessment of tha UCS contentions and in '
particular, to respond to the following environmental qualification j
issues / questions:
Identify any plants which cannot demonstrate environmental qualification for electrical connectors, splices, penetrations and terminal blocks with full doiumentation, and explain the le regulatory basis for permitting their continued operation. gal and Did the applicable environmental qualification regulations for all operating power plants specify actual testing prior to granting of an i
operating Itcense?
(The UCS petition of November 4,1977, specifically referred to GDC 3 and 4 of 10 CFR 50 Appendix A, Criterion !!! of 10 CFR 50 Appendix B,10 CFR 50.55a(h), and to the basic single failure criteria of 10 CFR 50 Appendix A, as the applicable requirements.]
i L
UCS asserts that the Comission concluded that no violations of HRC
=
regulations were presented in spite of the fact that (a) licensees failed to meet the comitments made by them in SERs and required by i
the regulations to environmentally qualify equipment, (b) licensees failed to perform tests to qualify equipment prior to operation, and C-16 i
I
~
(c) equipment was installed that actually failed when testing was finally performed.
Coment on these asserted " facts."
i Is there anything in the UCS attachment entitled " Chronology and Analysis of Staff Actions" that was not specifically brought to the Comission's attention by the staff prior to the issuance of the April 13 Memorandum and Order?
Review again the UCS ' Draft Memorandum and Order" of January 9,1978, and respond to the allegation that it was " virtually ignored."
Did the Comission express the correct formulation for use of the single failure criterion as it applies to the licensing review process in its April 13 Memorandum and Order, or is it just a
" theoretical formulation" as asserted by UCS7 As to those plants that were ultimately shut down as a result of the investigations that were performed by the staff in response to the November 4, 1977, UCS petition, is there evidence that the licensing reviews failed to ensure that CDC-4 was met and/or that quality assurance requirements were inadequate to detect noncompliance with 1
GDC-47 i
Provide views on the UCS characterization of the single failure criterion as stated in the UCS Petition for Reconsideration.
Provide the staff's views on the allegation that "there is now reasonable assurance that most, if not all, plants in operation use i
equipment which will fail when exposed to design basis event conditions."
Evaluate and discuss the request to issue orders to all operating plants similar to the order issued to Indian Point 1, requiring backfit of regulations for environmental qualification, fire l
protection, and the single failure criterion.
Provide the rationale for not backfitting IEEE Std. 323-74 to all operating plants.
Also provide the record of the Regulatory Requirements Review Comittee's determination that this standard was not to be backfitted.
Does the staff agree with the UCS assertion that "...without the motivation of the UCS petition, the staff would not have uncovered the disturbing facts which have since come to light?" Provide a listing of the staff actions completed or underway in response to the Sandia test results, prior to receipt of the UCS petition.
Provide a full chronology of the D.C. Cook Unit 2 situation, specificLily addressing the statements on page 17 of the Petition for l
Reconsideration.
Provide the staff's assessment of each of the UCS ' Analysis of Staff Action" sections, including an assessment of UCS contentions, 1
C-17
E quotations of previous staff submissions, and representation of the sequence of staff actions.
i -
Discuss the UCS assertion that the staff completely reversed itself on the qualification of terminal blocks at Connecticut Yankee.
i,
e Respond to UCS statements that the staff's January 27 submission to the Commission was " untrue" as to its promise of complete review of l
qualification data before Connecticut Yankee would be permitted to j
resume operation.
Respond to the UCS assertion that promises made by the staff and the licenses were not fulfilled with regard to connectors and j
penetrations at the Pilgrim plant.
Discuss the UCS assertion that the staff has attempted to Ilmit the applicability of the Sandia test failures.
i j
How does the NRC know whether installation is completed so as to preserve the capability of electrical equipment to survive an j
accident environment?
i Issue:
Installation and maintenance practices say invalidate EQ.
Respond to the UCS contention that the staff is inclined to excuse i
serious deficiencies by 392 has rationalization, and that the testing of Pilgrim spilcos did not determine their ability to survive i
a LOCA.
1
}
SECY-78-348. " Evaluation of Resoonded to IE Bulletins 77-05 and 77-05A.
Staff Actions Related to Bulletins. and Reisted Quality Assurance Entters." June 26. 1978 The staff provided the Comission with the results of its comprehensive
" lessons learned" evaluation relating to the qualification of electrical
- i connectors.
The staff's review was completed as directed by the 5
Commission in its Memorandum and Order of April 13, 1978. The staff's i
evaluation contained the following information:
IE Bulletins 77-05 and 77-05A were sent to 41 utilities operating an 4
aggregate of 65 reactors.
The responses to the bulletins indicated j
that 13 utilities had installed electrical connectors in the safety systems of 19 reactors.
Seven of the 13 utilities affected, representing an aggregate of ten 1
reactors, had electrical connectors for which qualifi ntion to i
regulatory requirements could not be demonstrated.
Ite ten reactors i
included D. C. Cook 1 and 2, Haddam Neck, Oyster Cre2L. Nine Mile j
Point 1, Browns Ferry 1. 2, and 3, Pilgrim 1, and Maine tinkee.
Subsequent testing drw nstrated that the electrical connectors for Oyster Creek, Nine Pile Point 1. Browns Ferry I and 2, and Maine Yankee were qualifie for use in safety systems.
C-18 h
e m
The electrical connectors at D. C. Cook 1 and 2 were qualtfled for LOCA but not for MSL8; the electrical connectors were replaced.
The electrical crnnectors at Haddam Neck and Pilgrim were replaced without performing any additional qualification testing.
The electrical connectors at Browns Ferry 3 were qualifted, but additional potting was required; further testing of the "unpotted" connecters was not performed.
D. C. Cook I was cited for not being in compliance with its quality assurance program. However, a response to the citation was not required. Conformance with quality assurance requirements was also investigated at Pilgrim 1, with no noncompliances identified.
i Ccnformance of the other utilities with quality assurance requirements was not pursued because of the generally undeveloped l
nature of the regulations in this area.
Information submitted for Haddam Neck and Pilgrim I was inaccurate, but enforcement action was not taken since the information was preliminary and was subsequently corrected.
Although D. C. Cook 2 had not yet been operated at the time of the staff's evaluation, enforcement action was taken for material false statements in the D. C. Cook 2 license appiteaticin.
Installation of electrical equipment that was not fully qualified was due in part to a lack of development of regulatory requirements.
The ten affected reactors were originally docketed between 1963 and 1967.
Construction permits were issued between May 1964 and March 1969, and operating licenses were issued between June 1967 and December 1977.
Although there were general requirements that equipment should be quallfled for the intended service, the r,neral Design Criteria were i
not established by regulation until Febrary 20, 1971 (two years after the latest construction permit was issued for the ten reactors i
a f fected).
Also, the formal quality assurance regulation (10 CFR 50 i
Appendix B) was not established until June 1970.
Thus, none of these plants were expected to have fully implemented quality assurance programs during their construction phases.
As with the regulations, the staff review process for nuclear power plant appitcations was also under continuous revision.
The Standard Review Plan (SRP) was first published in final form in November 1975.
lihile the SRP had been published for cosseent and was being used in varying degrees as early as 1974, even the final version did not explicitly address small electrical components or connections such as connectors, terminal boards or splices. The inspection program also did not address these compon,ents, but focused on motors, instruments, and cable qualificatfor..
i In general, the staff found that license application statements regarding qualification of electrical components were vague or nonexistent.
These findings were considered to be consistent with the status of industry standards and NRC requirements that were in existence at the time, C-19 I
i i
The staff has recognized that, as a result of the evolving regulatory requirements, issues like the environmental qualification of connectors may need attention on reactors already Ifcensed for operation, The staff has ongoing programs to address safety levels j
in Itcensed plants in areas where the regulatory requirements have changed.
t developed to address the evolution of Itcensing criteria on olde 6
i operating facilities.
Current reviews of Ilconse amendments on operating reactors i
frequently address issues where the licensing criteria have evolved.
)
In these cases, the IE and NRR staff assure adequate safety levels 1
l are maintained.
1 i
Based on its evaluation, the staff concluded that coordination and communication between NRR and IE and among IE organtrations could be i
improved.
The staff did not find that licensees had intentionally misled l
the Connission.
Based on these conclusions, the staff planned to:
implement practical improvements in interoffice communications; (b) (a) assess the need for a priority Bulletin procedure; (c) send a letter to all reactor licensees and semit holders stressing the importance of the accuracy of information sumitted to the NRC, both in applications and in response to Bulletins; and (d) assess the enforcement capabilit to the accuracy of infomation submitted in Bulletin responses.y related IE Circular 78-08 was issued on May 31, 1978, discussing the various environmental qualtftcation deficiencies that were identified and other j
" lessons learned."
Memorandum for the Commission from E. Case. " Union of Concerned Scientists' Petition for Reconsideration Dated May 2.1978.* July 6.1978 The staff provided its response to the asterisked items that were Identified in the SECY memorandum of June 21, 1978. The staff indicated that the response to the other items would be provided by August 25, 1978.
information was provided:In the area of environmental qualification, the following l
Regarding the allegation that "...there is now reasonable assurance that most, if not all, plants in operation use equi i
fall when exposed to design basis event conditions.pment which will j
..." the staff continued to believe that there was adequate protection for the pubite health and safety.
The staff referenced its memorandum of December 13, 1977, which previously discussed this issue and its i
memorandum of May 12, 1978, concerning the SEP program as, the basis i
for this position. However, the staff recognized that additional
{
effort was needed in the area of environmental qualification.
j Ongoing staff actions that were specifically mentioned included J
Circular 78-08 and associated IE inspection activities; Technical Activities A-21 (Main Steamline Break Inside Containment) and A-24 i
which was addressing EQ on eleven of the older opera) tin
{
and confirmatory research programs in the area of synergistic effects, aging and radiation source ters.
i l
C-20 l
De Specific actions taken by the staff and licensees following issuance of the Cosalssion's Memorandum and Order of April 13, 1978, to upgrade and/or satisfy environmental qualification requirements for specific power plants were discussed. For example, some plants were conditioned to require sequential testing of certain Barton instruments, changes were made to terstnal blocks and cable questionablespliceswerereplaced,andendsealsonmineraI insulated cable were modified to prevent moisture penetration during LOCA conditions, i
Memorandum for the Commission from H. Denton. " Union of Concerned Scientists' Petition for Reconsideration Dated May 2. 1978.* Auaust 31.
121A The staff provided its response to the various points of contention that were raised by the UCS as requested in the SECY semorandum of June 21, 1978.
The staff concluded that the Petition for Reconsideration contained no new facts pertaining to the issue of environmental qualification and therefore, the staff's position on the UCS petition remained unchanged from what was presented previously in its December 15, 1977, report on this subject. Nonetheless, the staff acknowledged that its recent expertence with environmental qualification over the past year has shown that additional staff attention was needed in this area.
It was the staff's view that, In a more general sense (as evidenced by the short-ters environmental qualification assessment of the eleven SEP plants), no significant safety concerns that required immediate remedial action were identified either by Itcensees or by the staff.
This peneral conclusion was also thought to be appitcable to the other operating power i
reactors.
The following information was extracted from the staff response:
l 8ased on its review and follow-up activities associated with IE i
Bulletins 77-05,77-05A, 77-06 and 78-02, the staff concluded that qualification documentation demonstrated adequate qualification of connectors, penetrations, and unprotected terminal blocks for LOCA conditions inside cor.tainment.
The adequacy of environmental qualtf tcation of splices, protected terminal blocks was being pursue,d during staff follow-up of IE Circular 78-08.and o The NRC staff was pursuing the broad question of environmental quaitfication of Class IE equipment, including protected teminal i
blocks, at operating reactors via (a) the SEP review being conducted for older plants and (b) by performing follow-up inspections of actions being taken by licensees in response to IE Cirevlar 78-08.
i While the Itcense reviews of environmental qualification have not been instituted on a plant-by-plant basis (except for those in SEr),
the IE follow-up inspections to Circular 78-08 were expected to i
verify that licensees have initiated a reexantnation of the l
environmental qualification of electrical components at their C-21 l
t l
4 fac111ttes. This reexastnation should address the problems identified in the NRC revfew of environmental qualification.
's The staff recognized that additional guidance was needed in the area of accelerated aging technic'ves used to establish a qualiffed life i
for electrical equipment and assemblies. The Category A technical activity on equipment qualification (Task Action P' an A-24 and the NRC research program were intended to provide additional gu)idance for the development of test methods and licensing review procedures on aging.
Applications for construction penetts filed since the issuance of i
Regulatory Guide 1.89 in 1974 were made subject to a requirement for j
documentation (at the operating Ilconse stage of review) which shows that a qualifted 11fe for safety-related equipment has been established in accordance with IEEE 323-74.
i i
Although staff actions were initiated as a result of the Sandla electrical connector test failures before the UC5 petition was flied, the staff's efforts were intensified and accelerated by the interest and concerns raised by the petition.
The Sandia electrical connector tests were ' overly conservative
- in that Sandla used the latest version of the IEEE t.est standard (i.e.,
i IEEE 323-74), whereas all operating plants had their connectors qualif ted to the eariter, less severe, standard f.e., IEEE 323-71).
NRA and IE staff concluded that failure of the sp(ecific connectors at i
Sandla was not proof that the connectors in operating plants would j
not survive a LOCA.
However, an IE Bulletin (IES 77-05/77-05A) was issued to request that Ilconsees consider,the generic tap 11 cations of the connector test failures and review their spectfIc qualification i
test data for any problems.
1 Note:
This logic seems to suggest that the IEEE 123-74 standard j
any be overly conservative.*
i At the time the electrical connectors were procured by Sandla for l
testing, none were available that were qualtfled to the later version of the standard and therefore, a procurement strictly in accordance j
witi Appendix B requirements was not possible, i
The staff's conclusions for each facility that the actual installation of equipment has preserved the capability of electrical equipment to survive an accident was based on positive findings with i
regard to:
(a) licensee commitments to meet ' appropriate standards for design, fabrication, testing, and installation;" (b) management and quality controls established by Ifcensees and their contractors to assure that the intended quality standards were met; and (c) NRC i
audits to verify, based on a sample, that licensee and contractor
{
control systems were working.
1 C-22 i
=-- - _. - _ _ -
)
SECY-79sll2A. ' Union of Concerned Scientists' Petition for Reconsideration - Environmental Oualification of Electrical Ecutoment."
March 15. 1979 The staff provided its response to specific questions and concerns that were raised by the UCS in its petition for reconsideration of May 2, 1978.
In the response to Question 6 the staff indicated that test results did not exist for many compon,ents, nor were they required at the time of initial Itcensing of the Cthen currently operating reactors.
Although the staff felt that the ievel)of assurance that equipment was qualified was adequate in the short ters, it was desirable to increase the level of assurance for the future. The Systematic Evaluation Program was an important part of the ongoing staff actions in this regard and, as a result of that effort, the staff instituted a program of augmented IE inspections and issued IE Circular 78-08 to provide feedback to licensees on lessons learned and to initiate actions by licensees to examine the installation and qualification documentation for squipment located inside containment.
The staff ultimately requested that the eleven SEP
[
licensees identify the specific conditions and methods used to qualify each piece of equipment.
Type testing of equipment installed in SEP facilities was not performed to the extent that was used in the more recently Itcensed facilities, and the major staff effort under the SEP i
1 program was being directed to more precisely defining adequate methods of environmental qua11ficat1on.
Memorandum for L. Cossick from R. T. Kennedy. "UCS Petition for Reconsideration." Auoust 7. 1979 Comissioner Kennedy requested that the staff assess the effects of backfitting the 1971 and the 1974 versions of.IEEE Std. 323 on licensees.
Specific questions in this regard were asked, and the staff was asked to provide its response no later than August 21, 1979.
dv from H. R. Denton. "UCS Petition for e n de at on p,
The staff provided its response to questions that were raised by the j
Commissioner on August 7.
The following infomation was provided:
{
The staff provided a comparison of IEEE 323-71 and IEEE 323-74 (see Table I-1).
The principle differences that were cited between the two was that the 1974 standard provided much more detail and included requirements for aging, margins, and documentation that were not i
included in the 1971 standard. However, both standards required guidelines for necessary interpretations and judgements.
The staff noted that current reviews of plants referencing the 1971 standard were being conducted so as to bring the level of assurance of i
equipment qualification in these plants to essentially the same level as using the 1974 standard would achieve.
i Note:
This does not seem to be a fair assessment of the staff's review efforts in that equivalent aging, margin, and qua11 fled life requirements equivalent to those required C-23
i e
i I
i
(
by IEEE 323-74 were not imposed on those plants that i
referenced IEEE 323-71.
Fifty-eight of the 70 power reactors that were currently licensed to operate (including Indian Point I and Humboldt 8ay) made no specific i
i reference IEEE Std. 323-1971 as the basis for equipment qualification. The licensees for the remaining plants had i
commitments in place to comply with the standard, t
The 00R Guidelines that were being developed by the staff will provide a level of confidence essentially equivalent to that which would be achieved from the appiteation of IEEE Std. 323-74.
- However, I
the Guidelines will only require that aging be considered for that i
equipment that has been identified as being susceptible to significant aging effects. As the staff's understanding of aging j
effects improves, the need for further backfitting of the aging j
requirements will be reassessed.
Advantages that could be realized by backfitting the 1974 standard would include a more rigorous and complete demonstration that aging effects are adequately accounted for, including establishing a qualified life for all safety-related equipment. Further, margins l
would be applied te all test parameters whereas the Guidelines will.
i require only tnat margins be applied to the most significant parameters ie.g. time).
A significant disadvantage is that many Itcenster would be required to retest some fraction of their i
equioecnt to comply with the aging and sanin requirements.
1 i
It was the staff's view that, when compared to the draft guidelines to be used in the upcoming staff reassessment of operating plants, i
i
)
neither the aging nor the sargin requirements of IEEE Std. 323-74 warranted backfit.
The benefits of backfitting these requirements would be a small, unquantifiable increase in the level of assurance j
that equipment is quallfted as compared to the significant costs that i
would be involved.
i Letter for All Construction Permit and Doeratina License Ano11 cants from D. Ross. " Issuance of NUREG-0588. 'Interin Staff Position on Ecutonent i
Qua11fication of Safety-Related Electrical foulement. '" February 5. 1980 l
l Construction permit holders and applicants for an operating Itcense were nottfled of the staff's resolution of Generic Technical Activity A-24 dealing with environmental qualification requirements. The addressees were informed that NUREG-0588, 'Interte Staff Position on Environmental Qualification of Safety-Related Equipment,' dated November 1979, was 1ssued for comment and estab11sh2d the staff's position on EQ.
license applicants were also informed that the staff will requireOperating j
j specific documentation of the degree to which they satisfy the criteria i
of NUREG-0588 in Section 3.11 of the Final Safety Analysts Report.
The
]
staff indicated that construction permit appitcants for which the Safety i
Evaluation Report was issued after July 1,1974, should comply with IEEE 323-74 (NUREG-0588, Cat. I requirements), with the following exceptions:
C-24
a-a^*'"'"'-"
- - - - ---- - - ~~~ ~ ~ ~
equipment which would normally be quellffed in accordance with
![EE 323-71 should be qu,alified in accordance with IEEE 323-74 (i.e.but w IEEE 323-was appilcable to all equipment purchased six months or m, ore aftar the issue date of NUREG-0588); and Note:
This position appears to have been everlooked by the staff Ce.g.
and Bellefon,te). current EQ reviews asseclated with Watts Bar i
beginning 12 months from the date of issuance of NUREG-0548 equipesnt which must undergo requalification testing should also be E-qualified in accordance with IEEE 323-74.
W Full implementation of the staff's position was not expected untti May i
1980.
Masorandum for Commissioner Kennedy f
- Environmental cualification of Class it foulneent " Februaryfrom W. Dirks. 'Off Bulleti i
- 11. 1980 i
dated February 4, 1980) about how EQ requirements for $[P being satisfied.
on the otb r plants by IE Sulletin 79-018 were also appilcable to the plants. However, since the adequacy of EQ for SEp plants was being 1
to the Bulletin. reviewed separately, it was not necessary for these licensees to respon
\\
Letter to Ooeratino Lleense Aeolicants from D. Anst. *0ualification of jafety-Related Electrical foulement." February fli 1980 The applicants for operating licenses were asked to evaluate and document compilance with NUREG-0588 as discussed in previous NRC correspondence l
dated February 5, 1980.
Commission Memorandum and Order (CLf-80-21) dated May 2L 1980 CLI-80-21 stems from a petition that was filed by the Unfon of Concerned Scientists down all ope (UCS) on November 4, 1977, requesting that the Commission shut j
rating plants and halt the completion of new piants.
j l
expressed specific concerns regarding fire protection for electricalThe UC5 cables and environmental qualification of electrical components.
j CLI-80-21 supplemented the decision that was made by the Consission on i
April 13, 1978, concerning this UCS petition.
j following major points and decisions were made by CLI-80-21:With regard to EQ, the The Commission concluded that current requirements in conjunction with the additional actions being ordered by CLI-80-21 provided t
reasonable assurance that the public health and safety was being i
adequately protected during the time needed for corrective actions to be implemented by licensees.
The Commission recognized that for the oldest plants, qualification C-25
~ '
0 was based on the fact that electrical components were of high industrial quality; for the newr plants after 1971, quellfication was judged on the basis of !((E 323-1971; and for the most recent plants whose Safety Fwaluation Reports wre issued after July 1 1974, qualification was based on Regulatory Culde 1.89 and IEEE 323-1974.
The Commission recognized that IEEE 323-1971, by itself, did not provide an adequate standard agatast which qualtffcation could be i
judged.
In short, the standard did not spectfy the accident conditions which electrical equipment must meet; there were no 4
specific requirements to maintain document flies; and there wre no specific requirements concerning margins, aging and other needed equipment specifications.
A program was currently underway to reevaluate the qua11fication of safety-related electrical equi i
completing this reevaluation, pment in all operating reactors.
In the 00R Guidelines w re being used to screen plants for EQ problems, and WRIG-0588 was being used as a guide for addressing the specific problem areas that were identified.
1 Issue:
Exceptions may have been allowed to the D04 Svidelines and to WRES-0584 during the staff's escreening and review
- of operating reactors, making it difficult to know just exactly what the sinta m acceptable standard was.
The Commission expressed the view that the 00R Guldelines ami WREG-0588 substantially improved upon IEEE 323-1971 and should provide greater assurcnce tant equipment is adequately qualiffed.
applying the criteria stated by the DOR Culdelines and WREG-0588, itBy was the staff's intent to provide a level of confidence essentially equivalent to that which would have been established by laplementing IEEE 323-74.
The Commission ordered that the DOR Guidelines and WREG-0588 form the requirements that licensees and applicants must meet in order to related electrical equipment. satisfy CDC-4 as it relates to environmental Further, the staff was directed to prepare additional technical spectf tcations for all operating plants to require the documentation provisfons stated by the 00R Gu'delines.
The Commission estab11shed that, unless there a re sound reasons to the contrary, the 1974 standard in WREG 0588 (i.e., Category I i
requirements) would apply to replacement parts.
The Comelssion recognized that the DDR Guidelines and WREG-0588 applied progressively less strict standards to the older plants and that justification for this position was not provided at the time the older plants are grandfathered from the provtstons of RG 1.89.
The Commission suggested that this problem would best be resolved by a rulemaking on environmental qualification of safety-grade electrical equipment.
C-26 I
I se e
The Commission directed that if the staff-proposed rule on [0 did not require plants to be upgraded to a single uniform standard along the Itnes of ths 1974 requirements stated la WREG-0548 then justification of a less rigorous standard for the older plants must be articulated in depth and be subject to casument in the rule saking proceedings.
Issue Otfferent EQ standards were required without supporting technical justificatten as tes (a) why more rigorous standards were warranted, and (b) why
- progressively less strict standards
- were adequate for the older plants.
Based on inadequate Itcensee responses to an !&E Circular and several Bulletins, and based on recent NAC inspection results, the Commission stated the view that the nuclear industry was not devoting the resources necessary to solve the environmental qualification problem.
)
i The Commission recognized that some [Q information vital to qualification judgements was not being shared among licensees due to the proprietary nature of the information. Consequently, the ssaff was directed to review anytrennental qualifthtton information la its possession to deterstne how much of the information could be released to licensees to aid them in making safety judgments. The results of the staff's review was to be forwarded to the Commission within 45 days.
The Cossatssion also directed the staff to pursue the
]
possibility of establishing, by the nuclear industry, a Nuclear j
Qualtfled Equipment Clearinghouse.
The Commission considered the staff's review of Ilconsee responses to Bulletin 79 018 to be of high priority, and the staff was requested to keep the Comunission and the public apprised of any further findings of incomplete environmental qualification of safety-related i
electrical equipment, along with corrective actions taken or planned.
The staff was directed to complete its review, including the issuance l
of Safety Evaluation Reports, by February 1,1981.
further directed that by no later than June The Commission t
i 30, 1982, related electrical equipment in all operating plants shall beall safety-j quallfled to the D0R Guidelines or WREG-0588.
{
In a separate memorandum dated May 27, 1980, the Commission asked the e
staff to address the delay associated with deciding upon an MRC l
anvironmental qualtftcation testing program.
l The Commission recognized that CLI-80-21 did not attempt to apply the i
lessons of Three Mlle Island to environmental qualification.
j issue was being addressed in the NRC Action Plan which was currently This under review by the Commission.
SECY-80-319. " Analysts of Alternatives for Conductino Independent i
Yerification Testino of Enytronmentally Dualified Eculoment." July 1219 1.
l As requested by the Commission in its Memorandum and Order dated I
l C-27
i i
April 13, 1978, the staff provided an analysis of alternatives for conducting independent vertf! cation testing of environmentally qualified L
equipment. Historically, the staff's practice in this regard had been to review some of the equipment qualificatten reports along with the Itcensee's safety Analysis Report. Currently, the staff was aise reviewing Ilconsee responses to IE Sulletins and a backlog of environmental qualification reports that had been submitted for equfpeent Installed in operating plants.
Under contract to the NRC, 5sadia reviewed a number of alternatives for addressing this issue (NURfC/CR-l!87) direct NRC review Based on its review of the Sandia
{
study, the staff concluded that some environmental qualification testing in conjunction with a certain amount i
of independent testing would be best.
It was the staff's view that i
indepth NRC coverage of about ten to twenty percent of the total number
{
of qualification tests would prevfde sufficient confidence that i
electrical equipment was being adequately quallffed and would also result i
in improved industry qualification programs. The staff proposed the following program to resolve the environmental qualtftcation issue for all nuclear plants required to meet IEtt 323-H:
a, Short. Ram:a Actf y1 ties:
perform indepth inspections and witnessing of selected equipment qualifIcationr, and perform selected independent verification testing.
b.
Lono-Rance Activitten:
standardize qualification criterta; establish (via industry) accredited testing laboratories; Improve test standards, specifications, procedures, and acceptance standards; and consider the addlt1on of ether equipment requlring qualification by test into this program.
The staff defined specific program tasks and indicated that action had i
been initiated on a few of these tasks.
For example, the American Society of Mechanical Engineers was contacted regarding laboratory.
j certification, an entsting contract with Sandla was extended to establish Indepsndent testing capability and to obtain technical support for the staff, and a separate branch was established to be responsible for environmental qualtffcation.
The following program tasks were
{
established by the staff:
d Initiate and participate in a third-party laboratory certification program.
J 1
Issue a regulation requiring equipment qualification to be performed j
by accredited laboratories, i
?
C
\\
=
I l
Identify weaknessas in the current qualification criteria and recommend changes to promote more precise criteria and inspectable standards.
i Develop standard qualification criteria.
Develop standard classes of environments.
j 1
Develop a afnimum set of standard test plans that encompass the i
accident profiles including aging requirements, etc., for all plants within the scope of the program (f.e., IEEE 323-74 plants).
a Participate in the improvement of existing standards for e
environmental qualification testing.
1 Recommend research projects related to resolving unknowns identified during IE's programmatic review in the environmental qualification i
field and direct significant results back to the industry for future j
laplementation.
Select one or more equipment prototypes from each category of i
equipment covered in the progras based on its volume of use in plants and relative safety significance for in-depth NRC coverage and review.
l Conduct site environmental qualtitcation inspections of plants under construction.
l Estabitsh a basis for selecting equipment that warrants independent
{
vertf tcation testing and perform verification tests on selected qualif ted equipment.
4
}
Catalog pertinent information relative to prototy qualtitcations, such as eqttpu nt identification,pe equipment i
{
conditions of quaitfication, test results, qualifled 11fe, etc.
i j
Subject to certain previous comments and directives the Commission approved the recommendations of SECY-80-319 in a mem,orandum dated i
September 16, 1980.
i j
SECY-80-370. " Order to Confirm the Recutrements of Commission trand 1
and Order of May 23. 1980.
Protection." Auoust 6. 1980Recardino Environmental Dualification and Fire 4
The Commission Memorandum and Order (CLI-80-21) established several requirements for licensees and the NRC staff in the areas of i
environmental qualtftcation of electrical equipment and fire protection.
1 One of those requirements called for the staff to prepare Technical Specifications for all operating reactors to specify requirements for documentation of equipeent qualification.
To address the Commission's i
request, SECY-80-370 proposed an order for Modification of License to be applied to all operating reactors.
proposed by the staff for Itcensee completion of qualificationA deaditne of Dece i
)
C-29 i
se i
documentation in order to support the Commission taposed deadline of february 1, 1981 of environmental, qualification for all operating reactors.and to allow time for th Other requirements of the Commission Memorandum and Order discussed bat not included in the ordering provisions included the June 30 1982, deadline for qualification of applicable equipment; provisions gov,erning replacement parts; and the obligation to modify or replace inadequate equipoent promptly.
$rCV-80 417. 'Isonet of TM1 Lessons tearned on uhe staf' Recutrements for Environmental Qualification of_ Safety Related E ectrica' fouinment.'
Sootember 19. 1980 The staff Informed the Commission of the impact of the TMI lessons learned on the currer.t requirements for environmental qualification of safety-related equipment.
This SECY Paper was prepared in response to a I
request that was made by the Commission in a memorandum of May 28, 1980.
The environmental qualification requirements stated in the 00R Guidelines and in NUREG-0588 did not include the recommendations of the TM! lessons learned study.
Consequently, based on a review of the lessons learned study, the staff identified the following issues that needed to be addressed:
Requirement for new scfety-related systems, including a reactor I
coolant vent system, reactor coolant and containment sampling l
systems, auxiliary feedwater initiation system, emergency power for pressurizer heaters, and additional accident monitoring instrumentation, needed to be established. The staff required these new systems to be included in the environmental qualification program for each plant.
Relief and safety valve systems and the emergency power system for pressurizer equipment needed to be upgraded to safety-related. The staff required these upgraded systems to be included in the environmental qualification program for each plant.
The radiation source term needed to be changed to recognize the potential for cireviation of highly radioactive fluids outside containment.
The staff was preparing a supplcment to IE Bulletin 79-018 to address this issue.
Temperature and pressure profiles in containment should include potential hydrogen burning effects. The staff was still investigating the hydrogen burn scenario.
Issue:
Hydrogen burn scenarios may result in conditions that exceeds the EQ envelope.
The potential for significant temperature and radiation stratification needed to be considered. The staff was considering the effects of both temperature and radiation stratification during the ongoing environmental qualification review of each plant.
l C-30
l,
.w Issue Temperature and radiation stratification may result in conditions that exceed the EQ envelope.
Order for Modification of License concern no the fnvircra-ntal_
l Qualiftention of Safetv-Re'ated flectric Loutonent. October24. 1980 Orders were issued to all power reactor Itcensees which modified their Technical Specifications in accordance with the Commission Memorandum and Order dated May 23, 1980 (CLI-80-21). The Commission directed the staff to prepare additional Technical Specifications for all operating plants which codified the environmental qualification documentation roostrements contained in Section 8 of the 00R Guidelines.
The Order imposed Technical Specification requirements applicable to all operating reactor power plants which stated that:
a) all safety-related electrical equipment in the facility shall be quellfled in accordance with the 00R Guidelines or MUREG-0588 by no later than June 30,1982 (copies of these documents were attached to the Order), and b) complete and auditable records must be available by no later than December 1,1980, and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detall to document the degree of compliance with the DDR Guidelines or NUREG-0588.
Further, the revised Technical Specifications stated that the environmental qualification records should be updated and maintained current as equipment is replaced, further tested, or otherwise further quallf ted.
SECY.81-477. 'Comoletion of Unresolved Safety issue A-24.* Auaust 7.1981 The staff informed the Cosmission of its resolution of Unresolved Safety Issue (USI by issuance) A-24, " Qualification of Class IE Safety-Related Equipment.'
l of NUREG-0588, Revision 1, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
Including Staff Responses to Public Comments.'
Since the Commission Memorandum and Order (CLI-80-21) made specific reference to the initial
'For Comment" version of the NUREG it was the staff's view that NUREG-0588 should not be modified untti the proposed EQ rule was developed.
clartitcations that are deemed appropriate.The EQ rule could then factor in any Consequently, NUREG-0588, Rev. 1, contained two parts. Part I was the original "For Comment
- version of the NUREG, and Part !! wts the staff responses and resolution of the public comments that were received.
SECY-81-504. "Eautoment Qualiffeation Procram Plan." Auoust
- 20. 1981 The staff informed the Consission of its plan for qualification of equipmen.t important to safety and requested the Commission's consent to implement the program.
of SECY-80-319 Reference was made to the Commission's approval and directives.(see above), subject to certain prior Commission comments This SECY Paper (SECY-81-504) discussed how the Commission's directives and the independent vertftcation testing program (SECY-80-319 program plan)(EQPP) which included an environmental, t.elsmic, a were integrated into the overall equipment cualtftcation qualification testing program; rulemaking activities; and research to be C-31
- - - - ~^ ~ ~^^~- ~ ~ ~ ~ ~ ^ ~~ ~ ~ ~ ^^^
6*
se conducted in support of the program. The objective of the (QPP was to provide a systematic approach to ensure that all egukmt laportant to safety in both operating and new facilities is pr r.t ir quallffed to perform their safety functions if subjected to i C les accident conditions or a salsaic event. The program var.
re ced to take about four years to accomplish the following activities; Review of the qualification status of equipment importsat to safety in operating facilities and identify unquallf ted equipment.
Enforcement of appropriate corrective actions, inc1 Wing relocation, replacement, or requalification of equipment.
Development of standardized NRC review procedures for equipment s
{
qualification to be used in the review of new facilities.
3 Develepsent of a rule on the qualiffcation of equipoent important to j
safety and the development of regulatory guides in support of the
- rule, i
Development of technology (analytical and experimental) in support of l
equipment qualtf tcation reviews.
l Testing and inspection of selected equipment by MRC to independently verify equipment performance under accident conditions.
Development and implementation of an accreditation program for test laboratories.
1 i
i i
i i
i c-32
- - - - ~ ~ ' ^ ~ ~ ~ ~ ~ ~ ~
1 1
APPD@!X 0
)
Informatten Concerning 10 CFR 50.49 The specific requirements for enilronmental qualification of electric i
equipment are stated in 10 CFR 50.49 (sometimes called the [Q ru?e.
i rule recefved a lot of attention and scrutiny from the Consission,)NRC staff, Thetg industry representatives, and others. By the time 10 CfR 50.49 was being i
finalized been expen,ded in addressing EQ, and (based on industry co
~
a rule that wovid tapose additional EQ backfit reoutrements.
The staff reviewed information concerning the development of 10 CFR 50.49 primarily to i
identify any potential EQ issues that were not fully addressed by the rule.
This appendix is a listing and descriptive suonary of the information that was j
reviewed in this regard, and potential EQ issues are listed in bold print.
Memorandun for R. Mlascue from H. Denton. *1tandard for Ovalf ficati j
Safety Re'ated faut;.; J t.* Sent_ Ler 2. 1980 The Consission's Memorandum and Order of May 23, 1980 JCL!-80-21), directed related electrical equipment.the staff to inttlate rulemaking on environme fication of safety-In this memorandum, the Office of Nuclear i
Reactor Regulation (MAR) requested that the Office of Standards Developmen 1
act1on, an amendment to 10 CFR 50 which would include:(05 i
i Ca) the rulemaking directed by the Cometssion in CLI-80-21 b the rulac2 king proposed in
{
SECY-80-319 on independent environnentaI q(ua)lification verLfic and inspection, and (c broadly address the qualification of both electrical and mechanical equipmen)t (including cheatcal process equipm hydrogen recombiners j
other environmental c)onditions.for seismic and dynaalc loading conditions as well be of greater benefit to the staff and give better guidance to th i
than one that addresses a more limited scope of equipment qualiffcation.
i was also reafnded that the rule should address requirements for operating OSD l,
reactors as well as fact 11 ties that are under construction.
Basorandum for W. Of rekt Oualification of f afety Grade Electric Eauta-ant.* January 6.1982from
{
1 The purpose of this meno was to report on the status and schedule for the i
The following points were made:rulemaking on environmental qualtftcati The environesntal qualification cf safety grade electric equipment
'1
,)
represented the flest of three related rulemaking efforts.
The other two efforts were the overall qualification of safety grade equipment i
includu rechantcal equipment, and the accreditation of laboratories m
i; that pe m re environmental qualiffcation testing.
Issue:
Laboratory accreditation say be necessary to assure that EQ testing is adequate.
\\
The Equipment Qualification Branch was reviewing all operating nuclear
l, l
a j
i power plants and those scheduled to be Ifeensed to operate by f6bruary 1 version).,1981, against the criteria stated in NURIG-0588 (for consent l
A revision to WRIG-0588 was scheduled to be issued by March 15, 1981, j
upon completion of the comment period.
t Concurrent with development of the (Q rule, a revision.to negulatory L
I Guide 1.89 was being prepared to specify in detall the acceptable methods for meeting 1 w rule.
It was thought that the revised NURti-0584 would w rve as the principle basis for the technical content of the regulatory guide.
The rule will address backfitting and the bases thereof in considerable i
detail as specified in the Coselssion's Memorandum and Order of May 23, j
1580 (CLI-80-21).
]
Note:
This evidently was not accomplished by the staff.
Memorandum for R. Minoous. N. Moselev. M. Thornburo. and 7. heter from W. Morrison. " Proposed Rulemakino and Associated Raou"atory Guide 1.89.*
March 2. 1981 Assistance was requested in reviewing the proposed rulemaking for EQ and the i
proposed revision to AG 1.89.
l provided:
The following background inforsation was The proposed rulemaking was being undertaken in response to the Commission's Memorandum and Order CCL1-80-21) dated 14ay 23,1980, relating 16 the environmental qualdffcation of electric equipment, including consideratton of backftt.
3.89 was being revised to state meth Ms acceptable to NRC staff for mening the Commission's requirements for environmental qualification of e!attric equipment important to safety.
Upon pubitcation of the final rule and the revistos to RG 1.89, it was the staff's intent to withdraw the 00R Guidelines and WREG-0588.
Note:
This indicates that the etaf' ild not originally intend to i
grandfather operating nuclear power plants under the EQ j
rule.
Memorandum for C. Knf ahton free 2. Rosrtocrv. *Pronosed Rulemak.ino and 1
Associated Reculatory Guide 1.89.* March 23. 1981 The Equipment Qualification Branch provided comments on the proposed EQ rule i
and the draft revision to RG 1.89.
comments:
The following viewt tv.n included in the The proposed ruiceaking should include setsafe requirements.
The statement that the 00R Guidelines and WREG-0588 apply progressively i
~
D-2 4
k i
=-
i i
Iss: strict standards to the elder plants is not approprfste.
There is no difference in the einfeus requirements except for the Cat. I and j
Cat. !! plants as defined in WR(G-0588.
Note:
l This view is not consistent with the tiew that was expressed by the Consission in its Memoranda and Order of May 23, 1980 the vCsee Appendix C). In fact, the Commissten expressed
(
ew that *...The Suldelines and IENLES-0548 spply progressively less strict standards to the older plants....'
1 i-The staff position related to use of the steam saturation temperature 1
(SST) corresponding to the pressure profile inside containment for the J
PWR plants and 20*F above t.w $$7 for SWR plants should be included in j
the rulemaking and also in RG 1.89.
i '
)
Leuter for R. Fett from L. Benmn ($andfa) rer revfew af aronesed f0 4
)
ru emakina and revision to Raouaterv Culde 1.89. Aart' 3. 1981 This letter forwarded Sandfa's cannents on the proposed revisions of NUREG-0584 and Regulatory Guide 1.89, and on the frepo i.
Sandla also made reference to previous consunts t at were subaltted by i
dated August 22, 1980, and January 13, 1981. Sandla made the following etters observations with regard to the proNised EQ rule:
i a.
General i
There appeared to be jurisdictional problems since proposed Regulatory Guide EM 008-4, " Plant Shlalding to protect personnel and Systems and Components important to Safety in the Event of an Accfdent
- dated March 5,1981,alsodiscussedsafety-relatedequipmentqualification.
It was suggested that the philosophy and coverage should be extended to all safety-related equipment, not just
- electrical.'
l b.
Rulemakino i
i There is a severe ratchet on old plants.
spectfic reference was made to synergistic effects.The state
}
!ssue:
Current ' state of the art capabilitlos' may not be i
sufflctently developed to support existing EQ requirements.
Some confilets existed between the proposed rule. RG EM 004-4, and Regulatory Guide 1.89.
while RG 1.89 requires consideration of rates.The rule only discussed " i The proposed rule i
confilets with the double peak required by RG 1.89.
Issue:
The ' double peak' requirement for E0 nsting may not be j
warranted.
l Although the rilue/!apact Statement states that a proposed revision to 4
D-3 i W
)
1 i
I i
!(([ 323-74
- ...is based substantially en the technical material in i
NUREG-0584....' in fact, the proposed revision to the 1(([ standard i
disputes NUREG-0584 almost on a point-by-point basis.
i t
j
!ssue:
The revised !!!! standard Cf.o., post-!Ett 32 b74) any be j
more appropriate for esta611shing EQ than the previous version.
j i
j Memorandum for R. Ifenzfnoer Jrom L Aeoarwa1. 'Pronomd Rulemakfna and Associated Reau atory Culde
.89 on inviror.s ntal Qua' ification.' Acril 7.
I 1H1 j
The minutes for a meeting that was held on March 26, 1981, to discuss EQ rulemaking issues was attached. Of particular f aterest, the meeting minute.s 1
included the following positions:
l "The final rule say not be identical to NUREG 0588 and the DDR Guidelines.
If the final rulo is significantly different free NUREG 0588, the schedule for enforcement of the new requirements should
{
be included in the rule...The Equipment Qualtffcation Branch (EQB) will j
provide a list of differences to us based on codified portions of the proposed rule, by April 15, 1981.'
' Reference to ' progressively less strict standards' should be deleted from the rule.
Reference to Category I and !! should be made instead.
EQ8 will justify why uniform criteria for All nuclear power plants i
cannot be applied.
This input will be provided to us by April 15, i
1981.*
Issue:
Technical justification was requested but not provided for j
why a uniform criteria w.is not applied to all nuclear power plants.
F l
Memorandum for C. Knf ahton frra I. Ros2toery re the draft E0 rule and charices j
to R.G. 1.89, dated May 11. 1981 The Equipment Qualification Branch provided consents on the draft EQ rule and on Regulatory Guide 1.89.
Of particular interest is the comment that ' Aging requirements should be expanded to laclude the requirements that equipment cust be aged in the same functional state it is used in the plant (e.g.,
energized,de-energized)."
Issue:
It may be appropriate to perform aging of equipment in the same j
functional state that it is used in the plant (i.e., energized or i
de-energized).
4 f
Memorandum for R. Mineous from V. Stello. "Pronosed Rulemakina. ' Environmenta and Selsaic Qualification of Electrical Ecutonent for Nuclear Power Plants.
i Sectember 4.1981 The Office of Inspection and inforcement concluded that it could not concur on the proposed rule.
A major concern that was cited was that the proposed rule i
D-4 s
=
added additional requiremts (e.g., seismic and dynamic) beyond those contained in the Commission's Memorandum and Order (CLI-80-?)) dated May 23, 1980.
SECY-81603. "Procesed Rulemakino. ' Environmental Oualification of Ecutoment for Nuclear Power Plants.'" October20. 1981 The staff submitted a proposed rule for environmental qualification of electric equipment and requested the Cosuiission's approval to publish the proposed rule in the Federal Reaister.
the Commission's Memorandum and Order (CLI-80-21) dated MayThe staff's SECY paper contained the following information:
23, 1980. The The proposed rule did not cover seismic and dynamic qualification requirements.
The proposed rule pertained to Class IE electrical equipment and some additional non-Class IE equipment and systems.
It was stated that "The proposed rule will codify explicitly the current NRC practice with respect to environmental qualification of electric equipment and will apply the same uniform performance criteria to all operating nuclear power plants and plants for which application has been made for a construction permit or an operating license.
Included are specific technical requirements pertaining to:
(a) qualification parameters, (b) qualification methods, and environmental qualification methods are prog (c) documentation.
The older plants."
ressively less strict for j
Note:
This seems to conflict with what is stated in the next bullet.
It was the staff's intent to issue Rev. I to Regulatory Guide 1.89,
' Environmental Qualification of Electric Equipment for Nuclear Power Plants," concurrently with the proposed EQ rule.
l The implementation section of the Regulatory Guide provided specific guidance for stating the qualification requirements of the proposed rule at older plants that i
and the practicality and cost effectiveness of conductin j
qualification testing of installed electric equipment.
It was the staff's intent to withdraw the DOR Guidelines and N i
upon publication of the final EQ rule.
i nIluead$1onI fee 1ntent to expand the scope of eE 1
t 1
fN$c t ft t e st!
p a na e eg d re s
s fication of electrical a$d eYar equ$pNn u
mpor to I
)
Issue:
The adequacy of EQ for equipment other than electrical, and 1
I i
D-S i
c.
for other conditions referred to by the GDC such as ssismic and dynamic loading effects was considered but never addressed, SECY-81-603A. "Pronosed Rulemakina. ' Environmental cualification of Electric Ecutoment for Nuclear Power Plants.'" NoveaeGr 4.1981 The staff provided three more versions of the proposed EQ rule in addition to the one discussed in SECY-81-603 that was sent to the Commission on October 20, 1981. In particular, the staff reconnended changes to the proposed rule to:
(a) strengthen criteria for environmental qualification of electrical equipment, and pending development of a va(b) defer the seismic and dynamic requirements lue/ impact statement.
SECY-81-6038. "Pronosed Rulemakina. 'Enyf rc.a-ntal Qualification of Electric Eculoment for Nuclear Power P' ants. '" November 163 Based on comments received during the Commission meetir:g that was held on November 10, 1981, SECY-81-603 was modified.
In particular, a new paragraph was ad6d to clarify that the requirements for seismic and dynamic qualification of electric equipment are not included in this proposed rulemaking, and the statement of considerations was expanded to clarify the conditions under which
- analysis alone" in lieu of testing would be acceptable for equipment qualification.
i l
Memorandum re; Affirmation of SECY-81-6038 dated Janusrv 7. 1982 y
i The Commission approved the proposed rule (SECY-81-6038), with noted clarifications / changes, for noticing and public comment.
The Commission also asked that the staff address a number of issues, including under what i
circumstances licensees are obligated to report to the NRC when equipment falls a qualification test, whether the staff believes that additional reporting requirements are necessary, and when reporting such a failure would come under the requirements of 10 CFR Part 21.
SECY-82-51. " Staff Reouirements - SECY-81-6038 - Pronosed Rulemakina.
j
' Environmental Oualification of Electric Ecutoment for Nuclear Power Plants.'"
l February 4. 1982 The staff provided its response to the January 8, 1982, memorandum from S.
Chilk requesting the staff to provide specific information to the Commission.
Information was provided regarding the adequacy of licensees' justifications j
for continued operation, reporting requirements were discussed, information on the capability of plants to achieve cold shutdown was deferred to a later i
date, reference was made to SECV-82-31 for the staff's position on enforcement I
i of EQ reoutrements, and the staff's proposed plan for equipment qualification
{
(including seismic) was deferred to a later date.
1 l
lilth regard to reporting requirements, the staff recognized that there were no reporting requirements per se for equipment that falls qualtftcation tests unless the equipment is installed plant equipment.
The staff did not consider such requirements to be necessary since such information would not require 4
further action unless the tests represented installed equipment.
It was i
D-6
9@
et presumed that a licensee would not knowingly attempt to install equipment in a facility that has failed a test.
Note:
This view did not recognize the possibility that similar equipment may be installed at other nuclear power plants, or that other nuclear power plants might be considering the use of stallar equipment.
Issue:
Additional reporting requirements for EQ problems that occur (i.e., probless that occur during qualification testing as well as problems that occur during plant operation) may be warranted given the uncertainties associated with establishing and maintaining equipment qualiflcation.
SECY-82-207. " Final Rule. ' Environmental Oualification of Safety-Related Electric Eculoment for Nuclear Power Plants.'" May 24. 1982 The staff provided the Comission with a final version of the EQ rule after making changes based on the public comments that were received on the proposed rule (SECY-81-6038). Among other things, the final version of the rule j
of Unresolved Safety Issue (USI) A-45, removed safety-rela i
i dynamic qualification from the scope of the rule, did not re j
that was previously qualified to be requalified under the new rule, ano deleted the requirement for a central file.
Issue:
EQ requirements for cold shutdown equipment should be assessed.
Memorandum for Flies from S. Trubatch. " Final Rule on Environmental May 28. 1982_Oualification of Safety-Related Electric Eculoment for Nuclear Power Plan S. Trubatch (OGC) indicated that differences between the proposed EQ rule and the final rule posed some legal problems. First, the proposed rule did not
)
include the grandfather provision that was included in the final rule.
S. Trubatch indicated that it might be appropriate to rerotice The public coment on this aspect of the rula.
Trubatch was that (stallar to the grandfather issue) the final ruleThe second issue ra j
effectively relaxed the qualification requirements that were established by i
the proposed rule for replacement parts.
for this relaxed position, and S. Trubatch indicated that it might be.The staff di l
j appropriate to renotice the rule for public coment on this aspect of the rule l
as well..
i l
Issue:
EQ requirements for replacement parts should be reassessed.
l t
SECY-82-207A.
- Final Rule. ' Environmental Oualification of Safety-Related Electric Eculoment for Nuclear Power Plants.'" June 9.1982 i
previously submitted on MayThe staff provided the Comission with a revised l
24, 1982, as SECY-82-207, which included D-7 c__________________-___-_-___-__-__-_____-_________-_________________
i m
meeting on this subject. consideration of comments that wera received at the SECY-82-207 was modified as follows:
The " grandfather
- provision was expanded to clarify the requirements for plants currently under review for operating Ilconses.
The definition for " safety-related" was clarified, and the statement of considerations was expanded to clarify that certain post-accident monitoring equipment is covered by the rula.
The staff also noted that the proposed revision to Regulatory Guide 1.89, was issued for public comment in February 1982 and, after later than Septemberresolved, it was the staff's expectation to issue the Eegulatory 30, 1982.
final rule should not be conditioned on issuance of RG 1.89 rule, in conjunction with the statement of considerations, was sufficiently explicit.
)
Memorandum for W. Dircks from S.Chilk. 'SECY-82-207/82-207A - Final Ru Power Plants.'" June' Environmental Oualification of Safety-Relateo Electric Eouloment
- 25. 1982 The Commission provided the following guidance for revising the final EQ rule (SECY-82-207):
a qualification req (u)irements for replacement parts needed to be direction was given regarding the scope of the rule; (b) rule; (c) the rule should assure that there is no relaxation of requirements contained in the Commission Memorandum and Order (CLI-80-21) of May 23,1980; (d) PTOL plants including Comanche Peak and later plants should be clearly required to meet the Cat. I requirements of NUREG-0588; and (e) an option for qualification of equipment required to bring the plant to cold shutdown should be included.
SECY-82-207C. " Final Rule. ' Environmental Oualification of Electric E
. Imoortant to Safety for Nuclear Power Plants. July 27. 1982 The proposed EQ rule was modified (SECY-82-207A) based on the staff requirements memorandus changes, the rule was rev(SRM) dated June 25, 1982. In addition to editorial ised to indicate the applicability of the Cat. I cold shutdown requirements were added. requirements of MUREG-0588, optio SECY-82-207D. "Sucolement do SECY-82-207C Final Rule. ' Environmental Qualification of Electric Houinment imoortant to Safety for Nuclear Power Plants.'" Auaust 26. 1982 SECY-82-207C was supplemented with the following information:
replacement parts; (c) the status of RG 1.89 was provided; a (a) the cold deadline for equipment qualification was clarified.
I 0-8
i l
Memoranem for the Commiqsion from L. Bickwit. Jr. (General Counsel).
final Rule 'Environmenta' qualification of Electric Eouiement freertant to
" Draft Safety for Nue" ear Power P
' ants.' $fCY-82-207C/207D.* Sectember 2.1982 The purpose of this memorandum was to provide OGC's views on the draf t fi i
rule for environmental qualification of electrical equipment.
l view that certain aspects of the rule were not well supported by the It was OGC's i
rulemaking record and that the solicitation of additional consents would be j
advisable.
4
'The Comission's discussion of grandfathering in CLI-80-21 and the i
decision not to include grandfathering in the proposed rule seem to
{
require the Comission to obtain coments on the proposed grandfathering 1
i in the final rule."
Note:
This indicates that the decision to grandfather plants was J
supported by the Commission.
4 1
replacement parts, a justification for that decision sho 4
j in the statement of considerations."
"The Comission is now considering whether to require the 1
of one train of equipment needed to achieve cold shutdown. qualification to us that there is nothing in the record of this rulemaking proceeding
...It appears to support a re cold shutdown..quirement mandating the qualification of equipment for to be provided with the final rule."..At the very least, supporting documentatio that the testing of equipment is a betcar sethod for qu i
equipment than provided by analysis or operating experience...g though the preferability of testing is somewhat self-evident, OGC
.Even believes that this aspect of the rule would be strengthened by the provision of a rationale supporting testing as the better method for i
environmental qualification."
" Industry believes that there is no technical material in the record demonstrating that accelerated agino can be used to reach a well-defined j
qualified life... 0GC believes that the rule would be strengthened if the record contained or referred to documentation demonstrating the i
wallability of accelerated aging techniques capable of resulting in a well-defined final state for aged equipment."
i i
Issue:
It may not be possible to establish a well-defined qualified 1
life given the current state of the art.
Letter for N. Palladino from N. Reynolds (NUGEO).
Reoardina Environmental Qualification of Electripal Eouioment." Seotember 10 IMZ The Nuclear Utility Group on Equipment Qualification (NUGEQ) submitted coments to the Chaiman regarding the latest draft of the EQ rule 4
0-9
j i
i i
i (SECY-82-207C relative to the previous draft (SECY-82-207). Among the several commen)ts, NUGEQ included views on grandfathering i
parts.
NUGEQ believed that, contrary to the Cossaission's apparent intent, the new proposed rule changed the provisions of the previous version of the rule and did not grandfather operating reactor licensees.
replacement parts, NUGEQ stated that the new proposed rule failed to includeWith
{
the Cossaission's direction to make allowances for " sound reasons to the
{
contrary."
Note for W. Johnston from S. Accarwal orovidino caa-ents on a draft response
}
to NUGE0 concerns. October 10. 1982 S. Aggarwal provided his coments on a draft response to NUGEQ concerns that were addressed to N. Palladino in a letter dated September i
10, 1982. 'With regard to grandfathering plants, 5. Aggarwal expressed the view that the proposed rule would not require licensees to requalify equipment that had already been qualified under the previous requirements, but that it was never.
the staff's intent to grandfather plants.
Therefore, it was possible that some of the existing equipment that had not been previously qualiffed would have to be qualifted in accordance with the requirements set forth in the s
{!
rule.
l l
Note:
This indicates that the staff's intent was not to grandfather t
plants from the provisions of the EQ rule, but rather to grandfather existing equipment that had already been qualified.
Memorandum for Commissioner Ahearne from W. Dirks. "Envirer.rertal j
Dualification Rule." October 26. 1982 1l{
The purpose of this meno was to provide the staff's interpretation of " uniform i
standard' as used on page 7 of the Commission's Memorandum and Order j
(CLI-80-21). The staff's view was that " uniform standard" referred to both the qualification parameters and the methodology. The meno stated "....In j
sumary the staff believes that the electric equipment qualifted in accordance with the DOR guidelines and/or NUREG-0588 (Categories I and II) will satisfy si ll the requirements of the proposed final rule.
need not include a specific ' grandfather provision.'Thus, the proposed final rule l'
Paragraph 50.49k was included in the proposed final rule to assure the industry that previous valid j
qualification efforts would not be wasted."
l Note:
This appears to be an attempt by the staff to justify the existence of 'different EQ standards' for the older reactor i
plants.
Contrary to the above stated position that was taken by i
the staff, the 00R Guidelines and NUREG-0588 l
satisfy the requirements stated by the propos Cat. II, did not ed final rule (e.g.,
different standards were established for aging, margins, and doctamentation).
.I' Affirmation Resnonse Shee". Chalman Palladino to S. Chilk. "SECY-82-207C -
I Final Rule. 'Enviror. antal Qualification of Electric Eautoment Imoortant to f-Safety for Nuclear Power Plants." November 1.1982 In his Affirmation Response Sheet, Chairman Palladino voted to approve the i.
l D-10 u
~ _., _ _ _ -, _,.
,l final rule.
However, the Chairman expressed the view in his conenents that "Section 50.49(k) should be replaced with one that grandfathers plants rather than equipment.
Operating reactors and plants under review for OLs that have been previously ordered to comply with the DOR Guidelines or NUREG-0588 s not be required to do any further qualification to meet the provisions of the rule."
Note:
This indicates that Chairman Palladino cencluded, without technical basis, that plants should be grandfathered from the provisions of the EQ rule.
t Memorandum for W. Dirks from S. Chilk. " Staff Recuirements -
Affimation/ Discussion and Vote. 3:30 P.M.. Thursday.
January 6. 1983.
Commissioners Conference Room (Ocen to Public Attendance)." January71 1983 I
'The Comission, by a vote of 5-0, approved for pubitcation in the Federal Reaf ster a final rule...to codify the methods and criteria for the addressing the issue of previous qualification efforts (f.e., t In proposed did not recognize that operating plants ju:,t completed qualification of equipment to the 00R Guidelines or NUREG-0588 down the drain, the final version of the rule was expanded to alleviate this concern (i.e.,) older plants were " grandfathered").
Note:
This indicates that the decision to grandfather older plants from the provisions of the EQ rule was purely economical.
Menorandum for H. Denton from T. Soeis. "SECY-82-207C/D - Final R 4
Environmental Oualification of Electric Ecutoment Imoortant to Safety for Nuclear Power Plants." February 4.1983 The issue of EQ for equipment necessary to achieve cold shutdown was i
discussed.
vehicle for addressing this issue, the memo was written to pr l
clarification of what was not within the scope of USI A-45.
identified in this regard included:
Issues that were shutdown eqaipment; (a) qualification requirements for cold equipment; and (c) me(b) postultted failure mechanisms for cold shutdown failure of cold shutdown equipment.asures to be taken to protect against environment Issue:
EQ requirements for cold shutdown equipment may not be adequate.
Memorandum for D. Eisenhut frg,fEljgr and R. Mattson. " Guidance for Licensees and License Aeolic:
' Environmental Dualification o'f Electric Eouiement Imoortant to Safety Nuclear Power Plants.'" Acril 8. 1983 The staff provided guidance to infom Itcensees of the infomation required in their May 20, 1983, response to the EQ rule (i.e., 10 CFR 50 j
recomended that the guidance be sent to the licensees as a g.49) and The major concern expressed by the staff was the method used by licensees for eneric letter.
identifying electric equipment important to safety, including non-safety-1 0-11
of safety functions by the safety-related equipment.related eq guidance addressing this aspect of the rule as well, which defined theThe staff dev electric equipment of concern, provided guidelines for assuring safety functions (including surveillance measures, operator actions, equipment isolation, etc.), and suggested methods for reporting the information required by the rule.
SECY-83-271. " Decision in Union of Concerned Scientists v. Nuclear Reaulatory Commission and United States of America. p. S. Court of Anneals for the District of Columbia Circuit No. 82-2000." Jul y 5.1983 The court ruled in favor of UCS, finding that the NRC did not adhere to the regulations when suspending the June 30, 1982, deadline that was established i
for licensees to implement EQ requirements without providing for pubile notice and coment.
Memorandum for R. Vollmer from W. Olmstead. 'Interoretation of 10 October 4.1983 In a memorandum dated August 31, 1983, the staff requested a legal interpretation of 10 CFR 50.49 the environmental qualification rule. The following interpretations were,provided by the Office of the Executive Legal Director:
The requirements stated in 10 CFR 50.49 do r,vt establish the complete requirements for submitting environmental qualification information in an application.
The applicant must meet the application requirements of compliance with all applicable Commission. regulations.Se The appiteant may not rely upon the limited reporting requirements of Section 50.49 as a basis for declining to supply the information necessary for processing the application.
In the statement of considerations accompanying the final r11e, the Comission clearly stated its intent to apply Category 11 requirements requirements to later-vintato plants whose CP-SER predated July 1, 197 subsection (k) of the rule,ge plants. Also, during development of the first plant to which the criteria of NUREG-0588, Cat. I, wouldth apply.
Therefore, in the case of any earlier-vintage plants that were 1
equipment, the CP-SER date should be treated as the basi establishing which NUREG-0588 category applies.
The qualification deadline established by subsection (g) of the rule does not apply to Itcense applicants.
Therefore, it is necessary to impose a license condition for license applicants in order to establish a qualification deadline.
Those plants that were authorized to operate at 5% power at the time the i
without completing qualification. rule became effective are not " appli 1
However, these plants should have 0-12
- \\
to 4
subeltted a justification for continued operation which fully considers
- i full-power operation before being granted a full power operating license
{
by the staff.
of Reg. Guide 1.97The rule requires qualification of all equipment in Cate Rev considerations acco,mpany. 2, as indicated in the statement of 1
ing the final rule.
must be qualified in accordance with the subsection (g) schedule.The e be fully qualified at the time of installation. Equipment that i
New equipment not previously installed in a plant (f.e., the equipment is neither existing or replacement) should be considered " replacement equipment" for the purposes of subsection (1) of the rule and qualified j
i accordingly. The intent of this provision is to gradually upgrade the criteria rather than the previousqualification level of all equipment to t
3 less demanding, guid3nce.
In some by the rule. cases, however, there may be "soun,d raasons to the contrary" as allowe j
SECY-83-457A. "NRC Resoonse to Court Decision Vacatino Interim Ru l
Environmental Qualification Deadline (SECV-83-457).* December 7. 19 The staff provided background information related to the environmental qualification issue and reconsended that the Commission proceed with the proposed " notice and comment" rulemaking to formally delete the June 2
deadline from all Itcenses.
30, 1982 Court's decision in UCS vs. NRC (see SECY-83-271).The staff's actions we I
]
Memorandum for the Commissioners from Chairman Palladino. *Env Oualification Meetina Scheduled for January 6. 1984." January 3. 1984 _
1 Sandia staff on environmental qualification.The Chairman indicate It was Sandia's view that'the HRC EQ program had problems.
s The briefing slides raised the following issues:
jj Qualification methodologies have shortcomings.
e Design bases have shortcomings.
Some inadequate equipment is installed in the plants,
=
l SECY-84-83. " Environmental Qualification Poliev Statement - SECY-l February 21. 1984
\\
Stellar to SECY-83-457A, except the Sandia views on EQ shortcomings were included -(see N. Palladino meno dated January 3, 1984).
l The SECY Paper bases (acceptance criteria), and the presence of inadequaj 4
J plants."
raised by Sandia "were addressed," the SECY Paper did not 1
j concerns were fully resolved.
Identified in the SECY Paper.
The Sandia concerns were not specifically i
D-13 g
s v
w e
w+-
w 0
r-t
lH1 m
APPDe!K E 5tandard Review Plan Section 3.11 The 51andard Review Plan Qualification of Mechanica(WREG-0800), $ect1on 3.11, " Environmental 1981, contctns the staff's criteria for as uring confermance with requirements of 10 CFR 50 Appendix A, General Design Criterton 4
- Environmentt? and Dynamic Effects Design Bases.* The purpose of the staff's of equipment (mechanical and electrical, including instr control normal,) abnormal, and accident environmental conditions.ar focuses on slid as well as harsh environments.section of the S Therefore, this Qualification of Safety-Related Electrical Equipmen 1
1979.
In general WREG-0588 Category !! requirements are called out for plants whose construction permit SERs were dated before July 1,1974, an NUREG-0588 Category I requirements are called out for plants whose construction permit SERs were dated after July 1, 1974 many other standards and Regulatory Guides for specific equipmentThe SRP refere qualification criteria such as for motors, valve operators, and penetration assemblies.
Two documents in particular, RG 1.89,
- Environmental Power Plants.* and IEEE Standard 323Qvalification of Certain Elect1
" General Guide for Quab rytog Class !
Electric Equipment for Nuclear Power, Generating Stations,' are reference source documents for stating the fundamental principles red criteria for qualf fying equipment.
i i
Issues:
(a)
The 3RP is very much out of date a
- central file" grove; Gere is ne(e.g., the EOS is listed the lead review is referred to contrary to uhat wasreference to the required by IC CFR 50.49, and R$ 1.89 and IEEE 325-74 are not recogn ued as the appropriate guidance documents for satisfyird the EQ requirements).
(b)
Perlofic maintenance and surve111ance requirements necessary to w intain EQ have not been defined.
(c)
IEEE Standards ethers) have not been endorsed by the IWtc.381, 535, 627 (d)
The 3RP suggests that IRNtEG-0588, R$ 1.89, and IEEE 323 may be applicable for qualification of mechanical equipment, but i
no specific guidance has been provided in this regard.
RELATED CORRESPONDENCE:
Dualification Reoutr a nts.* FebruaryMemorandum for D. Eisenhut from j
- 17. 1982 by uttitty representatives, and requested that the clar
. l a-
~
all operating reactor licensees.
clarifications during its review of licensee environmental qualtftcationIt was submittals. The following clartftcations were stated:
j All display instrumentation identified in the emergency procedures need i
not be identifted; only those operator display instruments which are safety-related need to be identified and have available qualification documentation.
Issues:
(a)
Non safety-related instruments that could impact the i
operability of safety-related instruments wre not initially included in the scope of equipment that was required to be qua11 fled.
(b)
RS 1.97 instruments were not specifically addressed.
Identification of the safety-related equipment installed at specific j
plants can be obtained from FSARs, Technical Spectftcations and other docketed correspondence setting forth NRC requirements or Itcensee i
coensitments; the necessity for upgrading non-safety-related systees to i
safety-related status will be the subject of other NRC reviews.
i Sound reasons wf',y replacement parts need not be upgraded to NUREG-0588 e
Cat. I requirements include:
a) replacement equipment sattsfying NUREG-0588 Cat.11 requirements l
was procured prior to May 23, 1980; i
b) replacement equipment satisfying Cat. I requirements does not l
exist; I
c) replacement equipment satisfying Cat. I requirements is not available to meet installation and operation schedules (Cat. Il replacement equipment may be used in the interis);
d) significant plant modifications would be necessary to support installation of Cat. I replacement equipment; e) operating performance and reliability data for the Cat. !
equipment indicates poor overall equipment performance; and I
f) the use of Cat. I equipment is Itkely to cause significant human factors problems that will negatively affect plant safety and performance.
Issue:
The reasons allowed by the staff for not upgrading replacement equipment to NUREG-0584 Cet. I requirements appear to be without merit and should be justified.
Equipment environmental qualification can be adequately demonstrated for use in st1d environments by implementing the following three programs:
a) a periodic saintenance, inspection, and/or replacement program; E-2 i
4 i
l j
b) a periodic testing program to verify operability within performance specification requirements; and j
c) l an equipment surveillance program which includes periodic
(
inspections, analysis of equipment and component failures, and a review of the re testing programs.sults of preventive maintenance and periodic Also, purchase spectficattons for replacement and new equipment must reflect those design basis environmental conditions that are bounding ior a11 app 1Icable equipoent local 1ons.
j
!ssue:
purchase specification considerations for replacement parts j
are not reflected in the sap.
t Qualification for submergence must be demonstrated and documented for i
equipment that could become submerged due to a high energy line break outside containment.
i The screening value of 4[7 rads is only appitcable to PWRs with dry typ containments.
However, plant specific analysis say be used in Ifee of i
the screening value.
j plant spectfic analyses. Plants with other containment types must also use
}
The staff values identified in the SER: of T, for PWRs and T.' + 20*F l
for SWRs can be used as the maximum in-contaYnment temperature for i
equipment quellffcation, or plant specific analysis may be used.
1 Plants subject to the 00R Culdelines or to NUREG-0588 Cat. !! that have tested equipment prior to May 23, 1980, i
to qualify that equipment to the required operating time plus anmay use test data appropriate margin.
The one hour margin need not be applied, tot subsequent failures should be shown not to be detrimental to plant i
safety.
Also, the one hour time margin rule is not appitcable to changes in the environment at the eqvfpment location. equ i
i
!ssue:
Use of an ' appropriate margin' instead of the one hour mergin that is currently required may be more reasonable and i
sufffctent.
i i
A qualtfled life need not be deterutned for plants subject to the 00R
=
i Guidelines or to NUREG 0588 Cat. !! requirements.
i An acceptable method and surveillance program with equipment and compone j
of inspections, and manufacturer's recommendations.and/or r i
t Issues:
(a) in lieu of defining a 'qua11fted Itfe,' it any be more 1
appropriate to develop methods for addressing in-serylce degradation.
i (b)
Preventive maintenance and survei11ance prograos for j
maintaining EQ have not been defined.
1 E-3
\\
i i
i APPDCII F i
Regulatory Sulde 1.89 a
i Regulatory Guide 1.89, 'Qualtf tcation of Class It Equipment for Nuclear power Plants,' was issued in November 1974 to provide the NRC staff's endorsement i
IEEE standard 323-1974, "IttE standard for Qualifying Class It Equipment for Nuclear power Generating Stations."
clartr(shortly after 10 CFA 50.49 was established), providing substantialRe 1984 i
with the requirements that were included in 10 CFA 50.49.ication i
j Regulatory Guide defined " qualification" as a *vertitcation of design 11alted Revision 1 of the to demonstrating that the electric equipment is capable of perfoming its safety function under significant environmental stresses resulting free desig 4
i basis accidents in order to avoid common-cause failures.*
1 that there are ' considerable uncertatattes* associated with the qualification The RG recognized process (especially with regard to preconditioning of e f
the view that an latograted approach to equipment quali lpeent) and promoted j
adopted.
cation should be samples employtag the Arrhenius theory and C2For example,
- ...a combination of (1) preconditioning of test 1
degradation processes that, based on expertence, are no survelliance, testing, and j
f preconditioning..." could be credited for satisfying [0 requirements.
Issues (a)
As stated by R$ 1.09, the purpose of env qualifIcatien is to avo14 'cemmercasse'ironmental i
j this) it is not clear why it is necessary tofatlures and (givon equipment to protect against single failures. qualify i
(b) i Elven the uncertainties inherent in the quaitficatten process, more emphasis may be needed en survet11ance, i
testing and maintenance to maintain the qualificatten l
status.
state of the art that entsted at that ttee.The regulatory position delinea Research programs were in j
progress to investigate such concerns as the effects of oxygen in a LOCA environment, the validity of sequential versus slaultaneous appiteations of steam and radiation environments, and fission product re accidents.
lead to a revised regulatory position.
the regulatory position that was established in RG 1.89:The following points were inc j
1 i
Complied radiation effects data on all classes of organic compounds show that compounds with the least radiation resistance have damage thresholds greater than IE(4 j
studies have shown failures )in metal calde seelconductor devices at rads.
i somewhat lower doses.
The regulatory guide established methods that were acceptable to the staff for determining temperature and pressure conditions inside contatnaent during LOCA and Main Steam Line Break (MSLB) accidents for equipment qualificatton purposes.
referenced in this regard.
Srectfic computer codes were i
4
,.m m.
i u..
.-- -- - - -- - - ~~~ ~ ~ ~ ~ ~ ~ ~ ~
t The regulatory guide established methods that were acceptable to the L
staff for detemining r qualification purposes.adiation doses to be assumed for equipment Additionally, the position was estabitshed that electric equipment that may be exposed to low-level radiat1on doses should not generally be considered exempt from radiation qualification testing without supporting justification.
Following the TMI-2 accident, the staff concluded that a thorough examinatt on of the source ters assumptions for equipment qualification was warranted.
time AG 1.89 was issued could lead to modifications in sourc assumptions.
Issuet In light of the TMI-2 experience, changes may be warranted i
in the postuinted source ters for E0 i
Electric equipment that could become submerged should be identified and qualifted by testing in a submerged condition to demonstrate operability for the duration required.
i Electric equipment located in an area where rapid pressure changes are
)
postulated to occur simultaneously with the most adverse relative humidity should be qualified to demonstrate that the equipment seals and to the degree necessary to maintain equipment functiona i
Issue:
Quallflestion for equipoent seals and vapor barriers en plants that are subject to the 00R Suldelines and IRRES-0588 should be reassessed.
performance characteristics that demonstrate the operability of testing throughout the range of required equipment op Cheetcal spray or deelneralized water spray that is representative of service conditions should be incorporated during simulated event testing.
)
{
5pecific guidance was provided with regard to test margins.
Synergistic effects that have been identified should be accounted for in the qualification test program.
Synergistic effects that were known at the time that RG 1.89 was issued included dose rate effects and resulting from the different sequences of applying radiation and elevated touperatures.
i The expected operating temperatures of the equipment under service conditions should be accounted for in thermal aging.
The aging acceleration rate and activation enerites used during qualtftcation testing and the basis upon which the rates and activation energies were estabitshed should be defined, justified, and documented.
i l
F-1 m.
4-Periodic surveillance and testing programs w re acceptable to account for uncertainties regarding age-related degradation that could affect j
the functional capability of equipment.
be acceptable as ongoing qualification to modify designatedResults i
qualifted life) of equipment and should be incorporated in theife (or eatntenance and refurbishment / replacement schedules, i
Sound reasons that w re recognized for using replacement equipment that i
was previously qualtfled in accordance with the DOR Guidelines or i
l NUREG-0588 in Iteu of upgrading to RG 1.89 requirements included:
i (a) the component is routinely replace 4 during nomal maintenance; (b) the component is a piece of equipaint that is qualf fled as an 1
assembly; (c) the component was on hand as part of the utt11ttes stock prior to
{
February 22, 1983; (d) replacement equipment qualifted in accordance with the provisions of 10 CFR 50.49 does not exist; o
I (e) i quallfled replacement equipment is not available to meet installation and operation schedules i
replacement equipment must be installe(d at the next availablein th outage of suffletent duration after it has been received);
(f) use of qualified re j
plant modtfication; placement equipment would require signtficant t
and t
(g) use of qualif ted replacement equipment would Itkely cause plant safety and performance,significant human factors problem i
Issue:
i The reasons allowd by the staff for not upgrading replacement equipment to the criteria stated by as 1.8g appear to be without merit and should be justifted.
Specific guidance was provided with reqard to the documentation j
necessary to demonstrate equipment qua' ification.
g g7ED CORRfSPONDENCE:
i Letter for R. Fett from L. Bonron (Sandia) re j
rulemakino and revision to Reculatory Guide 1.89. Acril 3. 1981 review of cronesed f,Q
~
I This letter forwarded Sandla's comments on the proposed revisions of i
NUREG-0588 and Regulatory Guide 1.89, and on the proposed EQ rulem Sandia also made reference to previous comments that wre subeltted by le dated August 22, 1980, and January 13 1981.
i observations with regard to the propos,ed revision to RG 1.89:Sandia made the foll j
j k
Regulatory Guide 1.89 appears to have some Internal inconsistencies.
i l
l F-3 i
l
= =, -..
The Regulatory Guide endorses Itt! 323-74, but the litt standard is currently being revised.
standard that will soon be replaced.Therefore, the standard will be endersin It is not, clear why the calculatten of temperature / pressure sheeld be Italted to a certain time at 340*F (page 4, itse 2).
Extrapolation should be allowed if data is available to support the extrapolation (page 13, item Sc).
Issue Extrapolation of data for EQ purposes. If adequately justified, say be appropriate.
Testing at humidities of greater then g05 is pushing the state of the art. Above 100*C pressurizedautocIave(page14,itsef).all testing would have to be done Issue:
Limitat!ons asseclated with humidity testing may need to be addressed.
page C-1 seems to indicate that all of the profiles have a desble hump, e
which is not consistent with the text of the RG.
Issue:
Use of the edevble humpe that is required in the accfdont h
profile for EQ testing should be reassessed.
The release assumptions seem to be discontinuous (i.e., 15 se11ds release for 30 days or less 505 Cestus release beyond 30 days; page D-4, top).
Issues The release assumptions required for EQ should be reassessed.
Memorandum for f,. Arletto from D. Sullivan.
- Backer w.d Informitten on Pronosed Rev1s1on to Reculatory Gulds t.Re. *invire;- atal Gua if1 cat 1on af Electric foulement for Nuc ear Power P' ants. Januarv 21. le8P This assorandum was issued to provide background informatles and to expedite the proposed revisten to Regulatory Guide (its) 1.09.
The RG was first issued in November 1974, endorsing !(([ 5td. 323-74 (with appropriate supplementary matrcial).
The proposed Revision 1 of the R$ provided more detailed guidance and aise f acorporated the guidance provided in IIUREG-0548, tatilag late account i
comments that were received during the NUREG comment period.
IIe significant comments vers received from the staff that required resolution, but the following significant changes were made subsequent r.o the staff's review and wert highlighted in this aseo:
Additional equipment, other than equipment essential to altigating the consequences of an accident, shecid be qualtfled for accident conditions tf its malfunction or failure due to the accident conditions will negate the safety function of safety equipment.
Analysis may be acceptable if testing is impractical because of size l
l F-4
..a,,
J
~
equipment was purchased prior to Mayitallattens, llalta j
23, 1900.
4 j
11 ef it.e remaining fission product solids was f acluded as part of the 4
a source ters.
I A basis was provided for not granting a general esemptfen free radiat quellfIcatfon te ting for equipment esposed to low level redIatIen.
I j
!ssue A general exemptten for radiation qualtficatten testing for 1
equipment exposed to low-level radiatten any be appropriate l
under certain defined circumstances.
i guallfled by test; the meaning of slid enstrennent j
Guidance for the qualification of replacement parts was provided.
i Guidance was provided as to how the RG should be applied dependin the licensing commitment of each plant; the application of the AG to j
replacement components and spare parts was a)se included.
3 deleted; the guidance for these analyses was prov
)
I j
the Standard Review plan.
sufficiently justiffed.The generfc temperature profile was deleted; the
!ssue:
i The generic temperature profile that was justified for plants that assumed this profile for ogvipment goalf fication j
any not be suffletent.
{
The methodology for determining the quellfication radiatten dose was updated.
selected saterials.Information was added on the thermal and radiation a mitiget1on was provided.A listing of typical equipment /functless neces j
Meenrandum for 1. Aoearwal from R. Clovoh f 5Ntt re:
i Guide 1.89. March L 1982 nr.,wned Reeviaterr 2
The memo discussed a probles that was identified where cable insulation w i
testing at higher dose rates would tend to underestim j
that occurred at the lower dose rates.
1 Issue:
j Dese rate effects are sometimes serpelsing and spm:1fic puldance any be needed in this regard.
k F-5 j
{
4 i
b randum for 1. K. Aeoarua' fram E. V. Marfrf t. "Amentatory Cufde 1 se.
i L. *invironmanual Qua trication af I ac tr ic f au t
.,1 Isaart ant to f a re t, l aw.
Inc Nuclear Power 7 ants'.* Octonse 20. 19a1
{
j The results of A(00's review of the subject regulatory guide were provided.
Issues During a case study, At00 identified fastances where quellflod equipment failed due to himidity or asisture fatrusten.
j Qualificatten practices and the status of EQ in this regard should be reassessed.
Me=arandum for CRct members from M. farler. 'Information Anlated to taca
]
Meetina No.14.
- Anuary IL le84 l
The results of the 0(DROGA staff review of Regulatory Golde 1.89, Revision I, 1
i was provided.
I j
The following comments were made:
a.
The full scope and meant of a "well-defined preventive maintenance program as p'elng essenti to [Q* 13 not clear.
i i
b.
Dose effect considerations may not be conservative given the TNI-2 j
e,,erience.
i 5NL views on [Q based on their technical findings may significantly c.
j tapact issuance of R$ 1.89, Rev.1. A January o, 1984, letter is i'
rsferenced in this regard. for example, Sandla discovered that for the same total dose, lower dose rates age materials more severely than higher dose rates (see UC5 letter free E.. Weiss to Commission dated January 5, 1983).
1 d.
accident can be deterstned and/or beended by curren I
i l
sothodologies Replacement of equipment and the upgrading of equipeont qualification e.
through the replacement process, including those reasons being set forth j
as to why equipment j
replacement process, qualification need not be upgraded through the should be examined.
4 Note:
All of the DEDROGR costeents appear to be valid issues etch have j
already been recognized.
i i
.e l i
l 1
l i
l F-6 t
I mumme um m i i i
~e
--- - - - - - - - - ~~
~ ~ ~
1 armers a I
004 suldelInos i
The guidelines for reviewing the status of equipment qualificatten la i
operating reactor plants, prepared by the Olvisten of Operating Reacters (
1 In the Office of Nuclear Reactor Regulatten (IRR Office of Inspection and Enforcement (!!) in a es)m,erandum for V. Stelle were transe' tied to the i
H. Denten dated November 13, 1979.
1 the 004 Guidelines The guidelines, commonly referred to as were developed to establish evidance for !! te use In reviewing Ilconsee, responses to It Bulletin 79-01 and asseclated qualifi documentatten.
equipment whose documentation was not sufficient to esta I
I assurance of environmental qusilfication, resolution of the identified problems. qualification was questi i
then operating nuclear power plants, including the 5tpThe Guidelines were ap j
include guidance for replacement equipment, plants, but eid not 1
l Class It equipment was defined by the Guidelines as *all electrical equip needed to achieve emergency reactor shutdown, containment isolatten, reacter j
core cooling, containment and reacter heat reesval, and prevention of i
significant release or radioactive material to the environment.
I bring the plant to. :old shutdown condition following Also lacluded resu' t in a breach of the reactor coolant pressure boundary together with a j
rapid depressurtration of the reactor coolant systee.'
i Issue:
Emergency shutdown systems *...used.to bring the plant to a cold shutdown condition following accidents unich de not result in a j
breach of the reacter coolant pressure boundary together with a rapid depressurtratten of the reactor coolant systee* were required to be quallflod by plants subject to the 00R Seldelines.
It is not clear, however, that this was the case for IRREG-0584 plants.
Also, since qualificatten of cold shutdeun equipment was not required by the EQ rule, it is not clear to what extent this j
requirement is valid.
i Licensees of power reactors that were subject to the D04 Guidelines were i
expected to use the following criteria for equipeer.t qualification:
Service Conditions i
from the FSAR analyses or other licenses submittals.In gener jl However, the 00R Guidelines contained specific requirements / guidance with regard to the
{,
following parameters:
}
A.
Service Conditions for a LOCA Inside Containment j
Tennerature and pressure Steam Condittent - Speelfic requirements were specified for pressure suppression type containments.
I Radiation - The normal radiation dose was required to be added to the dose received during the course of an accident.
Radiation service
condittens for equipment located directly above the containment sump, in the victatty of filters, or submerged in contaminated lieufds was j
required to be evaluated on a case-by-case basis.
4 i
i An assumed total gamma radiation dose of 2(7 RADS was acceptable for j
Class IE equipment located in general areas inside containment for pints i
with dry type containments.
j spp11 cation specific evaluation, Doses less than this amount required an k
Gamma dess radiation service conditlens for Stats and PtNts with ice condenser containments needed to be evaluated en a case-by-case basis.
j If it could be shown that the beta dose to radiation sensitive equipment I
l Internals would be less than or equal ts 105 of the total gassa dose to which an item of equipment was qualif ted, then that equipment could be i
plus beta). considered to be quat tfled for the total radiation environment (gass i
i j
e Su wenence - The preferred method of protection against the effects of j
submergence was to locate equipment above the water flooding level.
i
$pectfying saturated steam as a service condition during type testing was not an acceptable alternative for actually flooding the equipment l
during the test.
I j
!ssue:
Aging of equipment seals and esisture barriers may need to i
be accounted for on the 00R Guideline plants.
Containment scravs - Equipment exposed to cheetcal sprays needed to be l
quallfted for the most severe chemical environment which could exist
)
(acidic, bastc. and deelner'llred water)..
)
I
{
8.
Service Conditions for a R, 3 Inside Containment j
Teamerature and Pressure Stem conditions - Equipment quallfled for a i
LOCA environment was considered quallfted for a M5L8 accident l
environment in plants with automatic spray systems that were not subject i
completion of Task Action plan A 21).to single failures (the accept j
Otherwise, Class IE equipment j
needed to be qualffled for MSL8 based on a plant specific analysis.
4 1
Radiation - Same as for a LOCA except that a conservative gamma dose of 2E6 AADS was acceptable.
i l
su w roence - Same as for a LOCA.
Cheetcal Scrars - Same as for LOCA.
i C.
i Service Conditions for a HELB Outside Containment i
Service conditions for areas outside the containment exposed to a HEL8 i
was evaluated on a plant-by-plant basis as part of a program initiated by the staff in December 1972 to evaluate the effects of a HEL8.
This equipment was required to be qualif ted for the service conditions 4
G-2 4
1 i
reviewed and approved in the HEL8 Safety Evaluatfon Report for each specific plant.
1 D.
Service Conditions for Areas Outside Containment that are Exposed to Long-Tern Recirculation i
Teanerature and Relative Humidity - 1005 relative humidity was specified j
as a service condition in confined spaces.
The temperature and pressure as a function of time needed to be based on the plant unique analysis reported in the (SAR.
1 i
Radiation - Radiation service conditions needed to be evaluated o i
case-by-case basis.
In general, a dose of at least 4E6 RADS was expected.
i Submeroence - Not appilcable.
~
e Cheetcal Sorays - Not applicable.
i
!ssue:
i i
i With regard to service conditions in areas outside containment that are exposed to long-tem rectreulation equipment qualification may be needed for chemical spray.
l (e.g., actuation of fire suppression systems).
s 1
j E.
Areas outside Containment that are Nomally Maintained at Roos t
i Conditfons
{
Equipment installed in these areas was designed and installed using standard engineering practices and industry codes and standards.
Failures of equipment in these areas during a design basis event were or failures of air conditioning or ventilation systees. expected except for aging considerations discussed below, no specialTherefore, Class IE equipment in these areas provided that the area i
l maintained at room conditions by redundant air conditioning or ventilation systems served by the onsite emergency electrical power system.
Otherwise, equipment in these areas needed to be qualiffed for the environmental extremes that could result from a failure of the air j
conditioning or ventilation systems.
j Qualiffeation Methods t
Type testing was the preferred method of qualification for Class IE electrical equipment located inside containment.
As a minimus, the qualification for severe temperature, pressure, and steam service conditions for Class IE I
equipment needed to be based on type testing.
Qualification for other service conditions such as radiation and chemical sprays could be by analysis i
supported by test data.
l A.
Type Testing t
Simulated Service Conditions and Test Duration - The environmen G-3 J
W'W
' ' ' ' F
l
~
test chamber needed to envelope the specified service conditions.
i time duration of the test needed to be at least as long as the period The j
from the initiation of the accident until the temperature and pressure i
returned to the levels that existed before the postulated accident.
However, a shorter test duration could be acceptable if specific analyses were provided to demonstrate that the materials involved did not experience significant accelerated thermal aging during the period not tested.
i 4{
Test Soecimen - Any deviations between the test specimen and the installed plant equipment needed to be evaluated as part of the qualification documentation.
Test Seouence - The test sequence needed to model the environment for the postulated accident.
the test sequence provided: Radiation could be applied at any time during (a) the component did not contain any
+
materials which were known to be susceptible to significant radiation damage at the postulated service condition levels, or (b) the saterial's susceptibility to radiation damage was not an " unknown."
j radiation dose needed to be applied prior to or concurrent with exposure Otherwise, the to the elevated temperature and pressure environment.
The same test specimen needed to be used throughout the test sequence.
k Test Soecimen Aoina - Tests that were successful using test specimens which had not been preaged were acceptable provided the component did degradation due to thermal and radiation aging.not contain mat j
If the component contained materials that were susceptible to degradation, a qualified life for the component needed to be established on a case-by-case basis.
Arrhenius techniques were generally considered to be acceptable for i
thermal aging.
Functional Testina and Failure Criteria - Actual operational conditions for the component needed to be established during testing (i.e., if the 7
component was normally energized during plant operation, it should be' energized during testing).
Failure criteria needed to include instrument accuracy requirements based on the maximum error assumed in j
the plant safety analyses.
i If a component failed at any time during the test, the test needed to be j
considered inconclusive for the entire period of the test.
Installation Interfaces - The equipment mounting and electrical or l
mechanical seals used during the type test needed to.be representative j
of the actual installed condition.
needed to include an as-built verification that the equipment wasThe equip i
installed in the same configuration as it was tested.
{
B.
Qualification by a Combination of Methods (Test and Analysis) l An ites of Class IE equipment could be shown to be qualified for a complete j
spectrum of service conditions even though it was only tested for high i
i G-4 s
temperature, pressure, and steam.
such as radiation and chemical sprays could be demonstrated by ana j
Radiation Dualification - The effects of the calculated dose on equipment, based on material properties, could be evaluated.
general rule, the required time for equipsest M remain functional that As a should be assumed for dose :alculation purpo,et should be at least one hour.
{
Issue:
ilhat is the basis for the 1-hour criteria?
i Chemical Sorav Oualification - The effects of spray on equipment that a
was entirely enclosed in a corrosion resistant case (for example) cov1d be shown to be qualtfled for the spray environment by a suitable i
j analysis.
I Marcin The factor applied to the time that equipment was required to remain i
functional was the most significant for providing additional confidence in equipment qualification.
the time requirements stated in the Guidelines for functional testin i
the service conditions specified by the Guidelines. included conservatises
- Also, i
which provided even more margin.
4 no additional margin factors were required to be applied, including the j
suggested factors of IEEE Std. 323-74.
Aging It was the staff's position that the incremental leprovement in safety from arbitrarily requirfog that a specific qualified life be demonstrated for all i
Class IE equipment was not suffletent to justify the expense for plants already constructed and operating.
equipment using saterials susceptible to significant degradation due toTh t
thermal and radiation aging.
Component saintenance or replacement schedules needed to include i
considerations of the specific aging characteristics of the component matarlais.
and maintenance records to assure that equipment which is exh related degradation would be identifled and replaced as neces:ary. g age-i j
Documentation demonstrating equipoent qualtftcation. Cog 1 te and auditable records ne 9
The records needed to describe the had been fully tuplemented. qualification method in sufficient detail to demonst a design specification was not adequate.A simple vendor certtitcation of compilance with Issue:
It is not clear to what extent all test failures (for all attempted tests) were documented the equipment qualification recor,d. evaluated and saved as part of This may be an taportant i
i i
G-s
- - - ~ ~ ~ ~ ~
'~ ~
oO l
consideration since EQ is typically based on a very small test l
l sample.
Aneendices Appendix A Typical Equipment / functions Needed for Mitigation of a l
Appendix 8 Procedures for Evaluating Gamma Radiation Service l
l Conditions 3
i Appendix C Thermal and Radiation Aging Degradation of Selected j
Materials i
j i
gljffD CORRISPONDENCE AND OTHER INF0PMATION:
i Tad Action Plan Ites A-2L
- Main Steam Line Break Insida Containment -
[n untion of fnvironmental Conditions for foulosent Qua ification*
In the past to define eq,uipment qualtf tcation requirements inside PWR contai i
H:vever, preiteinary calculations indicated that the failure of a j
i temperature calculated for a LOCA and, therefore, possibly higher than the temperature for which the safety-related equipment was qualtfled.
1 action plan itse was concerned with the evaluation of environmental conditions This tash that would result from a main steam Itne break (MSLB) inside containmen the purpose of qualifying safety-related equipment.
j The 00R Guidelines specifically credited resolution of this issue for determining whether equipment that was qualtfled for LO*A condition qualtfled for MSL8 conditions (given automatic s; pray systees that are notcould also l
subject to single failure).
j ranking.
Resolution of this issue was given a low priority Memorandum for f. Beckterd from A. Gody. ' Periodic Review of tow-Priority l
Generic Safety Issues.* June 16. 1993 With ' regard to Task Action Plan Ites A-21. the staff indicated that the concern steemed from the assumption that PWR equipment qualification in containment was based on LOCA analyses, and the staff took the position that i
i the equipment qualification analysts was plant-specific and some plants' i
equipment qualification was based on a stone line break in lieu of a LOCA.
The staff also indicated that superheat effects in containment due to MSL8 i
been evaluated and it was concluded that the effects are lo j
concern therefore did not exist.
i 1
Issues:
(a)
A esso from E. Case to the Commission dated January 13, 1978
{
see Appendix C), indicates that environmental conditions or accidents other than for LOCA were not defined (at least j
El power reactors affected).
The staff's resolution of Tast j
j Action Plan Itse A-21 does not seem tn recognise this and therefore, may not be valid.
i i
G-6 4
. ~,, -
i (b)
With regard to the 'superheat effects
- ef a R5LS, the staff's resecution of TAP A-!! failed to consider single failure constgerattens. It is not clear to what extent single failure need not be considered such that *1ecal effects' can be excluded.
should be reassessed.
Staff practices in this regard
)
i l
I l
l l
l i
i 1
)
[
~
_,, - - ~ -
' ' ' ^ _ --
"~~
\\
~-
APPEW IX N WREG-0548 l
1 WREG-0584 (for Comment), 'Interin Staff Position on Environmental i
Qualification of Safety-Related Electrical Equipment fafety-Related Equfpent,' and to provide various staff positions on the sanner in which savironmental qualtf tcation require.aents should be satisfied.
i The WAEG was applirable to safety-related electrical equipment required fo accident sitigation, post-incident monitoring, and safe shutdown, on plants that were not yet oNrating and required to t aplement either the 1971 or the 1974 versions of IEEE Standard 323.
plarts that were already operating. The WREG was not appilcable to reactor i
The positions stated in the NUREG supplemented the guidance found in ti and the 1974 versions of IEEE Std. 3D; Category I positions being appilcable to the 1974 standard and Category !! positions being appilcable to the 1971 standard.
However, the staff's positions as expressed in NUREG-0588 were developed prior to completion of the TMI-2 event evaluation, and it was learned from the TMI-2 expertence.recoggtzed that additional staff positio areas of environmental qualtftcation, su:h as the effects of aging,Th and research activities that were being conducted by bo industry could very well lead to more detailed guidance in these areas.
Finally, WREG-0588 specifically recognized that seismic qualification of electrical equipment was not included within its scope.
establish a consistent, uniform criteria for demonstrat qualification of electrical squipment for plants that were under construction and not yet operating.
i This wa*. particularly necessary since IEEE 323-71 was i
auch less rigorous than IEEE 323-74, and the 1971 standard was very i
content and detall.
Nonetheless, the WRES allowed some laportant differences j
i to exist in the environmental qualtitcation criterla that were established for plants that were comaltted to the 1971 IEEE standard (Category !! plan l
j compared to those that vere committed to the 1974 standard (Category I plants).
Table M-1 contains a listing of these differences.
I i
WREG-0588 was revised in July 1981 to provide the staff's responses to comments that were received to the earlier "for Comment
- version.
Outdelines and WREG-0508:WREG, in its introduction, also stated what th The revised i
operating Itcenses that were issued on or before Maythe DOR Guidelines
{
with construction permit Safety Evaluation ReportsCategory !
I 23, 1980 before July I,1974; and NUREG-0588, Category 1, was(SERs) that were issued plants that were licensed after May 23, 1980, with construction permit SERs appilcable to those l
that were issued on July 1, 1974, or later.
l NUREG-0588, Rev.1, contained the ic11owing staff positions / clarifications:
l l
Imi i
' ~ ^ ~ ^ ~ ~
~ ~ ~ ~ ~
^'
l-l Alternatives or exceptions to the interia staff positions could be i
i proposed and found acceptable If justiffed.
Any exceptions needed ta be identified and evaluated on a case-by-case basis.
2 Note:
i The interis staff positions contained in NURES-0548 did not i
necessarily estabitsh a alnlaus acceptable stendard for qua11fication since exceptions were allowed (if adequately Justified).
The implementation and the degree of conformance to the Cat, !! criteria would be evaluated on a case-by-case basis. On the older plants,
{
backfitting an acceptable degree of conformance to the criteria was t
required where it was demonstrated that additional assurance was j
warranted.
l Note:
This tends to suggest that full implementation of the Category !! criteria was not required.
a further analysis and/or specific component testing would be allowed by i
the staff for estabitshing equipment qualification fer main steam line i
break (M51.5) environments on a case-by-case basis, in addition to the surface temperature method.
(Sect 1on 1.2(5) and Appendix 8 of the WREG)
)
An NRC sponsored research effort was investigating the use of a multi-i level and time-dependent releases of radioactivity from the fuel following the design basis LOCA. The results of these ongoinq research efforts would be factored into the final rule.
(Section 1.4(L) and j
Appendh D of the WREG) 1 The WREG was modified to include the assumptions for modeling radiation environments for BWRs; the "For Comment' version only provided i
assumptions applicable to PWR containments.
(Appendix 0 of the NUREG)
The staff would allow other assumptions in addition to ' radiation at the i
center point of containment" if adequately documented and justified.
(59ction 1.4(6) of the WREG)
}
i With regard to beta dose, the assumption of a sont-infinite cloud was t
allowed if it could be justified.
i NUTEG)
(Section 1.4(7) and 1.4(9) of the i
{
Though not specifically stated, the beta dose to cables free plate-cut i
needed to be considered.
(Section 1.4(9) of the WREG)
Cat. !! plants were r.ot required to demonstrate qualification to beta j
exposure as required for Cat. I plants. Instead, natorials that i
exhibited adverse effects to beta exposure needed to be addressed, and replacement equipment or equipment that had r,ot yet been qualifted 3
needed to conform to the Cat. I requirements.
(Sections 1.4(7) (8),
(9), (10) and (14) of the WREG)
Qualification of equipment to low levels of radiation could be performed H-2 wk
-.~
a Rhlysisthatwasadequatelysupported. (Section 1.4(11) of the 2
Only equipment required to mitigate the effects of a high energy pipe i
break CHEP8), or whose failure would be detrimental to safety, needed to i
be qualifted.
(Section 1.5(1) of the WREG)
Non-safety-related equipment did not necessarily have to be quallfled by i
test, though type-testing was the preferred method for active j
components.
(Section 2.l(3) of the WREG)
Equipment that was shown and justified to be unimportant to safety, was j
not required to be qualified.
(Section 2.!(3) of the WREG) j Establishin acceptable.g ' acceptance criterta" instead of " failure criterta' was (Section 2.2(1) of the WREG) i i
Exposing the same component to a combined or sequential LOCA and MSLB 1
additional requirements of doubling the number of transien
{
J (Section 2.2(4) of the NUREG) 1 Measurement of " bulk tewerature" rather than " surface temperature
- was acceptable; the intent was to demonstrate qualification to the accident enviro m at, (Section 2.2(6) of the NUREG)
}
Shorter test periods in conjunction with analytical extrapolation was acceptable for submergence qualification if adequately justified.
(5 action 2.2(5) of the NUREG)
For situations where undervoltage and underfrequency simulation duri i
seismic testing was more severe than for other hostile environmental j
conditions, the basis for excluding these conditions during testing of the other hostile conditions needed to be documented.
j of the NUREG)
(Section 2.2(10) i I
!ssue:
The most limiting unk rvoltage and underfrequency conditions plants (i.e.may not have been assumed for qualification testing for all conditions)., seismic vs. other hostile enytronmental j
equipment operability in dusty environments, but ra l
potential failure mechanism.
Some consideration was required for i
maintaining cleanliness or for operation in dusty environments.
(Section 2.2(11) of the NUrtEG)
Issue:
Specific asasures may be warranted to minimize the impact of 3
dust on equipment qualification.
applicable information was already available from te was essentially identical.
(Section 2.3(1) of the NUREG)
i Additional margin did not need to be added if it could be shown that e
adequate margin was already included. Margin was needed to account for uncertaintfes in analytical techniques, uncertainties associated with testing only a small number of units, variations in cosmercial i
production, and inaccuracies in test equipment.
(Section 3(1) of the NUREG) 4 3
Issue:
Overall margin requirements need to be better defined, with i
supporting technical justification.
i An interim position for determining the velocity profile in containment e
i during blowdown was provided.
upon completion of Task A-21.
A final position would be established (Appendix B, Item 2b, of the WRIG)
Issue:
1 Since the staff's resolution of Task A-21 was incomplete J
the staff's final position on the velocity profile may ne,ed to be re-examined, s
calculation of the qualification dose for equipment; usi 4
j multipitcation factor of 1.3 was not sufficient.
(Appendix D. Section 7 j
of the WRfG) 3 plate-out on centrally located equipment could be a significant
]
contr hutor to the accident dose and should not be ignored.
(Ap; Mix 0, Section 7A of the NUREG) i The amJe1 incorrectly calculated the gamma doses in the vicinity of the a
wa;1.
The doses near the containment walls needed to consider the sMtw (Ivx from all sources.
The NUREG was modified to remove the
- ndrsi.
Guidance was also provided to use gamma buildup factors.
(4pe+ dix D of the NUREG) i the qualf fication dose required.Dosts from ECCS equipment leakag (Appendix D of the NUREC)
RELATED CORRESPONDENCE:
Letter for W. Rutherford from L Bonron (Sandia).
Revis f on 1. Draft 1. dated November 1980." January' Comments on NUREG-0588.
- 13. 1981 i.
Sandla provided its comments to the draft revision of NUREG-0588.
Sandla commented that the revision was significant such that another round ofIn g industry comments may be in order, and the revision did not address many of i
Sandla's previous comments on the "For Comment" version 'of NUREG-0588, dat December 1979, and discussed in Sandia's letter of October i
10, 1980. Sandia noted that the previous connents' concerning "inspectability" were still 1
generally applicable.
spp11 cable to Regulatory Guide 1.89, and referenced a Sandla let j
August 22, 1980, which previously provided Sandia's consents on the regulatory j
golde.
While many of the Sandia comments on NUREG-0588 were editorial in nature, th i
i j
H-4 4
i I
following are noteworthy (coments are applicable to Cat. I unless otherwis noted):
Safety-related mechanical equipment should be included in the de finition.
l (Page 2, paragraph 5 of the WREG)
Issues More detailed qualtftcation requirements may be needed for mechanical equipment.
calculating the pressure and temperature envelopes Cat. II.
(Section 1.l(2) and 1.2(2) of the WREG)
Issue:
Justification for different qualification standards should be established.
As presently written, MSLB quaitfication is not assured because the LOCA bulk temperature may be much higher than that achieved at the equipm surface during the LOCA.
In this case, comparison of the calculated MSLS surface temperature to the LOCA bulk temperature will not assure j
that the equipment could survive the MSLB environment.
of the WREG) i (Section 1.2(Sc)
Issue:
Comparison of the calculated R$LS surface temperature to the LOCA bulk temperature may not assure that the equipment will 1
survive the MSL8 environment.
i
)
where the calculated surface temperature for MS
- i qualification temperature.
k (Section 1.2(Sc) of the WREG)
Issue:
Justification for different qualification standards should be established, i
i The aging methodology is not properly handled in that it does not e
I consider the possiblitty that aging could im Class IE component to perform during a DSE. prove the capability of a (section 1.4 of the WREG)
Arbitrarily chosen release fractions are inappropriate.
1.4(1) and 1.4(2) of the NUREG)
(Sections i
Issue:
{
The release fractions that are used for the sourcs ters may not be appropriate.
subjectedtothelargestsourceters. Equipment expected to survive will ba ottigated and the large source ters will not o (Sections 1.4(1) and 1.4(2) of the NUREG)
Issue:
A large source ters, given accident mitigation considerations, may not always be appropriate.
H-5
~
Heat due to beta radiation should be included in the calculation of the temperature profile for which equipment must be quallf ted.
(Section 1.4(8) of the NUREG)
. Dose rate may render instruments inaccurate before total dose effects become '.:.ortant.
(Section 1.4(13) of the NUREG) q
- Integrated beta and ganna" doses do not include rate effects.
(Section l
l.4(13) of the NUREG)
J There are a number of instances in NUREG-0588 where the qualification methods of IEEE 323-74 are not appropriate.
(Section 2.!(1) of the NUREG) i IEEE 323-74 allows several different qualification methods including type testing, operating experience, analysis, and ongoing qualification, as well as a combination of the above. This is inconsistent with j
Section 2.l(2) which requires test data.
(Section 2.l(1) of the NUREG) t The staff's position requiring test data contradicts the intention of i
IEEE 323-74, which is supported by Section 2.l(1).
(Section 2.l(2) of the NUREG)
Issue:
It may not always be appropriate to require test data to establish EQ, given the other qualification methods allowed by IEEE 323-74.
This section asserts that equipment is to be qualifled for the time required for it to function. This is different from Section 3(4) which i
requires a minimus time of one hour.
(Section 2.l(3) of the NUREG)
I i
Issue:
The one hour minimum operating time requirement may not be appropriate.
t' This is not consistent with Section 2.1.2 which states that the staff will not accept analysis in lieu of test data.
(Section 2.l(4) of the J
l NUREG) i for auch of the equipment, continuous monitoring or verification should e
be performed during testing, not periodic.
a (Section 2.2(7) of the i
Justification of the test sequence selected (required for Cat. II) should also be emphasized for Cat. I (even though IEEE 323-74 makes a i
loose reference to sequence justification.
demonstrate that the sequence may be very)important, especially d Recent Sandla data i
accelerated aging.
(Section 2.3(1) of the NUREG) 1 Section 6.3.2.4 of IEEE 323-74 states that:
"If the required radiation 4
level can be shown to produce less effect than that which would cause loss of equipment's Class IE function, radiation need not be included as 4
j part of aging."
NUREG-0588 should not allow this blanket statement 4
a H-6 m-
4 because it falls to consider synergistic effects during the accelerated aging program.
(Section 2.3(1) of the NUREG)
Section 6.3.2.4 of IEEE 323-74 requires aging to the end of quaitfled Ilfe.
This is not consistent with the newer standards, in that the most limiting condition may occur before the end of Ilfe.
the NUREG)
(Section 2.3 of The use of other qualification test methods contradicts Section 2.l which states that "...in general, the staff will not accept analyses (2),
lieu of test data..."
in (Section 2.4 of the NUREG)
The guidance provided by NUREG-0588 and IEEE 323-74 concerning is sometimes unclear or too severe.
(Section 3 of the NUREG)
Issue:
Specific guidance with justification needs to be established with regard to the margins that are required by IEEE 323-74 and NUREG-0588.
The margin requirement for thermal aging is too severe.
For many organic materials, such a margin requirement will roughly double the accelerated aging time.
(Section 3 and 3(2) of the NUREG)
Issue:
The margin required for thermal aging should be justified.
l defined in 323-74, another word should be used.The " mar take care of " production errors" but to take care of productionAlso, margin is variations.
(Section 3(3) of the NUREG)
For some equipment testing for an additional hour may increase the i
radiation exposure, requirements from less than 0.1 MRads to greater than 7 MRads. This does not seera reasonable.
(Section 3(4) of the NUREG)
Issue:
The one hour sinteum operating time may not be appropriate.
The wording of this requirement suggesh that it is not the industry's e
i responsibility to test for synergisms, rather, they must only determine if someone else has identified a synergistic effect.
requirement, samples should be aged via two methods and at varying do As an initial rates:
(a) radiation aging followed by thermal aging, and (b) thermal aging followed by radiation aging.
(Section 4(3) of the NUREG) f H-7 khNEM
_~
I DO e
TABLE H-I a
{
DIFFERENCES IN ENVIRONMENTAL QUALIFICATION REQUIREMENTS BETWEEN NUREG-0588 CATEGORY I AND CATEGORY II PLANTS 1
CthfRAL REQUIREMF9(T Diry[eTNCf fy cart [ eta Quebnesten parameters s
1.
LOCA Terroweture end Prosauro Conesens insede Category N plants were esewed te soeume perud reveportassen: other j
Centenment for PWRe with neeveremons thet reduce the terryerature roepense of the sentesernent were i
Dry Centenments eine esewed on a cese-by case bees.
t j
2.
htSLS Ten, eture end Pressure Cenesene inside Where oushf'ceden had as, been sempleted. Cat. N plette were eA*wed to e-estatash terrowsture and preeewre eenseene bened on seleuledens welag a 4
pien<speerf.o medd and tne sterf approved eseumpsons sentained in I
Appenas a sinn il of the NWAE0; the staussen whwe guetAeeson had j
streedy been swryleted wee not addreened Cet. I plants wwe roodred te estatAsh terremeture and peessure eenesens by veing a pime spesine model that wee redewed and approved by the etsff.
l ff the ceiculeted surfeso tertveredwre of equipment eseceded the webficonen temperature, Cat. E plants were esewed to prev 4de addtlenaf pstficenon te demonstrate that the " w would be etde se perform ito 1
i function. =hweee Cet. I piente wwe re, dred te either perfwm ressehficecon tee 6ng weing appropdate maryne or predde phyeled protecton for the equipment.
3.
Environmensd Cenesons Outade Centalronent The some techniques thet were esemed for soloideting sendtlens inside
{
commnment f or LOCA and he$La woro amewed for easedeeng eend#ene eutendo contenment, in the sees of the Cet. I plante the technMques wood were revnewed and approved by the etsff. in the sees of Cat. N plante. the techremsoe used were not spoo6Acady rev4ewed and approved by the etsff.
}
but the stoff approved esaurvysene sensened in Appends 8 Stern il of the f4UREG were impeeed.
For Cet. I plante. equipmerit not served by Class IE d 2.
.4 support eystems. or served by Claes IE eupport systems that seuld be soeured dudng 1
j pient opwstion or during shutd.wn eenetions, wee papacted to be guehned to the 1.rrutmo em4tonmente eendtiene ahet were posedeted for that l
loceton, soeverung a leae of the em4ronmental enoport system. Cat. N i
piants were pven ed60ensi Resitsty in that moretedng devices were
{
eA.wed to be used to siert the operatore to abnormel omtrosvnental corduone such thet appropriete eerrec9ve ac5one sendd be taken, Quahneseien heethode:
i j;
1.
Selecten of heethod For Cet, a plants. the queefles#en mothede were espected to eacderm to the requ6remente of IEEE Std. 323 71; wheroes fee the Cat.1 plante the i.
webf.ceoon methode were espeeted ire eenfwm to the roodremonte et stEE j
Std. 323 74, white the speaficeeen methods were genere#y the same. the
}
1974 vers.on of the standard wee trarch more specone and riperenta then the 1971 vers.on.
2 i
2.
Ovabf eDon by Test j
Cat, il plante were espected to see the items desertbed in Sween S.2 of if Et Std. 323 71 es guidance for estateereng teet preceduree: Cet. I pienes were espected to use the items deoenbod in Secuen s.3 ef IEEE Std. 3 23-1 74 es guidance for estatdefeng test procedusee, l
1 In poneral, the Cat. I repairemente are much more rigorous and dotated.
especietly with togord to emapment mounong and connectone, reesteen 1
4 4
i, H-8 i
i i
TABLE H-1 (cont.)
t j
CfNf R AL RFOt/tRflAfNT D*Ff ef Nef fN cRfTERfA i
2.
j ovencemen by Teet leentJ espeeure, teet monitoring, and erwesen www noemd sad esadent 1
conditsene.
j; Aloe, requirements that were 64eeed en Cet, I plants that were not opecdcolly erposed on Cat. 3 piente inehaded speedoesen of the nurreber of l
urete to be tested end requiremonte releeve to og6ng. morpne, test eequence, weiwetjen, and M %
3.
Test Sewence i
Cat. H plants were edad to predde justHieeWen for the adooseey of the test oomeence thet =as seiooted wide spoone paidenee wee setebbshed for Cet.Iplente, i
For Cet.11 plante, eeperate effoote seedng was not eenadored to be i
ecceptet>le fet wetet electrical epipment owch as penetro6 ens, eennectors.
cetdes, selves and eneters. and trenorrettere Secoted inside sente6nrnent er
{
espeeed to hoeide eisern ernerenmonie outswe sentai enent. For Cea, e plante, esperste effecte tweeng wee not en opt 6en due to the oogsental i
{
testmg gudence that was esteesehed in IEEE Std. 323 74.
4.
Qudneesen by Opereeng 4
i Empee6enee er by Analywe Cet, a plente were held to IEEE Sed. 323 71 and the enemiery standerde that wree endorsed et the time the Cp wee leaved: Cat. I plante were held to 3
j 4EEE Sid. 323 74 and the ensifery etenderd that were endorsed et the ame !
1 the CP wee isewed.
i
{
IEEE std. 323 71 le wwy general eerwered to the 1974 verwon Therefore 1
i=
ouelification by ensfreis or opersens empenonce for Cet. 5 plants weedd tend is b. very waist. we ow, ed = = more ei,-.we and ensie e4dence established by lEEF Std. 323 74.
Other Cenederasene:
1 1.
ungine i
l soih Cat. : and Cet. a pients wwe espeeted to ente margine. Hew.ww. in r
the cose of Cat. I plants. epoede pedance wee provided for " A-ecceptetdo morpns; for Cet. N plante the marpne prow 6ded in the dee6pn were esolueted on e case-by coes bee 6e (no speeds criserie wee estab6ehed).
1 For both Cat. I and Cet. E plante, whom the esehAeedon onweispe of
{
Appendia C of the NUREO wee weed for oostament inside sentelament the ordy additsenal mergine that were requered were those eseewtWng for 4
inaccutsc.es m the teet eespment. Hewever, Appendia C required Cat. I j
4 plante to espose equepenent to en esbitrary trono6ent emerosenent with peak j
conditiene leeting for ten mireates before returene to the inruel stereng j
i conditions. end then e pocond trenoient wee irgeoed to ammusste the eccedent scenario. Cet. Il plants were ordy regered to espose acepment to the second trer>sient sendtion.
{
2.
Aang l
j For Cat. i piente. eging eefects.n es e-;ua woe espected to be cormdered and included M the gudficamen program. For Cet. N plante.
i eyng of f ac te were not espected to be incewded M the spaebfbesten progra I
except = hen committed to for e epoesfie servponent wie enn of the enesitary standarde. For othee Wr.
- 4. Cat. E plants were espected to addrese 4
.png erv is th..mnt that e,meni thet is compo.ed. in p t. of j
v rnetonele oueceptstde to eging effects wee empoeted to be edlentded and o echedule for penedicacy repleeing the equement endic* metene6e wee i
i expected to be seteedshed.
i I
i 1
3 H-9 1
4
TABLE H-1
("ent )
GLhL%MWnLMLM Q:!!LEL%L:hL&lL?ah
~ iso, e.
...i n..........e...,,,,,n..u.,..,............#.....<.<..
i a,,..
- a. i
... ".. i. e... n..a
-i....... a..n av p..a..........i..a...
-..,~.a......e.,w4...,e............n........w
~e n.a., n.ei........ 4... i.~.+.pm.ai ai,.......a
.,e i
e m. p... ei.....a e n.., a. p....-
c i i p.ai.......p.ci.e i.....ua.a
- w eaned.'.*+.<. e a*
J
.n.,.
c.
isp.ai.... wi L
g
..,...i.ed..a.d av t he Nun t o '.' b.ia s -. e.com.,i.i..a..p
.i..a.
1 3
o..
ai....a ih. C.. i ad in. C.i et p ae.
I i
s k
)
i dj 3
b i
s e
d H-10 666
4 l
APP (NDl! l It(( 5tandard 323 Ihe intt it ut e of Iler t r1r al and ilec tronic (ngineern (l((() it andard 323,
)
- l(11 trial Une Standard:
General Guide for Qualifying (last l [lectrit i
Iquipment for Nuclear Power Ceeerating $tationt ' tilwed in April 1971, j
provided guidanre for demonstrating quellflaatton of electrical equipment i
Although lill St andard 219, *(rliert a for Nuclear Power Generat ing $t a'lon
]
8' rot et t ion Sys t ems,' and lifl 5t andard 308, *(rlteria f or C last if tiettric systems for Nutlear Power Generating 5tationt,' required equipment t o tie "spsa l i f t ed, ' specific guidance was not provided on how this should tie t
ai < naip i l shed uni l l l'c l 373 11 was titued.
The ters ' qualification
- was l
.5 fined by illi )?) is at t he " demons t r a t t on t ha t equipment mee t s de s ign eequirements lhe itandard required that the qualiflaation of IIatt l eisi t r is equipment Intlude Identiftietton of the (lass l pleiteis equipment he'ng qualified i
pteparatIon of equipment spot i f ti at loos adequale for
'he app ! is atton, 1
demonsteation that the equipment or tempunents increof are eapable of I
meeting the perfurwance spotIf1(atlons under ihe spot if led servl(e iondtttons, and doi umentation suf f it tent to permit an independent evaluation of the elutpernt qualifitatton i
i 1
lill 313 fl provided very general guidance with regard to quellfication method j
ti e type test, operating espertante, and an'a lyt t t ) and dor sment a t ion l
'equirements Although type letting that t mulated the service tondillons was the preferred method, partial type letts augmented by analytlt and/or operat ing esper tence, or analytit and/or operating espertence without part ial type testing that clearly demonstrated the required capability, was allowed for situations where tise or other practical limitations precluded type testing the standard required an evaluation for each modifftation that was j
made to quellf ied equipment to determine whether or not re quallf tr at ion was i
i eequired the analysis or dat a and evaluat ion that demonst rat ed the ef f ei t s f t he mod t f it a t ton on equipment aerformance was required tu be added I 4.
Ihr I
qua l i f it a t t n driument at t on
.I in l ebe ust y 191a, 1111 $ t d
!!! was revlied and a new litte was as ignea
- llit standaeo for Quellfying Llait 11 tuuipment fu Nut l e a r Powe r (,ene v a t iny j
Stations
- lhe trope of the st andar<1 was expanded to thclude ' component s or equipment of any interface whoto failure cauld adversely affect the performante of Ilast 1( t y t t oms and elet t ric equipment.
- lhe definitton of I
- equipment quallfiration* wat c hanged to *the generat ion and main.enant e of i
evidente t assure that the equipment will.perate on demand, t o snee t t he sritem perfo wance requirements
- At was *.he case with l[([ 17) 71, t t e e evised st andard recognited that quellfical ton may be accoleplished in severa t i
wAyt, in(luding Iype let{ inQg operatinQ eMperlen(e, and/or analytii, a nti the limit a t ions of eac h met hod were dist un ned.
Operating esperience wat cited as bring must usef ul f or quellfic ation of equipment located outside s sntainernt Qualif ic allon by analyllt was recognlied at most uteful for extrapolat un n' l
^
i test ilate and for atid re t t i n minor denlyn Ihanget llll std
)?l '4 ineluded j
and not
,v t Idreed io !-
,a'I of l 'i o J
Appendit e t that was e f or 4*.
at e
t l a tid a r d The appendic es prov ide t emple toit profiles for pressurlied water 3
era; tort, bulling water e a. t oe g, and fur hlQh I empr e a t ur e ga s 2 avled l
eesitors. and informallon was peut ided on test ahamber moisture'sontent in y e t' e r. l. lill !?) 74 was muth more det ailed and comprehent ive t han t he earlier vertion of the ttendard table l+1 is a (omparlton of the 1911 and 1he 1974 ret s ions of the llll standard RLLA!LD LORF.LLPONDLNL1 Neporandum {st L.J or11er_Roa.i JAnautt.str.Lt1Alna. Ls JLLL itandard JR.1L Jwl2 II.1511
% Hanauer stated the following opinion of II([ 5td 313 1)
- I iannot find a single redeeming feature in this worthlett doc umen t far feum heiny what itt title suggests, it tontains only the most general kind of stuff on how to quellfy something anything the body of the dut ument it not even spei t f si enough to be related to elettrlial equipment furtherwmre, the v ar ious ( l auset se e to general that it*t ententtalty impotstble 1o determine
.ompttante for t he s e re c tons t he re f e r ent ed doc ument in itt pr sent form it, as I ssid a tsu v e. without valus
'hDUf CdWRL ICLL.501110.f.IIDa P l.. Ltnna A_.*VL L Pti1119ft ! ut a
NtLJ'utdttillWh.' AWQual 19I9
( opp i s t i one r Kenned, roQae tt ed t ha t the staff assets the effert* of leai s f i t t ing Ihe 1971 and the 1974 versions of 11([ Std
- 11) on Ine lieenseen
'pe,iIis sp.ettions
'n this regard were othed. and the tIaff wat iequested to peuvide itt a s puri s e es o 1.i t e s than Aagust /l, 19f9 ds":'2 r Ada f e r f. om t n t oneL K enn ed y from H. L Denton *LK5 Pei M 1on !2I
'tT,0^1'JtrollGQ.* Auguit 1(. 1913
't.= staff peuvided ett entetietent of t he e f f et it of ba( k f It t ling the 1911 and the 19'4 versions of Ill! std ??) on licenteen at requented by C omp t t t i one r Kennedy in his memorandum of August 7 the f ollowing inf ormation wat provided lhe st af f provided a comparlton of the 1911 and the 1974 versions of
+
illi Standard 12)
Ihe print tote dif f erent et that were atied betweer*
the fiou standardt was that the 19< a st andarc provided mut h amre det all tnd itutuded raQutrementi for aging, margInt, and dot ume nt a t t on that i.e e not I nt l utied I n the 19f1
- a. tai.!ard Howsver, both s t a n d a e d.,
coquired guidelines for estestrer itit erpret at ions and juogement t Ib, stalf noted that current r eviews of pl ant s re f erent ing t he 19 71 g t anitard
~
wete beinq ronducted to 45 to bi*ing the level of assurance of equtpment qualif tt it non in those plant s to entent tally t he s ame level
.6 t h e.
1974 ttartIard would athieve Iiffy eIQht of the 70 power reasto,,,arientir litented to oppiat=
e
( i.it l eding indian Potot I ard Hueteldt it a y ) enade no spet i f it reiet et t e t2
L e
to l((( Std. 313 1911 at the bes t s for equipment qualification.
The licenseet for the other if t'antt h ad c ommi t ment t in plare to rempty with the standard, the DOR Guidelines that were being developed by the staf f would provide e
a level of confidence essentially equivalent to that which would be ac hieved f rom the applic at ion of I(!( $td it) 14 However, the 1
Guidelinet would only require t hat aging be c ons idered f or t hat equipment known to be tutt ept tble t o signif 1( ant aging a f f et t t the need for further bat t f t t t ing of aging requirement s would be r e at tes ted at the staff *t underttandtag of equipment.;tng improved Advantaget t hat could De re allied by bac k f it t Ing t he 197a standard would int lude a more rigorout and t emplet e demont t r allon that aging affeitt were adequately attounted for and would require that a qualifled life be established for all safety re l a t ed equ ipment Iurther, margint would be applied to all test p a r ame t e r s wh e r e a t the Guidelines would require only that marging he applied to the mogt glgnifisont p a r' a me t e t t {e g i i me )
A s lyn t f it ant disadvant age was t hat many litentees would be required to retent some frattton of their e qu i pment t o c omp l y wi t h the oging and margin requ t reme n t i o f l((( !?)- 14.
l t wa s t he s t a f f 's v l ow t ha t, when c ompared t o t he dr a f t putdelines to De used in the upcoelhg staff realtettaent of operating plants, neither the aging nor the sargin requirements of l[([ 5td. 323-?4 warranted backfit The benefits of backf ttting these requirement s would be a toall, unquantIflable increate In the level of alturance that equipment in qual t fled as t ompared to the significant cost s that would be involved Issue Given this view, 1(([ 323-74 may not be warranted or sufficiently justified at a necettary qual t f tcat ion st andard
'or power reactors, regardless of when the Coastruct ton Persit $th was titued l
l k
l
!-3
d TABLE I-l COMPAAIS0010F !EEE STDs. 323-74 AND 323-71 i
s e n stawo4eo 1
in seAvoam i
ogggg G.,.ee s no.eeemea.o evee e
a, 5,
.% ei. x See N,-em.ae.
w ag
- e. weve e arweseoeweee te wetage. heeeeenst. erg ee e
e..aeieseeea e > e wm easac e es eso. gen see
- s e.a.e e nosee.a
- p. ma==ee.vvece he seeteeng ewee o emhee boo Ge,we o eeeneemeane 8en ev,e S,.ehe esee,es,w.e eegereag a
2
- es.e.e es h hsme.a see.. e
,.ee e.,e.e e.e.ee.
ei.
. ee...e e,,e....as same-eae, o.e one to ewag e
a.aee
.. so.e m e.
e e,c e e
o.,, mea, a.ese o
I e
sees somn ond detened aweta ce SpeeN ond detes ed eewomease l
J Ce.s um.ae s trae speceN *ese.ement to meatona a
s ee erNeeens ewoohse sea metaede i
e.cwe'ieatom.a b e hemeeamone ee c%ea ma
- =
= teee seneem ieet psen.c she e, et i
4 ' e. t een Nee,
.e.a 5,e.A. i e., sea,as wdmg j
.e e e e,em.ais e.e we.
.e %
se mow tas a
ti senapetene toseewerbane se m esmag j
si meewtenag ei merya i
di teos peo,eace e s t eet seeeense I
e 4eemea epag e < eest a 4.ce ae.ee at oe.* et,ee N s edwesea ai es s es,eence c aion e
.ea e ewe.cm a wet.ac wo.4 e..
. oc c seisac e c e.,one ei epag I
e6 meepa i
sa.aepecten
)
1 Gee ses s,was et peeceowee a ehe 5eec N.ma emeate r
t h ee.ag l os.a ase j
hem of oete eeeeeem.aes j
j Owenheeseea 4Aeitwo i
Geases owman et p.ec eow e.a ehe s
e L eo N ei a o, ameev,ee Spec.N eem,...awnee teem of dote esoseememe
es eow s e
Romeeemeate noe.acewded
$p.cefic eew.e emente
?
O* Ge=ag CausseNewee 9eoseewee Rosetemeate not inceweed 8 Saeueied Sec.=ce Cecos.ea Descapinea end ' gweee show.ag t pp.s e ' me d Poe'ee es merpa teos.e.eael poes toeepeeptw e peseewee end umes e
d ei ope est.ea s peesev e end *ee'veestw e
ee di teme pee ed l
e l
1 l
1-4
)
)
APPCNDl1 J Compartten of (Q Review Celteria
'he NR( revtew criterta for env ironment al qual t i t t at t on ((Q) of electrital equtpment hat evolved over time at the s t a f f ha s bec one more aware o f t he problem and a s mors het been learned la the area of 10 through on90 tag regear(h attiv11les Plants that were litented before May IJ, 19so, were revtowed to the riteria contained in the DDR Cutdeltnet. *Culdeltmet for i v alua t ing l aw t ronment al Qual t f tc at t on of Cl a n t II (let t r 1( 41 l aulpment in Operating Rosstort
- dated November 1879, while plants licensed after May 2),
1980, we r e re v i ewed t o t he ( r li er t a c ont a ined in NUR(G 0564,
- Int eria 5t a f f i nv t ronment a Qualiiit at ton of Sa f et y Relat ed llet t rl( al Positinn on those plants li ented after May 23, 1980, the onet with iquipment pf t
r oms t ruc t inn pe rm i t Safety Ivaluatton Reports (5[Rt) titued before July 1, 1974, were rev lowed t o t he C at egory ll <riteria of NUR(G-0588, while the other plantt were revtowed 'o the lategory 1 (rtteria.
fhut, depending on when the tonstrustion perwit and opwr at ing i tr an te were lituod, power reac t ors were re tewed to the (riterta that are t d in either the DOR Guidelinen or in NURIC 0588 lable J I containt a f ate ly comprehensive comparison of these eiteria, and some of the more obv tout dif ferences include:
s The trope of the DDR Culdelines did not specifically include arc 1 dent monit or ing ins t rument at ion or non la f ety rel ated elec t ric al equipment that could impact the operablitty of safety related electrical equipment A s t r t( t Interpretatton of the (Q rule seeen to indicate inti sny of this equipment that wat not previously required by the staff to be qual' fled in attordante with the DOR Guideltnet, would have to be qual'fied
'n at c ordanc e wit h the provin tons of 10 (IR 50 49
- However, it it not > leer that thit was actually the case l
fontideratton of onidation gat diffusion effects was required by hvRIC 0588, Category 1, but not by the other criteria.
l ihe tonditions that were required to be simulated during qualtftcation i
testing varled signifitantly among the three (rlierta Ihe DOR r,u'delines required tubmergence and themical spray (when appiltable).
NURIC 0588, lategory 11, added fire, water, teltmic, and dutt; l
NURIC 0588, lategory 1, added humidity, vtbrations, jet forces, chemical j
c ompetit ton of the ambient anyironment, espected mechanical wear, and ele t ric al cont ac t degradat ion NURfC 0588, Category 1, required a " double peak
- during accident 1
timulation while the other criteria did not, i
NURf G 0588, Category 1, required the equipment qualification temperatures to be established by direct thermocouple readings, whereat NOR(C-0588, Category !!, allowed ans'.ysis based on thermocouple readings the MR Culdelines dio nnt include a specifl( requirement in this regard l
fett spe(tmen requirements were not spectitcally addretted by i
=
NURIG OSAA, tategory I and ll, whereat t he D()W Guide l ine t stated that o
l t he t est spec imen should be the name model and ident ic al to the equipment he t og quelli t ed
'e st monit o ing requirement s were not spec i f ic all y addretted by the DOR Guidelines. whereas the other c r it eria inc luded requirement s in t h e i regard NUW(G-0$88, lategory 1, included test sequence and f unc t ional requirements (e g inttlet inspection, normal operation f or baseline data, operatton to the estremot of the tpecificat ions, art ific ial aging, esposure t o n o rma l me c h a n t ( s ) and lettelt vibrations, operation durant design batin event, operstion post accident, disantemble, intper t, anc!
record findings)
Limilar requirement s were not spec ific ally stated by the other criteria the DOR Lutdelines did not require a quellflod life to be established provided Ihet the component did not cont ain materialt t h a t we re k n nwn t o be tuliepttule to "tignifitant* dsJredetton due tu thermal and radiation aging NUR(G 0$88, Category ll criteria required replacement schedules t o be est aDi t thed f or equipment containing materials that were subject to aging NURIG 0588, (stegory I criteria required the qualification l
tett plan t o inc lude aging t imul at ion, In(luding nynergittit effectt j
ti e radiatinn added t o ot her known degrading influent e t, tus h at i
temperature and vlbration, elec t ro mechanic al agulpment needed t o be l
operated to simulate wear, material phate r.hanget needed to be taken
+nto asiount, ett
),
to sitoblish a quellfled 11fe for the equipment l
The lok Guidelinet did not require any additional margins beyond e onter w at isen that were int luded in te st prof t le NuklG 0$n8, iategov, il <rtierla stated that margtnt would be e v a l u a t ed on a ease ti r a ase l si is, and imposed a almimum qual 1f1(atlon t ime of one hous NUNIG OL88, lategory I ce iteria listed specif ic telt margin reQulrement s for temperature, pretture, radiation, voltage, frequency, t i me,
i I
t'anstent 4ondittons, and vthratlon, margInt were alto required for ear h agisg tefluente NURIG 0$88, Category 1, was the only criteria that st at ed spec ifl(
requirementi for pont tett ditattembly, Intpection, letting, int luding appropriate documentat ton of post tett conditions NURIG 0588, C ategory I and ll, both provided instructions for spec i f ic l
doc ument at ton of telt results, while the DOR Guldelinet did not NURI G 0588, C ategory I and 11, both provided Instructions for the use and applic at ton of operat Ing esperience at a que l l f ic a t ion me t hod, while the DOR Guidelines did not The (04 Guidelines a llowed t he use of ana l yt t s t o e s t r apol at e equipment qualificatton based onl y on high temperature, pressure, and steam type f(st data (e g, quellftsation for radiation and themital spray to,14 be demon s t r st ed tar analysin) lhlt degree of flesibillty was not allowed by NU H l (, Otan, tstegory 1 and 11 Ihe IX)R Guidei nos required un going progeamt t o lie in plat e to eeview s u e w e t ' ! a ro e and maintenanie r ei ne d t in nrder to identify and manaje J?
l
\\
1 age related degradatlon.
NUR[Q Clet, Category I and ll, did not specifl(ally require on gotng review of maintenant e and surveillance recordt, but NUR[G Olla, Cate9ery I did require that a preventive maintenance schedule be sitabilthed for the installed life of the equipment.
Portodic replacement of equipment known to be susceptible to aging was required by NURIG 0588, Category 11 NUALG 0%58, l at egory I and l1, required an evalual non of modif14 at lons to be per f ormed to determine the e f f ect on equipment qualif ic at ion, wh il e t he DOR Gu ide l ine t did not require an evaluat 10n of modi f it at innt S ini e 10 ( I R $0 49 is silent w i t t.
r e g a r d t o e qu i pme n t mod i f t s s t i on s.
Ibere may be some que s t t on a t to what the spot i f f t r e qu i r eme n t s are issues-(a)
Differences that,utit in the [0 review criteria established by the D0R Suld61:nes, WOR (6-0444 (Cate9ery 1), and WUR[4-0488 (Category !!) should be justif ted.
(b) the qualification of electrical eeut poent for main steam line break condittent inside containment 15 pending resolution of talk action plan A fl luf fle tent technic al batts has not been established for resolution of A-!! (see Appendi x G).
4 J3 y
- =
I
)
i e
i Table J-l i
i Comparison of EQ Review Criteria l
e t t m ' 4.as 4
.itt m ".,e h. %. m o,,,i
- x. a.a c.,33 muerc esse cat e a vetc osas cat ;
e s c. ont
......t.w~.~..
o. it
~~
.m
+av s
- e a
.~. e,.
ia.. a e r...~.
- n...w...e.. - e.o.~
j
...o.....a..
p.~... c mi.p. a p...m
.~ -.a....
3
......a..,,..
.,,..,,,.........a
.~i j
u1
.~.....i.
.m.,.,..
1
,ac u.
6... a o. w
)
i
~~ ww.
=
2=.
1
... Y x a a.i.
i
.ic.s a o
m.=
e...,.w.c i o 4
a m.m...
.ac j
i t
s...-.,-~..a s,
, n e ov. ave ',
- Pt c..c. 4
....., e m
.a
,<..n.. v. r
.% s
- j. ' er s..v...d 6...e v. e s.. e. 6.so
- Spec.h< sp,.o.,,s e p e o o.a...
s om, e...i
....p.
- =o..o
{
........ at..
-.,pe 6
5 sc.ag me t ww..co.e w
..ma..
a..
A j
e.,.m i
4 a w.. e. 3 40 *
- 8w S w* e.v
...a swa eas
.c.ne.a J
i;
.. we i e. p.ai.
l J
t 3
2.o e. rwa.
t e a e.~.... i.
.i
-..,,.,...,..a j
w... -.
~.
.,w
-,. ~.
.th.e
.w,. ween. met e m.s
...w t.c.. ws.e s eae. wwas 1
e 1
..,v.e.e....
..e w
i s....tp.
4 e
p.. -f f I J ) 8
- i...o..... a s x A.
s,m.s op.,.Dao 8**. ww.
Ns..m s et.ag e.
mv.
s.m.
C e o
4
,1 es s.d.as e. I3.emem
- e. s.o ene s...
ah. p d.as.
6 3.e'
. m
.......ym.
.. ~
o..
. a.
..........e...........a 1
j n.~. ump
.a ee.
4.m s....ns.wo.ag m pa..a
...o.c..o...,,.....,
j e,,n.. p.e.a s.ai r a
..ov.u,.
v, a. ova (,,.o.
..a.re a.
l j
... c a...~.
J
- i.,....e....
,,,-.....i.m...,e,
....,..~...,o....
' w s.
.w,.
o p,Na.
d... J' p.ea j
Jt*.AOsa u-p,=.......a j
Pa. e,......,
... m.,
... c he.....
e..w. i.ea j
,...o,w.o.a.. p. o.o '..
P a.. J.,......
.v.
a i ecc.
j i
I ii J-4 4
h 1
4
i, Table J-l 1
l Comparison of EQ Review Criteria (Cont )
i af tf 3 2 3 71 ans
'i f f 3 2 3 7 4 aw Ne lG Otit Cat-a qu lc osse Cet #!
a 4
asu c s t.o e fle, w it 0;;n ',; a, a.,
i a onei.en et DC a s i.e as t Spec.hs pdence 'e*
See peewows pees See peewows pe9e e
e e-*e De*e seee eac+ weed dseeethea
'%ese% pommede as
'wa hee teae seesteoa eeg.need c eassene se be w%,esed
. u e<ene, ww.
e,A.i.ns,. woe owerwe gens e end t% e,,eee, dose -% i c ea..a X a-
,..,.me, oopec en.e the ope.as sene.sene te me
.aeewee He* eser wee
'wr*=dn e,
oppeed ewca se he.
ewet she,me e euewbeae
.ie,< e
)
j
.. e.,6
.. e
.. ee,, e ve d we.
.. e
...., ee
..-e-j
.ad 9,
- ,,.. e,. e.. e s-
.e...-
.., he.r e-. e,.. ea
-w ~ twcw.w. genc e
...G.
es eve oN p,.d e ic e p. e.= sed %,
i secto end e spes i ed ec e. 6..
6he-es s sp< e. 4Hecie end.0c ia4 e seases se
,r ease..oa 5.rw s,ee.f
,.a i
n e i,. e coed ie.
he,*w e me.e.
ene.*e l
J I
8 us a s r a
.c : c os.oi..o s I
- o. 6 ; e (eh
- a ha.or % '
j
)
e ea,e. es.. e e,w I
ome,a swese.ee 'e. t Oc a wwhe.,oweeksee.,a hee i De a Spec,N p, dear e p.e==ded so-eseowe6heds,ust3. peace
,e, Gied ea cam ais ps,e c ess w.es.ae ea...ee,wate me, e e e.. us. L e
..e> aw er*ws.s spe es erstems metoes shound be seeswasted emos ere e poem epec,6c meses s
s
'me' es e ae' meanies t to magne
., mag a pean speef=c meees oppsewed De the eted mwes be
'm.w... s seiesme#, e# tese me e se eie6 espe eved moeur*ip weed appe eec e pereag serasee eee e' heae cenaease sa the 8 V#t G e
8
- eet a s uea M
- A I II iesempeelese s==eeese are ee'e**as WWA and.se s endeases pseate ed se em orte<*=etve se psene s ea set se.eoe she permeaa pee 8ee Otanee. e psere seechc eneve.e opec.N medesel The sene pe e.=ded
- 0 @ep ed.8 VElC 0%$$ (St 4
. he* G h.,ep ht elp
- 30 p'tweve ef e* eased ie8 8ae 'hed eseg y I y &er'eseied.Se rest e4d'ee ed d eeil3 4 QC A < eas>te.ae 98 e s easide*ed dwe' tag ec hes e,em o
IWA end ate gemeenees pe erme e hg e t,en.
W$( $ an,eef es etsge
- em aset le wee the p *= eat p8e8e4 beoes we pese twfi es G ' op it. #. 9 4
p'ownded oft feue BW'JA f (e
'w*e e' 'ho t sN rtesef
.' 'ho 8
e 3,eb 5teteee servsmee pf w e.4 enw, 6 OC a s ens > pe =e.ee e eu eeo d
.'.,ae.de ed ow.ag.r.i.e.aw ee e
ewee N e t..,<
'es s a g.,
8 s et.se W5L$ sp,eehtalsoe e u
tspeed e# pest ( s*'rTo aeaf sw8
- evass ed w*
' e. e i e - weetwee
'8 'he weh s4 ' an.' t e*ig,e8 e',p 8
' e 8 L eeeed
- #De ce #.ee PP' v e.*
gi' v. es le.. e 8 tis
'he s aureg>ertent e evapeed edtle tie **e nee bN et.ee e 8ewqe e
- Le oma)..y ed 1 e, ese
' e We ' edl se a s tat. e i.a e *=
- ee
' etheae f st etv'
- teeDemig se emp teepn fue '. G e
- p. e,$e e
' e tedst ed of ewealted phy 84e 98eleC bea 8e*
Ihe s e.agptetent ee f 01pmeed e,,,3C A esseet e c ea Om.e. t o L OC A S re e,, s e g u a a a.
..,r uh3 g ame se s o
s J-5 f
Table J-l Comparison of EQ Review Criteris (Cont.)
l
.e t t 3 2 3 'd w=e j
,t fi 1 3 ' ' eae.
h*1kE*M 1 o w u L osaa Cai s
e' f l =3,1 yf.,d d'92!LOBt L ee % t.e a 3
s ee re **,e one.
e 2ee see,.. e. sees j
, m ee.e.e-. mas t 6 oc ec,e pw. w eso s e a:.g.e.,ausse,e f
5 - e. oc
- j se,-w e.ioca e,a e.
., e.
s - so 6 oc a w e.
% si a e g t ; peuCw' Oees Ow ' tiO4 C One f 4,eeng se-
'eaweew e ered peese=,ee saews S ome se C es e e
e
.af. 8 C.sene.ee y ea se the etere e
' 2 ' 2 news. owe m ee oneene be desea ease vene ehe emme
.. e,se eee she eseest e e' *e( L S ee<*wooi,eethei see's weem e *e oa e oseasspoe.he me I o.e 6 0C a. ewee eaowe e = ease showns awas eeeeese se.ves.gese me he esseeemos
)
==,ee me avses.es i
t.
P o a..oeae
' 00 % *eseove sweemmev meawmas C owe e8 the I C C S +=s ee ms
$sme es C et e s
m... wee.e,
.., se.hnee e,eces e,n,,e yes e..
e ;; n.- ;-r one fm we e
e on e.. s. net oe e wa.amde sea e,eshed se *=ethetead eew 'em e e
...w.was swaag song im ee s ea shee peneseeise seaseamea-we sew ew e.r.,
1 pe
. pea e
- w*w eew e mas p. eew, e w
e
-a,
no e e.w i.ne es w w onewee e.
.a ene.e a ie **
ama.
j i
e,eese s 5 a a ea.<e.e e.e
. e.
- >,.. i ea
...,v,. ease eewwm ee eso.seems i
eom...,ea oe, e e e,# i,eoe ~p
- s. e.s% c es ea e ese e,eene m ee..
a eees es c.eec ad e eaos.e e cies gue ege.w e e, e t hem.cas seense are see ee n e ees wa,
- c
%e e,ee.as enemmewomea eseeems t o,pmene.howed me ant.as f w sem. e C ee se..ceee c es e
eace se, wases seee penas see emt p ee ehe e,s,nea e s
e e.- c.as.u as '
- p.. sed met egeg ees,eenwaes ehe eenye e e
. eet.shes mas the seems are i.eae posewieted to oc we ewei
.n s se..
, a e,e s.,*.ag e... ee t o
- teate.aes se esem seassomas De onwneeeewate snowse be seeee.ee seed the eposetees to adve.ee
'adwneaal me werenoveng me owe os c ement.eae teet>ee eveterrw esewee tv the t empmeses aos seeved D, Clase et oe'. mee emee gens, mes tace pe =ee eyet.sm Otherones e sweeen evetems. en eerwee er, peae speek enesyme.e soonees Claes et suppen eyeser vi thee awy be seewees. eN,.ee t e e,em s.ee se the Irmaag ceashtneres ehet sewee seews seewreeag a 6ees of the o.epen erstems C oe se e
weate see peea the ope.*a e
.netophag meressaag seec es to esort the speestere se adeseos
. pas.o.was se thei s w..ec e.,e e
..se.
. e.,... se.,
I 1
J-6
i l
Table J-1 Conrparison of EQ Review Criteria
( C ori t )
.trt 323 ?..no
.t f f 3 2 3 7 e.as
.or/ 6 sts'. Le~nm
. s? wester u Wan_21181ALJ
%f'.DeGitaJ.ALJ e :;...e.c..o=
w...
..e.
,.a
.n, v
.a.
.e..e s.-.
C e.
e -ecs
.i..c.
~..
e c.. a w 6. =.
i.e
~ *,,. i a
in.
i.e.ae..,.
i.s a e
. e e.
........e.n..w.e.e w ee.. c..a
., i a.
.ase.,se... ece
... c i..i..,
,.m c.e ei..w
,im..anv
.... ~.
a..~,.
.w.,n.a..as..acw mw
- m. w.m..
ae
.c.
a.
c.,.
.a. c m
~
...a.
- a. c.ww I s>,,=.* thei %.'
ewas m.a '.
.e.aew..
o..h
..,c.
e....,
,c.
i.,..w
...a.
~i.cc,.....e w.
i,
...a.
..a.c.
,.aa. m
..ee,..c.eime l
eua.. 6 ~..
a....
.c..,
e c.m.m
-. m.a...
- n. De a
.*.v.s 6.
...e m
..e 1
1 1
s.,,.......e,e.,..~
i
.o.....~.c...
l aea..*..-.ae~..-
l l
........,.c...ei.,,,....
t m.m. m.,
e a a.v. a.
\\
. ~.....
(
l
.,s.....
- .,...e...., ~.
.a
' n..
.a e m.
.e e s e c i u p.m.
w.
...,,.r~.
.. is.
cw.c.e
,v...
m.
.m.,.
..w..eem moei. ws ow.e.a.. me. es..
..ae....,
eew e,c a.n.se m.m.,c.m.e,.no e
..e an e e e.e.
.aa.e ae.
.e
.a
.ae m.
cw. e e.. e.ac.e..
me n
r.. c
.c
..no
,c..e
.c..a v. u c.e.~
..a...a.m
..wic.eee m.... e.%.....< i. e
,m.
i
.c o...a
.m..a e...m am.ai e e..<.. cio....a.
....c........~..m.
- i.. 6oca as usts.
., evea, p.<*
.m.a.. m.. ipe l
me...e.e me.
...e...
a
- e.we
.as.%.s
..c..wm.e o...
csc.e c
.m e,s.6.. a.i hse.,.
..e wmm ano t>a.,mm
. c..a ec t c n ca....
t%. i.c p.a.%
e.....,
.e... ec.
a.
' h. iw ee cu,. i..n..n.
e.cc,.n..nois.i.e...wn.
a m..
e..e cw.se n. a.
s
.ae..
.e.. t,..ac a a.e e,.a
..e a. m.,
-i n.
.a.....e
...a
.an,
...e i.
.. e.aae
'w.........s
.....,e
..,,......e
.a...
e.. oc..
.,i,m..~..
......... 4. a,.
1 I
.i.,,...........,
..o i...n.
- ....>...a.
p..~.
.... s.,., c
..c m.a c se o. ihe
%..oo,....e... m.a e
% c moe,....e..,. w....
..#,...~.
..m.
.o..no
.m.c.,.. m.
..w..
maw m...eoc.
a, et i. c
.........as j
. w.pm.m 6. ae e.,*'..o
-~.*'a***.***a'.c.*('..
w' c. a a.'
...*.a l
J7
Table J-l Comparison M (Q Review Criteria (cont.)
i if f t 3 23 71 and 6f f t 3 23 T d erw WMG olas cet 1 9 sefaanon flamesit 00* G easarias wetrc.cles cet a l
eene e,s ea, epos.he enetweee se peew=de en awetaduse bee bet weea t
f ees I e wowa s ee peecewe pe,e
,he op.aee.ees.ene w w i.ee me demoneeeeied
[
ase.e-ease ee se
( etae i
i he s e eseesonneaseh was ee.
e on,es.,eed.a the es seng,e see, dois
,se Meeweenag te... ade.eeeed de 'seenwerea fee detec eny j
vesseeomeate see the eseistwee to Reee emeaes meen.av e s *,enges end mwee t.e e
e e,e mese.,eed ieamwd ay es cw e s... end ihe,was e e.s,e end
...*. ei os ev.ae t
.s.i asme <.
esos es.oa of eees merwee a*,g eens seehea e t erw eede The esme
.nseawat teet ween meseweemeaeo se*e fee secm wer.aD6e ohes prownde the esme cepend T he tempen etwee so w%e b em sp eaes et *ecei eeamesse meae.e enneemed ehewed be de *"med 6, eheemac ewswo eeee
'he seet :=ea ehowed emaiee pee f ormeew e and spacesaraoneas een
.*ge perteamenee s hee eetenehee showed be y,ched bef oes sit es obsee to be moeow.pd. eest ogws, meae e,we emeate sneswem0 e
end penedissey ownag seeteng oc sw ee.ee end ehe meeew eaisas ehe owphewt the eeoesed eenge ef e
speestreeev and the speeadnames seesens e ec etep by etep setans steeve showed be moceseems a en terwoweiy dwang leeosig 6emeeese the topiposenwee to w%eh esse e tervene me, be psebhed fee meat se enshhad showed be deba n
song seem tesong) ed Dv theenweeweee eeeenge.
portsamense ehereseeneesee showed be weahed betere enee erW peaegheaf, swangteenng sow owghows the veggwseed songe et j
ope atashly and the speseensbsy o'etwo **wwnd be pose wtoeed tea r
Sw.ey icebny e.$.e..oes e
e
'e *w wu s,
.at oevee me, be e,e b8.ed 8.ne
.a ng s e e.
s e e hng.
i Dee ie - e he oe n.e....e.ed.o.,e.e see
. e e - -. e se ee. e d e.e,...e se h em,.emeas e ese t ows ene end pose seted
. auess, se p*ec es.stwo ehe oestw coweere es pe e. t.. es.ie the pne e w w
ou.sene ehenwed be weed
.eied oaes eevnent
'he eeee se eied on. nmece
'te e.,,pgees sweew e essee eed ehouse %
e e, so toes ees,eas e ews een
.....e w
j e wapes been nearme uses ehnei e t.*
t e
S esIneesee owe ce eerweee ee*e
.. ed beoenne date og,e. eeien t o the De*,e 8 en be agobed et saw ihme
-e e..< epe e, e*,ow.d be.n est,s ee
.ee.e spe<e......ee,e
~.,ee
.eet.e.,e,w e pr.
ed. e s, o
s o,.t.e*,e e e..d
,~,g e.~.ei ed e.ee,.
e,ie t....e, e,e es.o..me.. eapose se mermes mes **eew, as esas ee.e
- eeense thes se e a mas se e. not se ee hng es the mes.m sta peseowee De swec optstane to emprehent end eer'wev etwee 6eaubbene thee att vibesheae ubea eio ownng j
aww d esewe when the speev ere Seonga bee.e e voat i.pe.e* s poe1 e
s ee.ee.ee sw*,oge se the eerwis e 4 enei.ea es.ese Othee..ee
'eeae e. sweis oc e. seat e.s e n e ses,ie. aspect sad es eed e. =s.e.ge
.se evea espeewee awei De eso pe.ed p**oe le os senswe s ent ** *t h S ep** *te ottocte toehng inet en empeow e to the steveted teve sehen to Cet 4 pf eaiol se not C oustet speev ehe.e*d toe ns orpee l
e emg semw st ed event test s v o end presewte eteern ese se ee.c opestre toe entaf mestnsal eted dw a
ee
- +pment owth as penelesteene
- ng et ehe me e.**wea ses e e s.,e e eruf wav e*weient ter9poe stw e c eree. no *het er r et te*e sets 6ee. emwee and e
mot oe s mest stoneastte*e secolod m ound est cwe a hea ehe oas. e e ae.c o c oat eere*ient me espeeed to speo, eyeieras a. n.e's J8 i
l
\\
.i l
Tabis J-1 Comparison of EQ Review Criteria 5
(cont )
1
.i t t a l l ' ' e**
'i t t b i b
- e **wn e r c ce g g C at_ j
- t <e.1G WU4' f
L o.'sntve beawen RO e "
.dai.*we
'e v 2
- e. e...w.
e
,...,.,or.....
.e.... e..
. e.ew. e e h wa.s e.
- e t. sew a e t e....
. e.g.
e
.he...
t w-w pe e e' e e.
.e.
e..e.......
i e e w..
e....
e... Da..
.w gh..we ehe.... e.
l we..... epe..s.ses,.
.ee
- ... e i. p.easei e
t w.w e C es la powe end.neam eeo, em e hee ns iones... s 8.a. t
- ee s.nv me. i.mam
~eet ioeted ehowed ee wi. w-iease
. p.eeeae the st ews esse.s et a ow.amea' showd be seabed gw. dense e'esee 'hei 'he 'vp*
- e s s ehese a-o d e op** *seea se w
h e.. e ett e.
ead ewaae ees ag e
e.he s eae,w e e e.
ea ease ehe, es,e.eia no as, eae.
- *w owgh we t e. eave es... od t he e e n emee e' *** ope **
haae e spe=ao **e ***8 ease s p, 4.* ine p..,o showed be..
see. ase.e.t o oped neecwaea to ne<mer mee" p r ee ownag.
sees etc s
' he ope.ee.d.ev eteew e' ew.e erec e sad oe.wm ww.o
,=s
.aeasehewd be me,eie.ed seas.a
.,*.es.ea dweae the demea beene
..., w, dwang nees.au 8 e.
mo e veae e w ope.enea pees e-eees.no me,eisaag ei ee
- s. 6. seas
.eie.aee,ves mes e e e,es. hee 8vrehe. spdense eiseee thee e meene awee be e ewwsed see e.e.
f neec eed es teemee e co.
..,.s
.,e. eave cw...
.., er ene. p r...g the ew.pm
..e e
wens e ehwwed te ee.v. e.: aw ~.,
move 9 e.,.
si e.s u.mte e.d
. e e t.ng mor. a.
e,w. a>va.e evang e.,e
< ee e..g
.... e. e e. e e e ~.s
.2,.........e e
,... a
.w,.....,,e e,,o.,,.
- e e.. e..
I w.umea. me.w
..a, e e... e
.~.........ae
- e.. me. +,... ee. shou.o n,e e.ue.me. a.w., ri...g e.... o...
6 e.e. e ce e. e...... c e,w
... no et...e.a o... ee,%e....aneo e..., men.. end the, arve c ea
.., e,e e ed e.... - g i.e..ng.
,.d
,*.v.....e
.... e erw eihe..e w..ed i... c. e e ene..... a s. he e...
e., nr.,s.e m,
,.e. ee ms,.,,es... a s
e es.p,ionen, ee a e i~..aepee... o h.w.d be.a
-,,,w
.. ea... p.. d ed
',J ed e.%e eg.a.gs**. sat mes> 8.t e v *..g* va oo.... o. e.*ip hoo.e e **eswed be p es.ed am s o,'vmon p.sheems tws m et de se he6es sist peepe e 4
we seed and neeeed seet e,s. y a es.nes 4.one A p ag
' este rhet mere oweeseet.d wane.g Jewees setN.eed D, ensanter.* eter*
A gp, g eff et t e a,a se oow.pment 3
.eneweing se.e. pet.s oftesse seaw e,% ease a; u ; thee mees foot peoeged derde that me,e se**weessed to e
mee tre some.de*ed eseeptedpe
.metense that see owt.ec t te showed be r uae.d e. e4 mes e w..
g. o v.ded.he g e,vpommat dose ac, ope.g eHot te thens a tre.ejent.f ed ed ' o. f. o.
- s. e. e * ' he es.
eat e e *%et o a sse ehet ere e me me e,d e othedwee chuw d be esteD p6ee ehenweJ. ve app t he e,
- g a
m d e t...e 1
be Owet optsts.e le e.grefit ent ached te Pehead.6 ee, 'esa.et e 8he e.,N.et.se 6 06 Dow 6 e
e f e. 8 ed s e.,s...e
. e e,inJ e..
en..e t see ans.so.3.
..h..
e,....
Gege mslet..* ewe.
8he**hW end e
e ed at.we aw g O he. ar.ee e
- e.e. es 4 e g. a.i.ag N e a. e e
- e.
- ee a
t o..is.e. s ew e e,.e g,...... e.e.
a,s.8 oc =le e,
. no g.,,*,u..e a.
.% gt be es t et, s hed we e snee og sg...g.e. e. e i e. u. a.o. %. = i e g
I e.e.. - s.*.e. > e
- en e.e
- e. e sa a.e,eee t.ee.e em weie e e s.
w'h
. ~...
I 1
i I
l Table J-1 l
Comparison of EQ Review criteria i
(Cont.)
i 1
etti 3 23 f t seus 4 t t 3 3 3 f e erwi n-Non..a bar iece DOa Gwsknee kuMC Otaa tm. s fntfil1G OMA Ca.1 the A, ewe meen eeney, se se -
i A,as m%w.we seehreo.,ee we pwwe see pee
,e per ese ee rene seeeeeeeed n
asetemme aew aee me, c.a eeres.ee eissme te.
epag Me me.e8 88te e..se.tened' e m.p.#
.W g
.se. oh whd be 99 nbed eHf $se
..e e.
...,. ee e., C o.
e u..eee-8 e.. pense.e.g m.t.aese e *ege a
- a. a.,
bie ire.d se a b.me
. ~
..r.g......e
%. age..qp.*g e.r.t
.eee m.c 'vu n.a.,ee ee=se a.. De w..e 4 em.e ma
=
.ne p#m e ft.ag
.ao
..e s..,w. e*=
M be sehn.e
.. e
.. ewee
>.a,.e e. a.e
.s e w,..s e.c ese,eyeee a e te j
emee.
,,ew *e. Ces e p. eases i
f i
v e
.O...e,s
.e e
j e
.e e.a I
e,ae.: e einem sonne
.newl# Dea e s,
ce-...,-.,e o
v e
.~ c si....e.,.,,,e f
. e cha. s at.al e the ee*. nee 6.*iebene thee are 6.e.pn s.m be evenweted em a teen *tg ami,et.as* wee pa ce.em.8te te
..pe, o.a. e e.e
.eee be.e
.eae,.
e.
e.e
,ne e.... s e.
,,e.
. ee.e j
,e
.O..
l s one.t eae.p.ofeed and the f ewsom e thee se esocoed se
..e son qpeeeNee pe eeawe e e ne.e e<.. : e => e..
.*ec a r wie
.c ome.a esaet as one emeed ea.en ee h
,.ac, ame
...ee e me, we
.e.s e.e.a e
- i. p.ev.aa.se swash.a..9.a e
...a. eat s e,en i..a.
ens.*e e I
ees.
b e 4..ai
' a.... e
.e
.., eh.a pea.4 ee w e e. 4
..e.
w..*..
e i..#.
f mee,.,
e
.e ae
... a e.e. nee e. % i.e.
,s., e...,........
.as e
..ne
,ae o e oo
. neo me eoense is o ie 's, m.m.v..o we %..t,e.,. ~ aea e ea.,-
w, se.ae m.
.e.e.....+.,,,,
l
...... e
.e e
e.
an.....r e j
. ei i one C ei e p.a...
i i
I r
- .e.
h. wie....e
%,...a.s..ee.e
.,p a.., ene i... e..e... e e.av i
9..
th. eew.larve.**. e8W b e.$..a =.a
- .68* 8 8.**
.e. i. - i se..ee....
4 l
... -,.a. e e
e,a e.
,.. e.
e~...e,
l
.no c e s %e s ensa... ei s e i
easm.newisti.a orieshw s a l
poete beerage aus.i< ease onesenoes e...i
......y
- e. e.
- ave tese te air. egeo erup ste.'
]
r
.etened so ow-ea.e e* =i se l
l I
l
' he 'miw e s eisee.e showed be g er
- e. C e.
i....a s...e e
w w e e,. tee.e enu,id mc e de s
a l
s... e c e.e....
i l
I J-10
Table J-l Comparison of EQ Review Criteria (Cont.)
it ti 3 2 3 7 9 and
'It t 3 2 3 t o end Les N et oa tiemeat D 0* G oede6.nes NumEGosascet s'
Nwef c Magggi,,)
esehenol tpedeae+ thei *e the 8
w e ; e.iee.e
.neiew ae ocew ec,e ee e e e st* hed pt.ee s e.. u ng e
e evenwouva e' the **** ewwete e
, e, e
me,.s e et emeen.no ehewed se x,e e %.e u s wesed pwa ehe
'e* et one t me dwang the tesi f % i.e seew.te show d somwn e
emeae sees not pe<*ve she v d se#e secembea e.... that t% eowemeas s on e ea.a ihe e
pee've m.ee eeenied Nas ova 've S ee. 'f 'wac be"w ' e*e"*d b e '**
o e i.e s oh w.d be 4.no.dee ed r
se seem e s e ws.beae poei.aesed egaraeae spece'.c enoae e
es ynows.e i =.in men pae ev.ng.s e inenessed e
hie f
"sws wmeatei pa ee '
4
- eve A en,modeeeeedime.
A ewmmee, of the test,,ew to
$mne se C es e4 e
e e, a,,,,,t e see v,eeer deswmeat et.on that com neteetoe the sees,et y
.e.,., - ee.eeo..
we the.,meeecouea pee,er.e ees noed The tw s test 6ete p%#et seates the teet esowito fee oesh test
- ncewdsag the obsee bwe aw mens tested. seet f aset, and sectevmeetateen.eth teocemessity eec eeds teet peesedweee teet ew wnery seis and essw ec, e
e c onc6w asone eos peevnenesteene approwel e.gnetw e and date s
r
'.mee e s.ng i es,... ens -
c,en ec ooewe
- a. ee soa.eeeed
? es e meiswed.o mesi e.,t eu.e...
Some e. c ee e.. a ado.t.w,.m esopmeat. h.ee e e uag.e p.,
gence ihee speeenng e spee
.ewded Du the pa est e e.de se the
.eni: e.e e me t huc c f bes e n ed, e e v
eow.pmens ben 0 ewesit.ed W hers e, e eene means se 3,en.f.g euge spee shng e speesenc e.e s e edeled bwt d' geset wee f ee suppsomens
.no yg,e,,et ng
.e.e mee t w ee tw, e
p ettial type f eese sa.a t e s um ponente of the saapment ehesad
'ee m mifweboa el oms.pmeae be c wmpieted se owppve' e' t%e ehet.e.#,c eted ow's.de -J' the me' mod e eas eameae Uwe=hcabea 8eef ed ^ eeeed Qwesificateen erse be beoed sa Qwes.hceben shee s easeet et tes esseung operet.e.g esew'. S ono, Gei se-e ne hen opeeebag empenence dete sea o.shng of the somement spees toene ereth 6eenga sorece comen
'T abene let betti the eW bene end peewag that the Ctese 8e be meshhad and the egepment
( perlermance ehoeoceeeobce o0
'ee. Nth speetebag espe <*ence se the eessemene..e mese es se e,meshoe, adenb4eben of f ee.
coed the eSe pment spec. fit aboa
'w ee be=ng medieeeed by itwo wridee deega seen e g eghe.pr.,
s me t hod coeveersten et the Periam type teehag me, be cow pment enf eemeben o.ammary necesseev te adde oee t o ma.t.une end sewece of spe eteng een w the, e,* ewt s o,s.ed o,.,e
.eng ease W ihe be wa. %ce.*be e epe ns ew e e,,. a e, e e. e,...
see.
- e. b e e a J o s
...e4 v u ee
- ap oo.,
e 4 g o., e e ' ab.e
,p-ea t,,etti n;3 'a I
J 11
i Table J-l l
i, l
Comparison of EQ Review Criteria (cont )
j i
I i
et it 3 2 3 7
- wie at t ! 3 2 3 14 ead Nu t G 0588 C at ;
n Nu t G OW Cat t!
o
' halc et ea Demer.t DOS G ed emos i a ne ve.e i
G oa.* w Gwd ens e Ae etwa of Qae ( egapm.as the mothed.e meet e,etems. fe, S ome se C es e onwe oesbeead
.et-o. ~. -
(
- e..e
~.
e..a.
p,.
.,,,,e p.,ve.e.,.,e e, ~
~
~~
eees,.es 1
e.ep e.hu wh
..i.e e.e
,en. - -
e.e,.. e speci,o,.,e - e
. ~. t...e e.e., i,,e e...
gi e,e.,,,e.....
- e.,. -
one e.
p wt.,,,t,.
ee..e.e,
~, 6. en. -
j etwee peeeewee and steem T he teste on =*ter ee-eneate of the the es1, ape 6 stoa of teet date end e
p e~ e.,ew
.e.
e.
.e.
, e, ~ en.e.e ee j,
.,o
.w -.
.e~,
_3,
,,e.,e e
,s en, e, e-
.a owp.e,..
,,,e e.,,s.
n.,
,,ehea e
mem thes peev, ewer, soeved i
.c w op eve me, be demease oved
)I f
ll i 0,.ae vme l
js
]
Ow.> %.s a t s e, enee.* e nd et a one s hwa Oweniheet,en e be beoed ea Q.o.hestmea ehes eenmet of e
{r a
. _:urer e, aspeef p, ef shee 1
De **** ase a c e ee e, m seh.et. ore we pee sete ar%cm sentene the news, I
I
==eso two ao spec hc ow.cesa mem ope.heesww..aaciese en the Cha 88 e,emme,w9w
,e... -.,s.he.
o.
ee e seie e.,
..,, e e s l
g
.e...e..e ese,,,
~ - - --
l
.....en.e..ee e
.,,.,e~
.~,...-e 9
=
t met. Set me.es. weed.+th appre eaweet emente
.a pensees t*,e 3
g DNele e,eliheeOca e seetnetrea p*eef 88,,,e9 De beoed a e.id l
ei er elWDsel mothede end res behod pe,acs58** epov eDag e sp**
i'
. w,,wiw p,ey eme we.
.e m e w e e-n. e v,......sc e j
[
l ewmm, se enevs,e as, e, a s.,eeene
.a.s m.ee As j
..i.e,on.e p.,*., men..
e.
c.i.eae..
- 6.,,..h i
.?
i we,eci aes.s e j
Speahc p sene.
in eepere is meshemensa m 4es.ae and l
esteepeeeeea of deve pvea en etti 323 le d
1
\\
' ova 688 if D. i8 f f he.noeemeaim ime,owomem.a Seww to D084 Gwi.e. nee issoe Ch,ethee wee. ene ee the se erv e, e, tete e,ev eeg,,a**e 'h*
- ****das *** **,mase end sw a eas p,*ei,me,we e,ewhcot,e,.
s ow vedene.,eoseemeate ther mwet m.
sw nee se e noec.hc oweef,.s ht. b. e.me a e
eineiec tw as D oes it ee,o m.~
b es o. i e,ee.n ow.s, ment o.ai. hc e s.ea
.e aet ew t.o.ent is e,ei,#, me e
Queehed ete showed be estes>
e sp.aee f e, esmate weeed, c ea sense beoed e,i the oevoar, es es,wreed man oscocae e,w i esag N e.
peeve.med the eenee,vesser.w wees.a the se H the senipen.as sontane mete' teseeseeea es see,ew,epws,ong thet see taie =a to 6. ewec.p c.,y ens v..m memwee L.cee to me,,heent degresecea
%i me, b...
.asse, e.ewmise owe i. %,mm w,s endnewea ca e e..ew poed on, enne sp ag e ouest.es see tw the cw.wea.-.e,,e 5. e.ie9. h.4 e,o, m
er e snee Drseee been s of 4 00ic Ja pw.~9 pe eg..m. e ouse...e' t w.emean 'hei e c e, wee.o v' a pe..v n
,.-ee.
..%.e n
me' ease she' *** ewes ecutae is e 8w ~.ne....e.','
.h.
M a i*dit%AsCf v me pasai to as e a ow.
a%c on
.m. eno m. a.
a ace..ce.ise i.
.p ng enecse c w.e p..sene e.,,meae
.. i,.....,
,,s,s.a l
e I
$, a t i. g 4,ec t se.... - *.e e m,.pm. ne. c..
. so see e ec heowse to. peaeo ows,ee,i es eg 5,meae
.,,,e,....
a sese 'opeecme t,= emopawm nea a f *h.fia t a.at N ' %
i J
J-12 4.
Table J-1
~
Comparison of EQ Review Criteria (cont )
il t f 3 2 3 71 ensi ot ti 3 2 3 14 ene L ea N at-a tweat 008 Ledc+aes Waf G 0$00 Cert Ji WafG osse cet I r a.m.4 ew.e.seasei.emag e
c 8
- 1 A *00*C e wi.ag age easied segrade end,e, meieasse eewwed be esten i
.ece. aonnes ee,w.ce e ms.t a.
j Maih'thaNCE i..a..s p..o.as.h.e one..
eenes
., a., e s,.e.,od sa ec cepiease an.in.4 e
.a gr.ag
.ehe I
ANC Ca p,e,,ed se n.see,,,
e 1waet% aNCt i..a
. ees % p a d.c seetag et %itut%eg eoc e ve the ow.om.a so she was i., w e ag ee, c. ea=w eameae
. 9 *%$ s.F*%e f
l
~l t e m o * < es ea te eae.o e S.'... e i e VOO.e.( a
- UNS
%..od.....e e.. n.,,, e,.,,.p..a s oo.
- 6.. i.
m.oo swi,..w a, si, ta e e w e.:
- e.o. i.e e.. m.,p upe es.ag esponeace espee..au oeaee enee m. e.eewei.e i.
r a:
a.ciea %.,
s.ee a.
ie e e
l :
-n. a + o%aehseeea got aq,, heessa e wne.
o e%
aeen.sease,eae e
O Det uf n t a
- o%
w c ene ewo,ien
. wee e
.ac ws.ag - e asea me.arwmeo.aeee,espv e
a.. p. e.
.me..a Mt.c
,a i een s..
io e.meaeie e e ewe + %
i.hcetea s ee sea a es.ea es e if f f 3 2 3 ie awee 6. awase.aea
.',..d e.a.e non a....s h e ket e uncew.ag aspeis emea..
e.meaese smag thee e,momem.e w., em.m e s inee e.iesse uneswa o.,eeh.e oms an te N p.no,
.ag sees seio teet seew is one meac e ees,*emeans Owess.
e
. e e oas ee i.
,y - - _ w.,.a ses e oeswm.menea ennuse e
t en.wmea eniume ye eceneemos
.aeewee ow,m at eeae,6c es a
- i. pwwe sa. n. a evaewe s eieg.a s enea ep.oheet a teet es eacee se e
e.ea setene a e.'. w i s e
on.
op.<inaq.aewawn.a eis 4
v 5 ;* f uA %f *, pus
- %. 0 0m G,d en.
.....,,eed.o g g g 3;; 1 i e,, ages, e
.I t t 3 2 3 7 4 e s aes.ag ee este
]
Juvt n.
- S ped o.a..a.a.i.e u e e eag es eting es me seisand.ar.m.
end aee, mat ea ehee..
.o.e se e
m ate ehe. e.d
....a n ees e s owppee e
,e>
e.
a e.3mm w e.,enoa
...e.ng pi ent e e
v eet.e, i%e w ee n.ag 6aveae
-..... m,
- h. ei.ee
.em n.
em.
. w r.g
,te..e,e....e,g e
j a...,a.....
... e i.,i n.....e e, i n.
., w eag,sen,cnea ei e *.c a p.em spec e.c e.ee s s
u oevegGpggg.,,. ;' ei g.,,, e d e l e ew a. DCteesec
't
% p.4.e.t N. pl b QSIll 'e' C es se s.eate
.e
.e *, p.,9e,awag the sieni
. o pe.. meat e en. p,.e.as e r ea
- ews, m.aie
.6 p,.o......,a ope.hs.eae.e.
w.....
o.a.t*l J23 ?i enis i
%.e.e et t f t.' I *e e e..
. %.e e see.6.
i%. soie e,w.m w
..... 4.s 6.mahseu e.evero.ng
.%.e
%. o.ie erst.aswmenea a e.e mee a a o.e.. woe om,p e., a,
ie seyniG ogsa e.ies
.pe,.,,.eas,mean e. c he m.as m.neh.enua e
j e.
i98' D Oa G d e.n.e e, e noe Poe Rev ' to e vat G cgga se,ee oppi.c ense to p. ease otea.ee se s Me, 2 3 '900 Pe.me, t se WRf G.0548 seeed swe,t951 C et i * *e,* em.m a u
s.e', ital Ces e eeewomeate see appbcease ne %.e p.
ais e.e app =< nese to th e poems twei. e.eceme.4 e... v.,23 e
%eeeee.ceasedeMe. We,23 1990 we. vae m.s e....
i,..=vi
' 9 80
..in c oastnechea pe mt Slme.eeved.a,s,'
974 e<
5tAs.se. e be e+e Adv ' 1974
- es ee s
J 13
i..
i APPENO!X R Generte Ceausnientient l
N.ey genera ce. uni.aueni hive been tiived by the ac it.ff te.ddrei.
i specific equipment quellficatten talves al they have been identified, for esemple, ou<tilfication talven that have been addrelled by the 6tpff in thin regard have included probleth With elepr'en) sonstratten assembitos, terminal i
es and r s' iswltchos. Othergener's blocks, electrisal sable sp) d tho staff grev'dhng is Ofte guidafies at te corro6pondente has been illue how 10 requirements theuld be la tsf'ed, effsettvely est b'ishing the preliminar programmatic reguLrementl for (Q before ' Pet fLe requirements were codified b 10 CfR 50.49.
Thts appendix sumarties much of this generte communical on, focusing primarily en (Q programatie requLrements. Potentia) lisues are highlighted in bold print.
IE Bulletin 77-05. *flectrical connector Assemblien." November G. 1977 The Bulletin discussed electrical connector failures that were experienced during EQ testing that was conducted by Sandia, l.icensees were requested to deterutne whether sistlar types of electrical connectors were installed in safety systems inside containment that could be subject to LOCA conditions, and to review the qualtf tcatter, status of any connectors that were identified in this regard.
Licensees were instructed to submit a copy of any cualification documentation as applicable for NRC review, and also to submit plans for resolving any electrical connector qualification deficiencies that were identified.
IE Bulletin 77-05A. " Electrical Connector Annemblies.* Ncvember 15. 1977
)
This Bulletin was issued as a supplement to the one that was issued on November 8. and expanded the scope of the eariter Bulletin to:
(a) includo j
electrical connectors installed in safety systems outside containment, and (b) include other events (i.e., not limited to just LOCA conditions).
H Bulletin 77-06. "Potentin' Problems with Containment Electrical Penetration Assemblies." November 22. 19??
The Bulletin discussed instances where discontinuity problems were identified with containment electrical penetration assemblies and asked that licensees respond to specific questions concerning this issue. Oral responses were due on iovember 25. 1977 and written responses were due within ten days.
IE C1rcular 78-OlL 'ETvironamnial Qualif' eat' on of Safetv-Reiated Electrical fouioment at Nue' ear bower P' ants.* Mav M.. 47A
!he staff provided information to the industry regarding the Consetss1on's 1
Memorandum and Order of April 13. 1978, and provided the results of staff follow-up actions.
The following information was contalr.ed in the circular:
As a result of its environmental qualification review of SEP plants, the staff did not identify any generic qualtftcation deficiencies.
- However, other NRC review effort identified a number of qualification deficiencies, including poor installt.tton practices, inadequate i
l w -
j 1
i consideration of subcomponents and omission of certain environmental parameters in the design, qualification documentation was inadequate in many cases, and some licensees indicatad a lack of detailed knowledge of l
the quality of installed equipment.
The staff emphastred the Comission's view as expressed in its April 13 I
Memorandum and Order that licensees must be knowledgeable and vigilant and must take more initiative in ferreting out details of potential S
plant weaknesses.
1 Specific deficiencies and issues discussed in the circular included:
i a
)
In certain instances, electrical connectors lacked qualification j
a.
j data and were of inadequate design.
I b.
In certain instances, electrical penetration assemblies (EPAs) j lacked qualification documentation and in one instance, electrical j
connections in the EPA were not qualified.
In certain instances, terminal blocks were not protected and in c.
some cases terminal blocks were unqualified due to poor design or l
installation practices.
d.
Limit switches mounted on certain valves lacked qualification.
documentation.
Some electrical cable splices used in penetration assemblies were e.
j not qualified.
l f.
Other potential problems under staff review included radiation and temperature effects on electrical cables; adequacy of l
qualification testing of components by separate effects versus sequential testing; temperature limitations on nylon components j
used in solenoid valms: and qualification of electrical l
transmitters, i
- Environmental Qualification of Class IE Ecutonent.'
February 8. 1979 i
The purpose of this Bulletin was to elevate the status of IE Circular 78-08 to 1
j the level of a Bulletin, thus requiring a 1teensee res y ase. This action was l
taken based on the results of recent NRC inspections t1st indicated that re--
i i
review of equipment qualification status and resolution of problems by licensees were not receiving the level of attention that was wreantec The i
NRC also found that unqualtfled stem-mounted limit switches wer! being used on safety-related valves inalde containment and appropriate documentation for i
qualified equipment was lacking.
In addition to other actions, the Bulletin specifically required licensees to submit written evidence of the qualification of electrical equipment required to function under accident conditions. However, ltt 6 Jn pl W
- being reviewed und9r the Systematic
[ valuation Program ($lF) M e not required to respond to this Bulletin.
I i
K-2 f
- Environmental.0ualification of Class IE Eouloment."
January 10. 1980 This Bulletin was issued to request more detailed information regarding equipment qualification at operating nuclear power plaats, and to increase the scope of the inittai request that was made by IE Bulletin 79-01 (the eleven SEP plants were not required to respond to this Bulletin). This increased scope was necessary to resolve safety cencerns relating to design basis environments and to establish Qualification criteria that was not originally required to be included in the Final Safety Analyals Reports for operating reactors.
Specific concerns that were addressed by this Bulletin tecluded high energy line breaks inside and outside primary containment, aging. and submergence.
The Bulletin was very specific in defining the information required as well as the format ~ that was required for sumitting the information, component upgrade requirements, evaluation ~ and reporting requirements.
The licensees of operating reactors were asked to evaluate equipment qualification against the staff's guidelines and criteria,
' Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors," which was included as an enclosure to IE Bulletin 79-018.
Supplemental information to be used in conjunction with the Guidelines, the "Interin Staff Position on Environmental Qualification of Safety-Related Electrical Equipment," was also included as an enclosure.
Sucolement to IE Bulletin 79-018. ' Environmental Oualification of Class IE Ecutoment." February 29. 1932 The staff issued generic questionf and answers that resulted from the workshop meetings that were held in Regional Officer to discuss implementation of IE Bulletin (IEB)79-018.
The following guidance was provided:
The scope of IE Bulletin 79-018 is limited to electrical equipment that is required a mitigate e design basis event which may be exposed to a harsh environment.
Note:
The scope did not include non-safety-related electrical equipment that could impact safety-related electrical equipment.
The containment temperature and pressure conditions should be based on the FSAR analysis, but the figure in Appendix C of NUREG-0588 may be used as a conservative measure.
Spare parts are required to meet the same criteria as the installed l
plant equipment.
The requirements and positions stated in NUREG-0588 are the same as those stated in NUREG-0578 for environmental quellfication of electrical equipment and components.
When it is determined that upgraded equipment is required, equipment that meets the requirements of IEEE 323-i4 should be used if it's available.
K-3
~..
.?
l 00 e-When a deter-inatten
'he existl.:g data is inadequate or that sufficient (0 per IE8 77- : 3
'*:, the condition is reportable-Ad11ttonal r f fert ar.d t -
radiation :crvice cendit: --
b necessary to establish the i
vnt if the radiation service j
conditions r re not calcu:
e-
- 1y using the methodology and assumptions di: cussed in :
l'~
Hunnlement No. 2 '. ?J LhHt t b 7 9
d r?nmental cualf fication of Class E fouineent." S C C ;r 29. ::
This supplement to IE Bullet;n 7:-c :
information and clartftcation r:rtain'--
- ued to provide additional nvironmental qualification requirements as a follow-up to 19 -
- that were held in the Regional Offices during the ecek of July :",,
Supplement No. I to the Bulletin w:rt
- e of the answers given in
- dcd by Supplement No. 2.
For y
example, since the Bulletin response e November 1, 1980, the equip-ant t' t -
m extended from April 14 to 4
i based on the TMI lessons learr..
ired to be added or upgraded 1:, required to be addressed in i
the Bulletin responte. Altheup.
',t' that was stated in the first su pi r ferent from the original position required to be installed before the in e-
- Bulletin, the TMI equipment was l
Safety Evaluation Report and necced to M voressed by the Itcensee.of the staff's I
following positions were stated by th-The i
By June 30, 1982, all safety-r:1:.
.1;ctrical equipment that could be exposed to a harsh environment in ;. nts Itcensed to operste on or before June 30, 1982, j
to NUREG-0588, as appropriate.must be gurlifi-d to either the DDR Guidelines or i
Safety-related electrical equi w. :
those required to bring the j
2 plant to a cold shutdown condition
--d to mitigate the consequences of 1
an accident.
Note:
This is a s1;,1fic:r*
Nestion as to what equipment i
was required to be o.i.
.. and it is not clear to what extent Itcensees :.=11y implemented this position i
since 10 CFR 50.49 did n:t specifically require it.
s i
1.fcensees stost also evaluate and docu-nt the qualification of safety-related electrical equipment to f~nctien in eavironmental extremes that are not associated with accident conditions.
Qualification for elld environments must be completed by June 30, 1982.
i All operating reactors as of May 23. DSO, will be evaluated against the i
DOR Guidelines.
In cases where the Dr. Guidelines do not provide sufficient detall but NUREG-0583 Cat. 11 does, the NUREG will be used.
i All plants licensed after May 23, 10:0, shall confors to NUREG-0588.
j As stated in Regulatory Culde 1.89, f"l'D-05BS Cat. I requirements are applicable to facilities whose constrr: tion parett 5ER is dated July 1, 1974, or later.
NUREG-0588 Cat. !! r: airements are applicable to i
ra
\\
mmmmmet n
e'acilities whose constructirn unless the licensee ce--ittM
~P is dated before July 1,1974, Cat. I requirements).
'!EE Std. 323-74 (i.e., NUREG-0588 1
The staff indicated that while there are differences brtween the Category !! column of HUREG-0P1 rad the DOR Guidelines, the differences are in detatis and in the cpticral ; art of the documents.
The minimum requirements set forth by these documents are general and compatible, applicable to ors and to NT0Ls.and the minimum standards set by either Issue:
It is not clear ihi" - 1s of the D0R Culdelines and NUREG-0588 the staff -
sidered to be " optional," which makes it difficult *-
' blish to what extent these standards were ret" irmented.
When a deteretnation has been r
t reasonable assurance does not exist that the Class IE electrict.
- isment or component can perform its safety-related fr-tion, that Mition should be reported In general, cps should report via 50.i".(c), and operating plants should subrait a LER.
Issue:
Reporting requiremer,ts n:y be warranted for EQ problems that occur.
The information required to be submitted by November 1, 1980, is for safety-related electrical equipment which could be exposed to a harsh environment during an accident, and r"st function to mitigate the accident (LOCA and HELB inside or cuistde containment) and/or to bring the plant to a cold shutdown condition. Equipment implied by the post accident sampling and moniter'r' equipment, and radiatione monitering equipment, must be addrc: -d.
Equipme.it added to the emergency procedures as a result of 7.1 lessons learned must also be addressed.
Systems and equipeent rust be included regardless of the original classification that was made when the plant received its operating license, and the Q list may be incomplete in this regard.
Note:
This provides significant clarifying information regarding the scope of the 00R Guidelines and NUREG-0588, and it is not clear to what extent this position was implemented by licensees.
A listing of equipment and functions typically needed to sitigate a LOCA or MSLB accident and/or to bring the plant to a cold shutdown condition was provided for information.
The methods and procedures of IEEE Std. 650, " Standards for Qualification of Class IE Static Battery Chargers and Invertors for Nuclear Power Generating stations," relating to design stress analysis, aging of electrical / electronic components, and stress testing are acceptable for qualifying the balance of plant components which are not exposed to harsh environments.
- {
K-5 o
g, _.,_ _ -
+"
~
3 j
l I
Both the 00R Guidelines and HUREC-05C3 accify that sufficient fnformation must be available to verify that safety-related electrical i
equipment has been qualified in acccrduce with the appilcable guidance l
and requirements.
Detatis regarding tM information and documentation required for type tests, operating experience, analysis j
and extrapolation of test data from operating experience, ar,e provided in Section 5 of NUREG-0588 and in Se: tion 8 of IEEE Std. 323-74, t
i Note:
}
This provides significant clarification regarding documentation requirements for power reacters subject to the
{
00R Guidelines, and it is not clear to what extent Itcensees have implemented this positten since it was not specifically j
required by 10 CFR 50.49.
The staff will accept sumary test reports maintained at the utility's 1
central file which reference the actual test reports and data available i
in a single location at the NSSS vendor's facility. The licensee or applicant must make the determination that necessary information and documentation to support qualification of equipment is in confomance i
with the 00R Guidelines and NUREG-0538, as appilcable.
This vendor information file must be maintained current, auditable and 3vailable j
throughout the life of the referencing plant.
i
{
Test reports are not required to be submitted.
Test report references must be included in the plant submittals and these reports sust be I
available for staff review on demand.
The staff has proposed Technical Specifications changes to the Commission which would require that licensees establish and maintain j
centrally located files containing the information necessary to verify the qualification adequacy of all safety-related electrical equipment.
1
{
Procedures for accessing the data base currently being developed by the staff should be available to the industry oy mid-1981.
\\
The format for submitting the required infor:ation should be in accordance with the fonnat that was provided in IE Bulletin 79-018 or in f
j accordance with the letters that were.ent to the licensees of plants in j
the SEP progras.
Either format is acceptable.
1 l
Tha staff does not require nuclear instrumentation or its associated j
components to be qualtfled for a LOCA or HELB.
The nuclear i
instrumentation system is used for transient conditions but is not
}
The staff does require that equipment signed to perform its safety-related function within a short time into an event be qualtfled for a period of at least one hour in excess of the time assumed in the accident analysts.
The staff has indicated that time is the most i
signtficant factor in terms of the margins required to provide an i
acceptable confidence level.
This judgement is based on the acceptance 1
of a type test for a single unit and the spectrue of accidents (small and large breaks) bounded by the single test.
K-6 1[ 7 NddLE M
- - ~ ~
' ~ ~ ~
~
_ m m m aeairlatif,J,% B %.
1 i
i Issue:
It is not clear how an ' acceptable confidence level' is established for electrical equipment that must operate
- 1ong-term" that is comensurate with the 1-hour utnista i
test time requiree.ent that was established for "short-tors' electrical equipment.
Sequential testing requirements are spectfled in NUREG-0588 and in the 00R Guidelines.
l spp1tcable document. Licensees must follow the test requirements of the i
j Note:
Only NUREG-0588, Category I provides specific guidance with j
regard to test sequence (see, Appendix J).
i If the test has been completed without aging in the sequence, a.
justification for such a deviation must be submitted.
i b.
If testing of a given component has been scheduled but not j
inttlated, the test sequence / program should be modified to include aging.
I j
c.
Test programs in progress should be evaluated regarding the ability to comply by incorporating aging in the proper sequence.
Note:
This provides significant clarification pertaining to the i
00R Guidelines and NUREG-0588, Category II; and it is not j
clear to what extent this position was leplemented by
?
licensees since it was not spectfica11y required by 10 CFR 50.49.
i When test failures occur due to non-EQ related problems, it may be difficult to extrapolate the test data to demonstrate qualification.
{
may be more prudent to correct the failure and continue with the test.
It
,The parameters to be considered for harsh environment outside i
containment (e.g., a spectrum of main steam and feedwater line breaks,
{
etc.) are temperature duration of exposure,, aging, and submergence. pressure, humidity, caustic spray, r
Mechanical and flow-l Further guidance for selecting the piping systems and co reviews are delineated in Regulatory Guide 1.46 and Standard Review Plan Section 3.6.1 and Section 3.6.2.
j
!ssue:
EQ with regard to mechanical and flow-induced vibrations and
}-
seismic effects may not be adequate.
In genera.l, the containment temperature and pressure conditions as a function of time should be based on the FSAR analyses.
The 00R Guidelines and NUREG-0588 provide guidance and considerations for determining if the plant-specific profiles encompass the LOCA and HELS condittohs inside containment.
As an alternative, the generic profile i
provided for pressure-suppression-type containments can be used as long as the pressura and humidity conditions as a function of time are accounted for.
t i
\\
K-7 I
e
- w In determining the qualification status of existing equipment purchased from a vendor, where a OA program did not exist, the utility should consider the following factors:
The complexity of design, complexity of the manufacturing process, a.
and end use.
b.
Past performance of the vendor.
Past operating history M products, especially stallar products, c.
made by the vendor, d.
Procedures, equipment, and results of environmental.qua11(f eation testing relative to other equipment for which a QA progras was applied.
Note:
This provides significant clarification ragarding qualification considerations and teiidt to relax the l
positions contained in the DOR Guidelines and in NUREG-0588, and it is not clear to what extent licentees have used this methodology.
After May 1980, all replacement and spare parts used to replace presently installed parts shall be qualified to NUREG-0588 Cat. I requirements "unless there are sound reasons to the contrary."
Nonava11 ability and the fact that the part to be used is a spare part that was purchased prior to May 23, 1980 are among the factors to be considered in weighing whether there are," sound reasons to the contra ry. "
When developing radiation source terms to be used for equipment
~
qualification, consideration must be given to those conditions that are most bounding.
NUREG-0578 (with the RCS assumed to be intact) should be used forF I
guidance.
For non-LOCA HELB conditions outside containment, NUREG-0588 1
(with the RCS assumed to be intact) should be used for guidance.
For conditions inside containment, the larger of NUREG-0578 (with the RCS assumed to be intact) or NUREG-0538 (with the RCS assumed to be either intact or not intact, whichever is worse) should be used.
Gamma equivalents may be used when considering the effects of beta exposure.
Qualification by use has limited application. Often, the equipment has never seen the harsh environment and no conclusions can be drawn as to its operability in a harsh environment.
A determination of 'long term
- for qualification of equipment should be based on:
a) the time period over which the equipment is required to bring the plant to cold shutdown and to mitigate the consequences of the accident, and b) the ability to change, modify or add equipment to provide the same safety-related function during the course of the accident or in mitigating its effects.
K-8
I to establish that the component was subjected to the mos temperature environment that is postulated to occur.
These temperature measurements are required to be made as close to the component surface as practicable to ensure that they are representative of the test environment.
The component surface temperature is considered to be a conservative measure of the test temperature environment.
Sucolement No. 3 to IE Bulletin 79-018. " Environmental Qualification of C IE Eautoment." October 24. 1980 The purpose of this supplement was to clarify environmental qualification requirements for equipment required to bring the plant to cold shutdown if the licensing basis of the plant is a hot safe shutdswn condition.
i maintaining a cold shutdown condition and qualify the equ The staff that method (path),
Issue:
It is not clear why the single failure criteria was not imposed in this case.
submit qualification information on equipment that was bein result of the TMI lessons learned study, since the previous supplements provided confusing and conflicting guidance on this matter.
Generic letter 81-05. "Information Recardino the Procram for Environmenta Oualification of Safety-Related Electrical Eautoment." January 19. 1981 This generic letter was issued to provide information regarding the EQ requirements that were established by Commission Memorandum and Order of j
May 23, 1980 (80-CLI-21).
generic letter included equipment necessary to achieve cold shutdown, replacement parts, Three Mile Island (TMI) Action Plan (NUREG-0737) equipment and the June 30, 1982, implementation deadline.
a.
Cold Shutdown The intent of !&E Bulletin 79-01B Supplement No. 3 was not to change the Itcensing basis - plants licensed to a hot " safe shutdown" condition are only required to qualify that equipment necessary to achieve hot shutdown.
to be implemented on a case-by-case basis.However, the Bulletin provided a Regulatory Guide 1.139 which EQ is a part.contains the implementation plans for cold shutdown requirement I
I b.
Replacement Parts EQ requirements for replacement parts are stated in 1&E Bulletin 79-018 Supplement No. 2.
deviations from the Category I requirements of NUREG-0588 in anIt is auditable manner.
l K-9 i'
gg"E;".a'vm N -
a NUREG-0737 Eculoment.
c.
I The qualification requirements for this equipment are described in Appendix B of the NUREG.
NRC is contained in I&E Bulletin 79-018 Supplement No. 3.The 1
Note:
It is not clear to what extent licensees have implemented the requirements stated by NUREG-0737, since these requirements are not specified by 10 CFA 50.49.
5 d.
June 30. 1982 Deadline Since the deadline was imposed by an Order, the staff was not authorized to grant relief.
- Environmental Oualification of Safety-Related Electrical Ecutoment." Acril 20. 1982 This generic letter was issued to provide clarification regarding certain EQ requirements, a.
Doerator Disolav Instrumentation Licensees need only identify and have available qualification documentation on those operator display instruments which are safety-related.
After other related NRC review activities are completed, relatef and/or qualified. additional instrumentation may be required to
~
b.
Safety-Related Eculoment The Consissien Memorandum and Order (80-CI.1-21) only required environmental qualification of safety-related electrical equipment.
]
Identification of safety-related equipment installed in harsh environments at specific plants must be supplied by the licensee.
4 other NRC reviews. necessity to upgrade other equipment and systems will The c.
Reolacement Parts Conditions were identified which reflect sound reasons why qualification standards for replacement of equipment in a harsh environment need not be upgraded to NUREG-0588, Category I, requirements.
]
d.
Mild En_vironment Established that equipment qualification for existing equipment located in a mild envircnment can be demonstrated by use of a periodic maintenance, inspection, and/or replacement pro program; and an equipment surveillance progra:a. gram; a periodic testing equipment, the licensee must also establish and document theFor replaceme environmental design basis for the equipment locations and purchase j
equipment accordingly.
X-10 i
i- -
'%g e.
Submeroence Outside Containment Safety-related equipment that-could become submerged due to a high energy line break outside containment must be qualified.
f.
Radiation Clarified that the screening value of 4E7 rads is only appitcable to PWRs with dry containments. Other plants must use plant-specific analysis, and PWRs with dry containments may also use plant-specific-analysis.
l g.
Containment Service Conditions Provided clarification that thef staff values of T(sat) for PWRs and T(sat) + 20*F for'8WRs need not be used as the maximum in-containment temperature for equipment qualification - the temperatures can be based.
~
on other NRC approved analysese
_t h.
One Hour Minimum Ooeratino Time Clarified that the one hour minimum operating time need not be applied to D0R guidelines or NUREG-0588, Category II, equipment that was tested prior to May 23, 1980. The test data and analysis may be used to qualify the equipment to the required operating time plus an appropriate margin. Also, the one hour requirement-is not applicable to equipment whose safety function is performed prior to significant changes in the environment at the equipment location.
l i.
Aglag cl' Clarified that the DOR guidelines (Section 7) and MUREG-0588,
[
Category II (Section 4.2), do not require a qualified life to be established for all safety-related electrical equipment located in harsh i
environments. However, a quallfled. life is required for equipment,'
including replacement parts, qualified to NUREG-0588, category I, requirements that is located in a harsh environment.
-s
- w w,
-9 Generic Lt 'er 85-15. "Information Re'atina to the Doadlines for C - liance with 10 F R 50.49. ' Environmental Qua' ification of E'vetric Eauinment Imoortant to Safety for Nuclear Power Plants.'" Aunust 6. 1985 l
The Generic Letter was issued to remind the industry of the. requirements
[
stated by 10 CFR 50.49 for environmental qualificattor, of electric equipment.
l The rule established that the second refueling outage after March 31, 1982, or March 31, 1985, whichever is earlier, was the deadline for qualifying electric equipment important to safety. However, extensions to this deadline could.be granted by the Director of NRR not to exceed November 30, 1985. Only the Commission could grant extensions beyond November 30. The Generic Letter advised licensees that the Commission would grant extensions only in rare circumstances and that enforcement action would be taken against licensees that continued to operate their plants with unqualified equipment beyond d
K-11
.j t
l:
l
=
re.
i November 30, 1985, without extensions approved by the Comission. Specific guidance was provided with regard to how civil penalties would be assessed.
e
]
Generic letter 86-15. "Information Relatino to Comoliance with 10 CFR 50.4L.
' Environmental Oualification of Electric Eautoment Imoortant to Safety for Nuclear Power Plants.** Seotember 22. 1986 The staf f provided further guidance on appropriate actions that licensees should take in situations where environmental qualification of equipment'is i
j suspect and on current NRC policy with regard to enforcement of 10 CFR 50.49.
it was the staff's position that when a licensee discovers a potential defstiency in the environmental qualification of equipment (i.e., a licensee does not have an adequate basis to establish qualification), the licensee shall make a prompt determination of operability, take imediate steps to establish a plan with a reasonable schedule to correct the deficiency, and h n e a written justification for continued operation. The licensee's justification did not require NRC review and approval.
1 The Generic Letter indicated that the licensee may be able to make a finding of operability using analysis and partial test data to provide reasonable l-assurance that the equipment will perform its safety function when called upon to mitigate the accidents for which it is needed.
if the licsnsee is unable t
to demonstrate operability, the provisions of the plant Technical Specifications must be followed.
For equipment that is not included in the le d nical Specifications, the licensee may continue reactor operation if the safety fonction can be accomplished by other equipment that is qualified or if
~
limited administrative controls can be used to ensure the safety function is performed.
The enforcement guidelines related to Generic Letter 85-15 were included as an enclosure.
Generic Letter 88-07. " Modified Enforcement Policy Relatino to 10 CFR 50.49.
' Environmental Oualification of Electric Eautoment Imoortant to Safety for Nuclear fower Plants.'" Acril 7. 1988 Upon review, the Comission found that the EQ enforcement policy stated in l
Generic Letter 85-15 and Generic Letter 86-15 could result in impositier. of civil penalties that did not properly reflect the safety significance of EQ r
violations with respect to civil penalties imposed in the past for other i
issues (i.e., the EQ civil penalties could be inflated by comparison).
Therefore, the Comission determined that the EQ enforcement policy should be revised.
Generic Letter 88-07 promulgated the staff's revited enforcement j
policy for E0 violations which more closely reflected the relative safety l
importance of E0 violations as compared to other enforcement issues. However, Generic Letter 88-07 reiterated the actions that licensees should take when a potential E0 deficiency is found as' listed previously in Generic Letter 86-15.
I i
K-12
i i
w APPDf0!X L -
1 ResearchActivitiesand8tudiesRelatedtotQ i
A substantial amount of IQ research has been conducted by the NRC and also by f
the nuclear industry in ordsr to address questions and concerns that have been i
ratted in this area.. Much of this work was done in order to address concerns that were expressed by t,he Valote of Concerned telentists in a Petition for i
Comission Action that was submitted in November 19771 and mere Petehtly, i
additions) research has been conducted partly in support of the license I
renewal initiative, but also to address plano, aging concerns.
This appendix providen a numary of the research studies that were reviewed by the staff in i
j laarch of potenttal C0 tttuea und concerna.
nated on a review of the.many research studies that are sumarised below, i
quite a number of potential tatues were identified for further staff tonstdoration.
These, along with several noteworthy coments, are llated in 4
l bold print for ease of iden,1fication.
$0 0413. 'llaf f Renartage fnviranmental nualtflaattan af anfatv.Ralated j
. lu GIcY Inu nment.' TabruarwElR7g,
2 The staff performed a cualitative assessment of the environmental qualification requirements for electrical equipment and in particular, focused on the adequacy of those requirements t11th regard to older plants (i.e.f the those i
that were currently operating. The staff also examined the adequacy o (0 methodology beinq used for)the newer plants,r,taff conclu ed tha, there was retsenable assuranet that pub)ite health and i
lased en it review, the l
tafety was adequately protected in the int 6 rim while a more systematic, complete resolution of the issue was pursued, and that no additional imediate acttoni were warranted, With regard to the adequacy of qualtf tcation j
reoutrements, the following information was presentedt i
i (nvironmental condittens of interest intluded these that would exist Inside containment due to a LOCA or MILD event, Qualification parameters of interest included chemica) espesure and submergance, in
[
addition to temperature, redletion, hur.idity, and pressure. Aging was not addretted.
Prior to 1976, equi ment quallf tention was primarily based on conditions resultingfromaL0bA, However, in 1978 the staff determined that a postulated main steam line break in PWR= type plants with dry l
containments could result in temperatures tigier than those that were i
predictedforaLOCA(butonlyforashortduration).
Note The staff's resolut10n of Yask Action Plan A 81 falls to recognise this, (quipment that wa* required to be qualtfled included the ICC1 and l
containment iso).'llen and cleanup systems, as well as instrumentation and controls nect'sary to isolate a steam generator affected by a H5LB i
accident.
In pcrticular, the instrumentatlon needed to initiate salvty systems and to provide diagnostic information to the plant operatori j
(e.g., electrical penetrations into containment, any electrical i
l l
l o
connectors to cabling which transmits signals, and the instruments themselves), and control power to certain motor operated valves (e.g.,
ECCS and containment isolation valves located inside containment) were of concern.
l For older nuclear power plants (those that were reviewed by the staff 4
prior to about 1967), specific nuclear environmental qualification i
requirements had not been established in the industry or by the Comission.
Licensing conclusions at that time were based on an awareness that nuclear components were of "high industrial quality.'
Plants reviewed during the period from 1967 through 1974 were assessed
)
j against IEEE Standards 279-68 and 323-71 (or perhaps to some preliminary
)
form of these standards) as they became available.
Plants that were l
J reviewed by the staff after July 1974 were evaluated against the "more comprehensive' guidelines specified in IEEE Std. 323-74 and associated J
ancillary standards as they became available.
Section 3.!! of the i
Standard Review Plan was issued in September 1975 and provi6ed specific 4
review guidance for these later facilities.
There have been occasions when licensees have been directed to perform j
system upgrades to satisfy EQ concerns.
For example, on June 23, 1976, the staff ordered that the reactor protection system be upgraded to meet i
4 j
IEEE 279-68 and that other safety-related equipment not previously environmentally qualified be qualified or replaced. Also, upgrades were required to satisfy the new ECCS requirements (10 CFR 50.46), but licensees were >ensitted to establish environner.tal qualification based on either sultaale analyses or test results.
Operating events, though not entirely representative of design basis accident conditions. established some degree of confidence in the i
environmental qualification of electrical equipment that was used in operating reactors.
Specific events appitcable to BWRs included:
,a.
A reactor scram at Dresden Unit 2 on June 5, 1970, followed by a main steam line safety valve discharge into containment; 250,000 pounds of primary coolant were released into containment causing i
an increase in drywell pressure to about 20 psig at 320*F.
Although some equipment damage did occur, the ECCS and containment isola' ion systems functioned properly, j
j b.
An event similar to the Dresden Unit 2 event occurred at Dresden Unft 3 on December 8, 1971; drywell pressure increased to about 20 psig at 295'F. Again, while some equipment damage occurred (although much less than during the Dresden Unit 2 event), all systems necessary for safety remained operable.
c.
Other steam discharge events have occurred at other BWRs in which the drywell was only slightly pressurized and tedperatures remained below 200*F, tio damage to any electrical systems was reported for any of these events.
Events applicable to PWRs included:
i L-2
l i
l 1
i a.
A partial depressurization of the primary system occurred at Davis-Besse Unit 1 on September 24, 1977, which resulted in a release of about 11,000 gallons of water (in the form of steam) into containment through the pressurizer quench tank rupture disk.
t The steam release occurred over approximately 15 minutes and the j
only equipment damaged was within the vicinity of the steam discharge; no equipment that was needed to mitigate the event was j
damaged.
i b.
A break occurred in a feedwater line at Indian Point Unit 2 on i
November 13, 1973, The containment reached about 110*F and a relative humidity of about 501, but no containment pressurization was reported. One cable tray was partially submerged in water (the cable was designed and tested for submergence), but no adverse effects on electrical equipment were found.
c.
While in a hot shutdown condition, a reactor coolant pump seal failure occurred at H. B. Robinson Unit 2 on May 1, 1975, which resulted in the discharge of about 130,000 gallons of reactor coolant in containment. The containment was pressurized to about 3 psig with a temperature of about 140'F. Although high i
humidity conditions remained in containment for several hours, no adverse effects on electrical equipment were reported.
i Other considerations that provided a level of confidence in the j
environmental qualification of slectrical equipment at operating plants Included equipment timing considerations coupled with redundancy requirements (that is, much of the safety equipment is designed to function very early into a LOCA, before the environment can be i
significantly affected, such that at least one of the redundant places l
of equipment is sure to function); the inspection progras under the Office of Inspection and Enforcement e phasized review of environmental l
Qualification test results for engineered safety systems; and operating experience reviews (i.e., review of licensee event reports) helped to identify whether electrical equipment was degrading.
1 On a more generic level, the staff had reviewed the qualification i
a i
documentation related to connectors and penetrations. Based on these
]
4 ef forts, the staff concluded that there were sufficient alternatives to l
assure safety while the qualification of components was fully
)
established. Additionally, the staff was relying on the Systematic i
Evaluation Program (SEP) to determine whether electrical cabling built to earlier standards wa; adequately qualified.
With regard to SEP, licensees of eleven of the oldest operating plants would be required to evaluate the environmental qualification of all electrical equipment that they deemed necessary to mitigate the consequences of design basis events.
The staff planned to review the information developed by the I
eleven SEP licensees and determine if any specific issues remained to be addressed either individually by the l' 'nsee or on a more generic basis.
In the event that documentation of enviro' wrtal_ testing did not exist or was insufficient to assure environmental qualification, the following alternatives would be considered by the staff:
]
L-j f
1 l
Evidence of qualification of identical or stellar equipment, either in another nuclear facility or by another industry.
4 Importanca of the safety function associated with any questionable i
equipment would be considered. Demonstration of adequate facility response to all DBEs without credit for the function of an unqualified
)
component may be justification for not requiring environmental qualification of a specificEcomponent.
I If important safety equipment was not environmentally qualified, l
consideration would be given to alternate ways of performing the safety i
function by using different systems, including the use of non-safety _
grade systems.
s Consideration would be given to possible means of protection components j
from adverse environmental conditions, such as by enclosing it, coating j
lt, or providing other protective features.
Other alternatives that proposed by the licensee would also be 4
considered.
I Note:
The above stated methodology, if it were used for the SEP j
review, would not provide assurance of EQ at these older i
facilities.
The other issue considered by the staff dealt with the adequacy of current i
environmental qualification testing requirements. With regard to this issue, i
the staff recognized that NRR had established Technical Activities A-21 and A-24 related to main steam line break considerations inside containment and related to qualification of safety-related equipment, respectively. Ongoing i
confirmatory research included an assessment of synergistic effects (i.e.,
squential vs. simultaneous EQ testing), confirmation that accelerated aging methodology can be used, and definition of the radiation source term. The study indicated that, as part of the accelerated aging research, cable elongation was being used as a relative damage ind'cator for comparison with naturally aged cables; tests to determine rate effects were underway (in particular those of oxygen diffusion and radiation); and a study of alternate damage indicators was underway for use in conjunction with elongation; and naturally aged samples were being collected to be used for comparison j
purposes.
i.
These same issues are still being pursued today, over fifteen j
Note:
years later.
I included as a major part of this qualitative assessment, was a chronology of l
the development of EQ requirements, industry standards, NRC review criteria, and a grouping of plants by their construction permit (CP) application dates.
Some of this information is provided in Tables L-1 through L-3.
4 NUREG-0458. "Short Tern Safety Assessment on the Environmental Qualification I
of Safety-Related Electrical Eautoment of SEP Doeratina Reactors." May 1978 i
I in its report to the Commission of December 15, 1977, "NRC Staff Report on Union of Concerned Scientists' Petition for Emergency ~and Remedial Action" j
L-4 L
i
TABLE L-1 Development of IEEE Standards Related to EQ IEA8 IEEE STANDARD 1968 IEEE 279-68 " Proposed IEEE Criteria for Nuclear j
Power Plant Protection systems' i
1971 IEEE 279-71 (revision to the 1968 standard) i IEEE 317-71, " Standard for Electrical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations' IEEE 323-71. " General Guide for Qualifying Class !
Electrical Equipment for Nuclear Power Generating Stations" 4
4 IEEE 334-71, " Type Tests of Continuous Duty Class ! Motors Installed inside the Containment of Nuclear Powered Generating Stations" j
1972 IEEE 317-72 (revision of the 1971 standard)
IEEE 382-72, " Type Tests of Class ! Electric Valve Operators for Nuclear Power Generating Stations" l
1974 IEEE 323-74 (revision of the 1971 standard) i' IEEE 334-74 (revision of the 1971 standard)
IEEE 383-74 "Ty e Tests of Class IE Electric Cables.FieId$ptees,andConnectorsforNuclear i
Power Generating Stations" l
1976 IEEE 317-76 (revision of the 1972 standard) l 1977 IEEE 381-77, " Type Tests of Class IE Modules Used j
in Nuclear Power Generating Stations" I
4 L-5 j
TABLE L-2 Development of AEC/NRC Requirements and/or Guidance Documents Related to EQ j
11AB AEC/NRC REOUIREMENT AND/OR GUIDANCE DOCUMENTS 1965 Draft General Design Criteria (GDC) Published for Coment (11/65; GDC-16) 1966 AEC Peb11shed a Guide for Preparation of Safety Analysis Reports (6/66) 1967 Revised GDC Published for Coment (7/67; GDC-4, 23)
" Codas and 50 (150.55a)1/69) 1969 Revision to 10 CFR Part Standards," Pubitshed for Coment (1 1971 GDC Incorporated into Regulations, Appendix A to 10 CFR Part 50 (5/71; GDC-4, 23) 10 CFR 50.55a, " Codes and Standards " Incorporated intoRegulations(7/71;endorsedIEEEStd.279) 4 AEC 1.G.2 Published (10/71) 1972 AEC Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports," Rev. 1 Issued (10/72) 1973 AEC Regulatory Guide 1.40 issued (3/73; endorsed IEEE 334-71)
AEC Regulatory Guide 1.63 issued (10/73; endorsed IEEE 317-72) 1974 AEC Regulatory Guide 1.73 issued (1/74; endorsed IEEE 382-72) i AEC Regulatory Guide 1.89 Issued (11/74; endorsed IEEE 323-74) 1975 NRC Regulatory Guide 1.70, Rev. 2, Issued (9/75; revision of the 1972 R.G.)
NRC Standard Review Plan issued (9/75) 1977 NRC Regulatory Guide 1.63, Rev. 1, Published i
for Coment (5/77; revision of 1972 RG endorsing IEEE 317-76)
NRC Regulatory Guide 1.131 Pubitshed for Coment (8/77; endorsed IEEE 383-74)
- 1. - 6 l
y J_) g.
l:'
~
I l
TABLE L-3 FacilityGroupin9byCP!ssuedate Facility Name/ Criteria
Pre-1967 Dresden Unit 1 (DOR)
P 5/56 9/59 Yankee-Rowe (DOR) 11/57 7/60 Humbolt Bay Unit 3 (DOR)
'11/60 8/62 Big Rock Point CDOR) 5/56 3/62 5/60 8/62 Indian Point Untt 1 (00R)
San Onofre Unit 1 (DOR) 3/64 3/67 Haddam Neck (DOR) 5/64 6/67 Lacrosse (DOR) 3/63 7/67 B.
1967-1974 Oyster Creek (DOR) 12/64 4/E9 i
Nine Mile Point Unit 1 (DOR) 4/65 8/69 Dresden Unit 2 (00R) 1/66 12/69 Ginna (DOR) 4/66 9/69 Millstone Unit I (DOR) 5/66 10/70 Oresden Unit 3 (DOR) 10/46 1/71 Indian Point Unit 2 (DOR) 10/66 10/71 Ouad Cities Unit 1 (00R) 2/67 10/71 Quad Cities Unit 2 (DOR) 2/67 3/72 Palisades (DOR)
.3/67 3/71 Robinson Unit 2 (DOR) 4/67 7/70 Turkey Point Unit 3 (DOR) 4/67 7/72 4
Turkey Point Unit 4 (DOR) 4/67 7/73 i
0 owns Ferry Unit 1 (DOR) 5/67 6/63 Browns Ferry Unit 2 (DOR) 5/67 6/74 Monticello (DOR) 6/67 9/70 i
i Point Beach Unit 1 (DOR) 7/67 10/70 Oconee Unit 1 (DOR 11/67 2/73 l
Oconee Unit 2 (DOR 11/67 10/73
)
Oconee Unit 3 (00R 11/67 7/74 i
Vermont Yankee (DOR) 12/67 3/72 i
Peach Botton Unit 2 (DOR) 1/68 8/73 Peach Botton Unit 3 (DOR) 1/68 7/74 Diablo Canyon Unit 1 (Cat. 11) 4/68
[11/84)
Three Mile Island Unit 1 (DOR) 5/68 4/74 Cooper (DOR) 6/68 1/74 Ft. Calhoun (DOR) 6/68 5/73 i
Prairie Island Unit 1 (DOR) 5/68 8/73 Fratrie Island Unit 2 (DOR) 5/68 10/74 j
Surry Unit 1 (DOR) 6/68 5/72 l
Surry Unit 2 (00R) 6/b8 1/73 i
Point' Beach Unit 2 (DOR) 7/68 11/71 l
Browns Ferry Unit 3 (DOR) 7/68 7/76 l
L-7 l
l
i g'
- l l
TABLE L-3 (cont.)
j Facility Grouping by CP !ssue Date j
Factitty Name/ Criteria *'
CP Issue Data OL Issue Date 8.
1967-1974 (cont.)
i Kewaunee (00R) 8/68 12/72
)
]
Pilgrim Unit 1 (00R) 8/68 6/72 1
)l Ft. St. Vrain (00R) 9/68 12/73 l
Crystal River Unit 3 (00R) 9/68 12/76 i
Sales Unit 1 (DOR) 9/68
'[12/76?
Salem Unit 2 (Cat. II) 9/68
[5/81; Rancho Seco (DOR) 10/68 8/74 Maine Yankee (DOR) 10/68 9/72 Arkansas Unit 1 (DOR) 12/68 5/74' 2
l Zion Unit 1 (DOR) 12/68 4/73 Zion Unit 2 (00R) 12/68 11/73 D.C.CookUnit1(00Rl 3/69 10/74 i
D. C. Cook Unit 2 (D0R, 3/69 (12/77)
CalvertCliffsUnit1(DOR) 7/69 7/74 Calvert Cliffs Unit 2 (DDR) 7/69 8/76 Indian Point Unit 3 (00R) 8/69 12/75 v
i Hatch Unit 1 CD0R) 9/69 8/74 Three Mile Is'land Unit 2 (D0R) 11/69 (2/78) i Brunswick Unit 1 (DOR) 2/70 9/76 Brunswick Unit 2 (DOR) 2/70 12/74 Fitzpatrick (00R) 5/10 10/74 Sequoyah Unit 1 (Cat. !!)
5/70 19/80' i
l 9/81)I Sequoyah Unit 2 (Cat. II) 5/70 i
Duane Arnold (00R) 6/70 2/74 Beaver Valley Unit 1 (D0R) 6/70 1/76 Diablo Canyon Unit 2 (Cat. II) 12/70
[8/85)
St. Lucie Unit 1 (00R) 7/70 3/76 Millstone Unit 2 (DOR) 12/70 8/75 i
Trojan (DOR) 2/71 11/75 l
North Anna Unit 1 (00R) 2/71
[4/78h North Anna Unit 2 (Cat. II) 2/71 (8/801 Davis-8 esse (D0R) 3/71 4/77 Farley Unit 1 (DOR) 8/72 6/77 Farley Unit 2 (Cat. II)~
7/72 f3/81)
Fami Unit 2 (Cat. II 9/72 17/85)
Zimmer Unit 1 (Cat. !)
)
10/77 (N/A)
Arkansas Unit 2 (DOR) 12/72
[9/78)
Midland Unit 1 (Cat. !!)
12/72
'N/AL Midland Unit 2 (Cat. !!)
12/72 lN/A Hatch Unit 2 (D0R) 12/72 6/78)
Watts Bar Unit 1 (Cat. II)
.1/73 N/A).
Watts Bar Unit 2 (Cat. II) 1/73 N/A) i L-8 7
Wi~'l J ~
- s TABLE L-3 (cont.)
Facility Grouping by CP issue Data facility Name/Criterta*
CP Issue Date OL Issue Date B.
1967-1974 (cont.)
McGuire Unit 1 (Cat. II) 3/73
[7/81 1
,4/84l
,5/83 1
McGuire Unit 2 (Cat. !!)
3/73 Washington Nuclear 2 FCat. !!)
3/73 Sumer Unit 1 (Cat. IL) 3/73 (1/82;
- N/A)?
4/89 Shoreham (Cat. !!)
4/73 Forked River (Cat. II) 7/73 LaSalle Unit 1 (Cat. !!)
7/73
[.2/821 LaSalle Unit 2 (Cat. 11) 7/73 l3/84l1 San Onofre Unit 2 (Cat. !!
10/73 9/82' SanOnofreUnit3(Cat.Ils)
(11/82l
'9/83 10/73 l
Susquehanna Unit 1 (Cat. 11) 11/73 Susquehanna Unit 2 (Cat. II) 11/73 6/84;
[N/A' Bailly Unit 1 (Cat. II) 5/74 (8/8Q Beaver Valley Unit 2 (Cat. 11) 5/74 Limerick Unit 1 (Cat. 11) 6/74
,8/85 i Limerick Unit 2 (Cat. II) 6/74 l8/89l Nine Mile Point Unit 2 (Cat. II) 6/74 ll3/87,;
7/87 Vogle Unit 1 (Cat. !!)
6/74
- 3/89J Vogle Unit 2 (Cat. II) 6/74 C.
Post-July 1974 North Anna Unit 3 (Cat
!)
7/74
[N/A North Anna Unit 4 (Cat. 1)
- '/74 lN/A Millstone Unit 3 (Cat. 1) 8/74
- 1/8 ]
Grand Gulf Unit 1 (Cat. I) 9/74
(.ll/84)
Grand Gulf Unit 2 (Cat. 1) 9/74
'N/A)'
Hope Creek Unit 1 (Cat. 1) 11/74
?7/86)
Hope Creek Unit 2 (Cat. I) 11/74 lN/A) i Waterford Unit 3 (Cat. 1) 11/74
- 3/85)
Comanche Peak Unit 1 (Cat. 1 12/74
'4/90'-
Comanche Peak Unit 2 (Cat. !))12/74
'4/93h Surry Unit 3 (Cat. 1) 12/74
[N/A)
Surry Unit 4 (Cat. 1) 12/74 "N/A?
Bellefonte Unit 1 (Cat. 1) 12/74 lN/Aj Bellefonte Unit 2 (Cat. 1) 12/74
[N/A) i Catawba Unit 1 (Cat. I) 8/75
[l/85)
Catawba Unit 2 (Cat. 1) 8/15 L5/86)
South Texas Unit 1 (Cat. I) 8/75 L3/88)
South Texas Unit 2 (Cat. I) 8/75 (3/89)
Washington Nuclear 1 (Cat. 1) 12/75 (N/A) l Byron Unit 1 (Cat. 1) 12/75
!;2/85 i
Byron Unit 2 (Cat. I) 12/75 Ll/87 L-9 f-l
I 1
TABLE L-3 (cont.)
Facility Grouping by CP !ssue Date j
l Facility Name/ Criteria' CP lssue Date OL Issue Date C.
Post-July 1974 (cont.)
Braidwood Unit 1 (Cat. 1) 12/75 l7/87p Braidwood Unit 2 (Cat. 1) 12/75
'5/88p Clinton Unit 1 (Cat. 1) 2/76 l4/87, Cilnton Unit 2 (Cat. 1) 2/76 LN/A)
Seabrook Unit 1 (Cat. I 2/76 l5/89)
Seabrook Unit 2 (Cat. I 2/76
.lN/A;l Callaway Unit 1 (Cat. I 4/76
[10/80)
Callaway Unit 2 (Cat. 1) 4/76 lN/A)
River Bend Unit 1 (Cat. 1) 3/77
[11/85)
([N/A)'I River Bend Unit 2 (Cat. 1) 3/77 Palo Verde Unit 1 (Cat.1) 5/77 6/85 Palo Verde Unit 2 (Cat. 1) 5/77 4/86'l Palo Verde Unit 3 Cat.Ip 5/77
[;ll/87h Hartsville Unit 1 Cat. I l
5/77 JN/A1 Hartsville Unit 2 Cat.Ih
'5/77 i N/A' L
Hartsville Unit 3 Cat. Il 5/77 JN/A HartsvilleUnit4(Cat.Ih 5/77 llN/Al Perry Unit 1 5/77
[:ll/86)
PerryUnit2(Cat.1)
(Cat. 1)
$/77
'N/A))
'6/85 Wolf Creek Unst 1 (Cat. I) 5/77 St. Lucie Unit 2 (Cat. I) 5/77 j6/83)
Steritng Unit 1 (Cat. !)
9/77
'N/A' ShearonHarrisUnit1)
J)/78)
'l/8b)
LWashington Nuclear 3)
.l4/78) lN/A)
The criteria (i.e., 00R, Cat. I, and Cat. !!) was determined strictly based on the CP-SER date and the OL date for the facility.
Therefore, any special exceptions that were allowed by the staff to the criteria stated in Section 5 of this report are not reflected in this table.
[]
Information contained in brackets was not included in NUREG-0413.
(also issued as NUREG-0413 in February 1978), the staff concluded that reasonable assurance existed that electrical connectors and penetrations would perform their required functions in the environment of a loss-of-coolant i
accident (LOCA) and that no imediate remedial action was required for operating reactor facilities. However the staff also concluded that it would beprudenttodetermine,inarelativelyshortperiodoftime,whetherthe scope of the environmental qualification review effort should be expanded from 4
tae two specific types of components (i.e., electrical connectors and electrical penetration assemblies) to include all safety-related electrical i
equipment lucated both inside and outside containment for all design basis events (DBE;).
The Systematic Evaluation Program (SEP) was used as the
)
vehicle for perfor1ning this *short-term" assessment, and NUREG-0458 provided the results of the staff's assessment in *.his regard.
The staff's assessment of this SEP review topic considered:
(a) the ability of the facility to adequately respond to all design basis events; (b) the importance of the safety function and alternate ways of performing the safety function such as the possible use of non-safety systems; (c) all available testing and qualification data; (d) any existing protection of the equipment from the environment; and (e) any other basis for acceptability the licensees may identify.
1he following conclusions were stated in NUREG-0458:
No significant safety deficiencies requiring imediate remedial actions were identified.
The results of the staff's short-term assessment did not change the j
staff's basis and conclusions stated in NUREG-0413.
I The changing and upgrading of older operating plants is a continual process. Although not part of a comprehensive program, this process has provided and continues to provide additional assurance that safety-related equipment will perform its function when required.
Although no imediate remedial actions were considered necessary, the staff re m nended that an IE Circular be issued requesting that licensees submit specific information and to facilitate a review of the installation and 1
i environmental qualification documentation of specific electrical equipment located inside containment of all operating reactors.
This review was to include staff follow-up inspections to resolve any questionable design practices.
Aside from the conclusions and recommendations stated in this report, the following information was also discussed:
A substantial amount of information had been collected on the status of environmental qualification of electrical equipment which showed that for certain components, documentation of environmental qualification as currently required could not be immediately identified.
The questions about environmental qualification of safety-related electrical equipment were, in part, the result of the evolution of regulatory and industry efforts in the development of the current criteria, regulations, and standards.
L-il l
I The design basis for functionability of safety-related equipe nt under I
accident conditions, including the adverse environment caused by the 1
accident, had been recognized for many years and was implicitly included in the design of even the oldest nuclear power plant. However, while j
this general design basis for equipment qualification was reasonably l
l well adopted in the nuclear industry, specific requirements had not been
'i establistedandconsiderablevariationinlicenseeagdesignbasis."proaches existed for the older operating reactors in satisfying this l
For pressurized water reactors JPWRs) Itcensed prior to 1976 the
+
environmentalconditionsforwhichsafety-relatedequipmentInside containment was qualif ted were established by the LOCA analysis.
However analysis by the NRC staff indicated that the calculated temperature inside containment associated with a main steam line bretk (MSLB) could be as much as 100*F to 150*F higher than that predicted for j
i a LOCA for a short seriod of time early in the MSt.B accident. A similar i
condition was posst ale for BWR reactor types with the Mark I containment i
whereby saturated steam becomes superheated during release to the drywell. An event of this nature occurred at Dresden 2 when a transient resulted in reactor system safety valves becoming stuck in a partially open position.
Note:
The staff failed to recognize this information during its
)
resolution of Task Action Plan A-21.
One of the significant changes that was included in Regulatory j
Guide (RG) 1.89 (which was written to endorse IEEE 323-74) as compared to the cr;teria previously established by IEEE 323-71 was the " qualified life' requirement for safety-related equipment.
The NRC's Regulatory l
Requirements Review Coealttee recommended.that the RG be applied only to j
l future construction permit applications (i.e., it should not be backfitted).
This decision was based on the consideration that the i
l Incremental improvements were not significant to safety and that full i
implementation of IEEE 323-74 required the further development of other e
l ancillary standards to provide guidance on specific safety-related equipment and components.
Subsequent public comments and review by the ACRS did not alter this position.
Note:
It appears that IEEE 323-74 was imposed on the industry l
before it was fully developed.
It was recognized that additional guidance was needed in the area of
+
accelerated aging techniques used to establish a qualified life for electrical equipment and assemblies. Task Action Plan A-24 and the NRC i
research program wire intended to provide guidance for the development t
of test methods and licensing review procedures on aging.
Issue:
Determination of a " qualified life' may not be possible given the current state of the art.
The considerations included in IEEE 323-74 pertaining to margin considerations (the difference between the most severe specified service d
condition of the plant and the conditions used in the type tests or l.-12
analysis) and sequence of testing (the most severe sequence was to be used during type testing) was cited for providing additional assurance that the type test of a single unit or analysis with justification was adequate.,
IEEE 323-74 required that equipment qualification be demonstrated by type testing, operating experience, or by analyses based on appropriate test data.
These quallfication methods may be used individually or in any combination depending on the particular application.
The source term required by RG 1.89 represents core degradation substantially beyond that which emergency core cooling systems were designed to prevent. Also, other assumptions, such as uniform distribution and exposure calculations at the center of the containment with no shielding, represented a significant conservatism.
Issue:
The degree of conservatism built into source term requirements may be excessive.
It was not likely that equipment required to function only during the firsthour(gliowingtheLOCAwouldbeexposedtoradiationdosesin excess of 10 rads and therefore, the safety function of this equipment wouldnotbeimpairedsincemostelpstomersbegintoexhibitdamagewhen exposed to doses on the order of.10 rads or higher (minor damage is initiated for certain materials at somewhat lower doses but not to the extent of functional impairment, and radiation damage to metals is not a concern at tnese levels).
Note:
This information may be useful in focusing EQ program requirements.
Equipment needed for long-term pccidegt mitigation, qualification to radiation doses ranging from 10 to 10 rads, depending on plant size, was acceptable.
Historically, beta radiation had not been explicitly considered in' environmental qualification due to the inherent protection much of the equipment has to beta radiation.
The potential incremental effect of l
beta radiation on organic materials would be considered during the long term SEP evaluation.
I EPRI NP-lS$8. "A Review of Eouloment Aoina Theory and Technoloav." Seotember lHD i
l The Franklin Research Center, under contract to the Electric Power Research Institute (EPRI), performed a comprehensive review of the theory and technology of equipment aging, especially as they relate to environmental qualification of safety-related equipment at nuclear power plants.
The report sumary stated:
4
...The dominant picture that results from the study is that there 1
is no comprehensive, scientifically rigorous solution to the l
problem of accelerating the aging of equipment. Aging that can be l
l L-13 l
l-
I i
accelerated in trays that yield verifiable correlation between real and j
simulated aging is an exception rather than the rule....the documentation provided in this report should facilitate acknowledgement and acceptance of current technical limitations.
This will in turn Sesh thee recenctuuttwa et Witkauem fogstegneau artth paract6cM means of meeting them.
Major advances in equipment aging technology are not expected within the foreseeable future. The huge diversity in the kinds of materials, components (combinations of materials), and assemblies (combinations of components) makes the dimensions of the probles quite large.
An enormous effort is needed simply to discover, accumulate, and catalog the appropriate physical constants for even a respectable percentage of these items, New materials, modifications to old materials, how components, and new combinations of old and new materials and components make this a never-ending undertaking.
l Although equipment aging on a rigorous scientific basis is beyond the current state of technology, it is nonetheless possible to I
satisfy the purpose of aging in equipment qualification.
This is true only so long as the intent of aging is to assess i
l qualitatively the vulnerability of equipment with respect to aging effects and not to achieve aging in the strict sense...."
l The report summary also concluded that synergistic effects do not appear to result in degradation that is very different (physically) from what is seen when aging influences are applied sequentially (though the rate of degradation may be different), and suggested that equipment surveillances that are keyed to the known degradation / failure mechanisms might provide an acceptable alternative to accelerated aging. Given the conclusions that were reached as a result of this comprehensive study, EPRI stated the following perspective at the beginning of the report:
Proiect Descriotion i
...Since the issuance of this standard (IEEE 323-74), there has been debate as to the technical feasibility of conducting I
accelerated aging that can be correlated with aging in real time.
This project was intended to steer the debate toward the development of useful and realistic qualification procedures that are consistent with the state of the art."
Preiset Obiective "The objective of the project was to investigate accelerated-aging technologies in sufficient detail to discriminate between what can be accomplished in a technically supportable manner from that which cannot...."
Prolect Results "The main conclusion of this work is that, with the possible exception of certain simple materials, the prediction of a L-14 Ik J
quellf f ed life through accelerated aging is generally not feasible at the present time. This conclusion must be interpreted very carefullyt l
It is not the usefulness of aging in qualification that is challenged; I
rather, it is the notion of demonstrated qualified life. Accalainted t#afag (as distinguished from steelerated attag) is useful for the identification of dominant' failure modes in advance of field applications and for comparative analysis of alternative materials and component designs. However, aside from the limited number of materials for which aging effects art well characterized, the demonstration of qualified life 'hrough accelerated aging is generally not feasible.
(This conclusion is supported by a fairly substantial military sffort in the 1960s to assess the usefulness of accelerated aging as a basis for i
predicting long-term equipment performance.)
The use of age conditioning should be encouraged with the objective of determining wiether equipment is minersMe to aging stresses....*
Note:
Once again, the ability te establish a
- qualified life' for equipment is questioned.
FRC Lecture L-A5300/EPRt.
- Review of Aoino Theory and Technoloov."
Scotember 17. 1980 The subject lecture wms prepared by S. P. Carfagno for presentation at the EPRl/NRC Equipment Qualification.Information Exchange Meeting.
The lecture notes contained background information on research that was conducted by the Franklin Research Center (FRC) under contract to the Electric Power Research Institute (EPRI).
The results.of this research activity were documented in EPRI NP-1558 (discussed above)4 The FRC lecture notes illustrated the difficulties associated with accelerated aging by reviewing some of the uncertainties that exist with the accelerated aging of electrical cables. The following problems were cited:
the Arrhenius model may not be applicable;
+
extrapolation of Arrhenius data may yield erroneous thermal aging parameters; the rate of circulation and the rate of fresh air input in the aging oven are significant parameters, but their influence on the degree of acceleration are usually not known quantitatively; oxygen diffusion that takes place in conjunction with background a
radiation over a long period may not be adequately shulated in the I
short time taken to simulate radiation exposure; l'ng-term degradation that may take place at the interfact between the o
a conductor and the insulation is probably not simulated; l
since we know very little about the aging influence of moisture, we can not be certain that this is adequately accounted for; l
L-15 1
l l
4 in. in. mai er.eieni inai esi.. aere.. ine inauisii.n waen in. ean. 4.
energized may influence aging; the electrical stress due to energizing the cable (and from occasional a
electrical surges) may have an aging effect; degradation may be most significant at points where the cables make e
contact with supporting structures; as the cable ages and the insulation tends to become embrittled and lose e
its flexibility, degradation may be accelerated by vibrations and other forms of mechanical disturbances (e.g., motion caused by electrical surges or by vibration of cable support structures);
if the cables are coated with a fire protection material, aging may
=
differ from that of an uncoated cable; and Issue:
EQ relative to fire retardant coating (and other fire
~
protection features) may not be adequate.
the aging of cable terminations may be more important than the agi'g of n
the cable itself.
Given the inherent weaknesses and uncertainties that existed in the qualification methodology, a " practical solution" for equipment qualification was suggested which included:
(a) exclude aging mechanisms that are I
considered to be insignificant (e.g., the aging mechanism does not adversely i
affect the ability of the safety system to perform its function); (b) use the l
Arrhenius model with awareness of its limitations; (c) restrict aging to weak l
links; (d) use survalliance to obtain advance warning of significant degradation; and (e) use reliability data (e.g., for equipment in controlled environments in cases where the effect of a DBE is amenable to analysis).
The lecture notes concluded in part that:
'The status of aging technology is not adequate for the specification of scientifically rigorous accelerated aging procedures for simple devices, let alone for complex eculpment.
However, it is possible to devise practical acceleratec aging l
procedures to degrade the specimens before they are put through a l
qualification test.
The correlation of aging with real time to estimate a qualified life is difficult at best.. Considerabia engineering judgement must be used in making this estimate."
Issue:
Given the state of the art that was in existence at the time IEEE 323-74 was developed and the limitations discussed above, it would seems that EQ program requirements may have been misdirected (especially with regard to the required determination of a
' qualified life' and the absence of surveillance requirements for obtaining advance warning of significant degradation).
L-16 l
l
i l
Letter from R. Eckart IAtomic Industrial Farunt to H. Denton nrovidino fication of electrical i
l Industry cosition caoers on environmental cua ecutement. Januarv 4. 1982.
j Forwards industry position papers ont a.
one-hour minimum operating time margin requirement and i
b.
pre-aging cur.:orns for seismic palification.
The industry position was that the arbitrary requirement of a ens 4 7
operating time was unnecessarily conservative.
Instead time re g..dments shouldbeestabitshedbyaddinga10% margin (asestabilshedinIEEE323-74) to the time required based on system dssign and accident scenario analysis, l
Issue:
The one hour operating time requirement continues to be a controversial issue and should be revisited.
I Letter from R. Eckert (Atomic Industrial Forum) to W. Direks orovidino an i
industry notition nacar on environmenta) nualification of alactrical soutomant.
'abruary 1, leal.
l The position paper argues against the proposed NRC program for accreditation i
of test factittles that perform qualtf tcation testing of safety-related equipment.
The industry endorsed the use of cvality assurance and control procedures developed and implemented to accortance with 10 CFR 50 Appendix C for assuring high quality 16st facility services.
ggelear Industry Position Paner en Surveillance and Maintenance Proorams.
l oresented at a workshoo soonsorac by NSAC. Aoril 23. 1982.
I 1
The industry's position on maintenance and surveillance program requirements i
in support of environmental qualifiution was provided. The position was tak.en that existing programs (i.e., corrective atintenance; preventive maintenance; periodic performance testing; surveillance; manufacturer,
,.ttitty and NRC comunicationst and operational and performance parameters menitoring and evaluation 1 were sufficient to ensure that the functional capabilities of safety-rehated equipment are maintained.
Note:
This position deos not appear to be valid, given a historical perspective. I.xisting programs have not been well directed at identifying, coassunicating and correcting EQ problems and deficiencies.
Letter from R. [Jert (Atomic Industrial Foruni te R. Vo11% JfavidLno industry oosit a oneers on environmental cuaLification of electrica t touloment.. Auo.fst 24. 1982.
Forwards ind.astry position papers on systems operating times and on radiatton
- onsiderations.
For equipment qualification purposes, the industry proposed that equipment be placed in one of four categories:
t) short term for equipment operating durations of up to 20 minutes; b) intermediate term for equipment operating durations of up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; c) long term for equipment l.-17 m 2.
i s
1 i
l operating durations of up to 15 dayattand d) extended term for equipment operation of up to 1 year.
O i
a the The industry postston on radiation considerations was thattinstantaneous release as l
i, unrealistic and overly conservativet and b) diation dose level.gnize and the NRC should reco endorse the concept of a threshold damage ra Letter from R. Eckert (Atomic Industrial Forum to R. Vol' aer orovidina fication oJ electrical l
industry coittien cacers on environmental cuat ggu_tement. Becember 28. Ital.
1 Forwards industry position papers on aging evaluation methods and on margin considerations. The position is stated that existing state-of-the-art can not i
support a determination of qualified life for complex electrical equipment, j
and that licensees should only be required to implement NUREr.-0588 Category !!
i requirements, The position paper referenced studies done by Sandia and EPRI in support of this position.
With regard to margin, it was the industry's position that the application of margin on top of margin during type testing is excessively conservative and not warranted.
i Issue:
Tha use of ' margin on top of Nrgin' continues to be controversial and should be justified..
Letter from S. Gescner (Gese'Ischaft fur Reaktorsteherheit - GRS) to i
S. Accarwal recard' no the 1903 Symoonium on Nuclear Power iv h beto ber 5. 1981.
b j
The purpose of the letter was to forward a technical paper, " German Philosophy l
and Practice of Aging within Qualification of Electrical Eculpment for Safety Systems of Nuclear Power Plants,' which was to be presentet at the !!83 Symposium on Nuclear Power Systems by S. Gossner. The paper out11aed the approach used by the Gemans for establishing and maintaining equipment qualification. [ssentially, the position was taken that accelerated aging of l
complex electrical equipment should not be done at temperatures that are very 1
dif ferent from those experienced during nomal operation so as to avoid increased uncertainty.
This lead to the conclusion that it was generally not possible to prove that the qualified life of electrical equipment will extend over the design life of the nuclear power plant by use of accelert.ted aging because the duration of the accelerated aging would tan too long.
In Germany, it was enough to establish that the expected life time of electrical equipment would be "lony' !!.e., extend over a number of y)ars), and then prove continuously by se J monitoring, surveillance and maintenance that the end of life has not bo o n W ed and that a suffletent expectation of remaining Ilfe is mainwesi io support additional plant operation.
Previous experience with specific my ;pment applications was used as the primary basis for making the determinatit,n that the expected lifetime of electrical equipment would be "long ' but in certain insto ces type testing procedures were used to ma k this determination. hwever, it was noted that qualification or electrical equipment for accident conditions was not yet addressed by the German rules 3r standards.
L-18 1
i
l Note:
The terman approach seems to be more in keeping with the state of the art.
MAC-EB. "A Gui da to Qualificatian of fiactrical fouinment for Nuclear Powar F ant 1.5 lantamsar llR3 The guide provided a baste overview of the equipment qualtf tcation process and j
program implementation.
fouinmann Maintenance and Qualif tentien Newtiettar. October 1923 (Innued January..984).
j l
A sumary wan provided of a paper that was presented at the Tenth Annual y Division Conference American Society for Quality Control Hattonal Ene San D ego during September $8-21.1983. ' Overview of Environmental held it:
Qualtftcation of Electric Equipment important to Safety at Nuclear Power Plants.' The paper, which was authored by A Mascisntonio R. LaGrange and V.NoonanoftieU.S.NuclearRegulatoryCommission,iden[ifiedthefollowing-l shortcomings:
The number of test samples used in type tests are not statistically significant to assure that the results are representative of the equipment population as a whole.
The margin applied during tests is very small in some cases.
Aging t.chnoing, was not deveio>ed to the point where the entreme extrapolations used to estabitsi long qualified life can be accepted at face value.
A survalliance program must be estabitshed to verify that uninown age degradation is not taking place.
The interpretation of surveillance test findings was subjective.
if degradation is taking place, at what point does corrective action need to be taken to assure that the equipment can still perform its function j
during an accident?
Note:
This informatten does not appear to be reflected in the current NRC EQ prngran requirementa, 1
NUREG/CR-4144. "Imoortance of Ranki ngh _ n Aoino Considorations of l
Comoonents included in Probabilitt< c Atak Annonaments.* Anr' )
986 3
Extstingprobablitsticriskassessment(PRA)studieswereexaminedtogain insights about the relationship between aging of nuclear power plant components and pubite risk.
T.ie potential risk significance of aging was evaluated by calculating the increase in core melt frequency for assumed increases in component failure rates (the logic being that aging results in increa' sed component failure rates). The components that cause the largest change in core melt frequency were most sensitive to aging effects. The results of the analysts indicated that the most risk significant components at a plant depend on a number of factors including plant system design, testing, maintenance intervals, and operating procedures.
Based on the PRAs that were studied (0 cones, Calvert Citffs, and Grand Gulf), many of the most risk L-19
' kW g
significant components are those in the auxiliary feedwater system, the reactor protection system, and the service water system.
Specific component types included pumps, check valves, motor operated valves, circuit breakersAssump And actuating circuits.
be recognized included:
(a) tse time dependent behavior of the failure rate wasbeyondthescopeofthestudycyb)noassumptionsweremadewithregartito the PRA sample which components are most suscept'b.e to ag*ng processes. (c)he actual scope the results of the studp were limited by t was small. and (d)hree PM: that were stedled.
l and deptt of the t j
Note:
Perhaps there is a place for PRA in focusing EQ requirements.
NUREG/CR-4301. *$tatus Reoort on feutnment Dualif tention innuen Ranmarch and ketelution.* tentamber 1918 1
i The Qualification Testing Evaluation (QTE) program was begun in 1975 to i
address equipment qualification issues and to identify and investigate any new issues. Morespectftcally,theobjectivesoftheQTEwereto:
(a) obtain i
data needed for the confirmation of the suitability of current $tandards and i
i Regulatory Qutdes for Class I oculpmentt (b) obtain data that will provide improved technical bases for mocifications of these Standards and Guides where
)
appropriate; (c) establish data-based and standardized test methodologies for j
equipment qualification programs; and (d) support the NRC licensing process with Qualification expertise and test capabilities.
Results of the QTE were incorporated into Regulatory Guide 1.8g, ' Environmental Qualification of for Nuclear Power Plants,' IEEE 323
[lectric Eculpment important to Safetf Equipment for Nuclear Power Generating i
j
- lEEE 5tanc ard for Qualifying Class i Stattens,' 10 CTR 50.49, "Envi o nmental Qualification of Electric Equipment for Nuclear Power Plants,' and other Standards and Regulatory Guides.
This NUREG/CR sumartzed and documented what had been learned as of 1985 Lelose to one hundred reports and papers had been written? ional research that wasdiscussed the re implicationsoftheresults,andidentifiedaddd needed to resolve any remaining questiols.
In total, thirty separate issues 1
were discussed encomaassing three generic areas: accident simulation methods;
{
agir.g simulation motiods; and special topics.
The report provided a l
description and sumary of research efforts applicable to each issue, and a summary of the findings to date, ' Table L-4 is a listing of the issues for each of thr,se generic areas.
The following insights and observations were eXtracteo from the information contained in NUREG/CR-4301:
l l
While there are some exceptions, synergistic effects during accident l
simulation do not exist for most materials.
Electric cables are the single equipment item that has been studied i
extensively.
Protected or enclosed equipment does not respond quickly to transtant environments, and short-term thermal transients may not need to be simulated.
l It may not be necessary to simulate superheat conditions since saturated conditions appear to be more limiting.
4
- c TAILt L-4 Listing of Qualif tsatten Testing tysluttien Progns !stues A.
Accident Simulation Methods:
simultaneous / sequential exposure luperhetted/64turated lttam effetti thermal shock and steam impingement dose-rate effects beta / gamma radiation effects oxygen effects chemical spray ef fects acceleration of post accident environments sensitivity of accident simitlations to agtag methods hydrogen burn influence on accident simulation methods submergence simulation B,
Methods for simulattu Antne candittenal realtiLic ambient environments?
Ilmitations of the Arrhenius method dose-rate effects eimultaneous/ sequential exposures mechanical stress effects oxygen offsets humidity effects analytical and experimental techniques for correlating natural and j
4 artiftstal aging t
Comparison of artificially aced and natur411y sced scutoment 2
C.
Soecial Tootes:
complimentary test approaches and considerations TMI I expertences evaluation of qualification procedures for specific equipment types Regulatory Guide 1.97 requirements advanced systems qualifIcatton istues l
realistic accident environments and calculational models criteria for selecting almulation methods review of Standards and Guides battery aging methods r Jiation damage thresholds I
s l
L Il l
Fabrication processes, surface preparations and installation procHures can impact equipment qualtf tcation and should be considered.
t features f askets e rings cable gromets, etc.) are Comkonentsealinkdorations ?!he mosl comon finding at THI t was crt ical (Q cons moisture intrusion into son'ed housings).
Note This is ' good to know' information.
Maintenance and survet11ance activittaa can very easily invalidate equipment qualtftcation espetta))y with regard to stulpment seals and embrittledelectriccabies.
The self heating effect' taused by iontatng radiation must be a
considered when testing temstrature senshttve s 9tpment. restrictio
" hts may It may be prudent to adopt a untform source signatur as a replacement for nucilde fractionation specifications to asture use of an identical generic data base by Itcensees.
In an accident that is successfully terminated i.e., does not result in significant core degradation). the gap release w(ould be the only fission product source.
Cffects due to beta and gama are essentially identical.
The presence of oxygen can influence material degradation in either a positive or negative fashion, depending on the specific material compositten.
Radiation followed by thermal aging is typically more severe than thermal aging followed by radiation.
i Hydrogen burns in containment may be significant, but the hydroqen burn environment is complex and the survival of safety systems canno", be treated simply.
Further, it' is nearly impossible to accurately define a j
hydrogen burn environment." Global burns can produce gas temperatures as high as 1300'K (1880'F) and pressures up to 4)0 kPa (58 pst). Methods for demonstrating equipment survival in hydrogen burns have not been estabitshed.
A great deal of uncertainty currently exists in the state of the art
(
procondItioning and :imulation techniques.
I Humidity effects for polymeric materials are probably not of major e
importance in radiation environments as long as materials known to have humidity resistance in thermal environments are isletted.
Synergistic effects do not appear to be significant in LOCA type tests I
of electrical cables.
Generic [PR response does not occur, the actual EPR femuintion ti important to equipment qualtftcation.
- l. It
0
'y^
Ellatnation of in containment terminal bletka in safety-related circuits should be conaldered.
Issue:
To what extent is this being pursued?
Very little (Q research has been conducted on pressure switches, RfDs, pressure transmittsrs and valve operators.
Issue To what ettent is this eensidered to be a problem?
Research is only just beginning to assess the adequacy of EQ for RG 1.g7 a
funcitons.
The areas of primary concern include reactor coolant level Instrumentation, core exit thermecouples, containment area radiation monitors, halogen and particulate sampilng, and coolant activity measurements.
Issuoi What is the current status?
Testing materials of component construction can help to estabitsh equipment qualtftcation, but material use and configuration in the end productcaneffectqualification(i.e., table.vs.sacetconfiguration) and must be considered.
Note:
This given support w the netten that analysts een be eredited Ln the tusM f'satten preessa as long as the 11mitations are recognised.
Radiation damage thresholds for various types of materials and equipment can be estabitshed based on research that its been conducted.
Notet WhileNURESlCR-4301.frevidedanexcellentsunnersofresearchthatha c.,ieted i rou.he way,of equLpeent' qutilficatten given the existing state of hi i
is di,i nei revide.ueh 4eeul en whai een reasonabiy be expected in t l
l the art, and specif te findings and reconsundations did not seem to be very well focused in this respect.
It should be possible to reach some meaningful i
conclusions that bear en IQ progrannatic requirements with all of the research that has been conducted, and perhaps this is an indication that IQ research has not been well focused and managed by the NRC.
Currently, it seems that research has not established a sound technical Ipasts for the existing I
quaitfication process and requirements.
EPRI EL /NP/CS-5914-SR. "Proenedinot t 1993 EPRI Waritthan on Power Plant Cable Conditton Monitor' no.' July sea The Electric Power Research Institute ([PRI) sponsored a workshop on power plant cable condition monitoring,kshop was attended by 103 engineers and which was held in San Francisco, California on February 16-18, 1988.
The wor researchers from the United States England, Canada Japan, and Sweden to share knowledge and experiences related to cable fallure mechanisms and currently available testing methods for assessing cable condition, and to discuss needed research.
(pRl programs for both fesell and nucliar plants I. 23
- -nnn_.
y e
.a v u,
.9 : f !;
M uf :,
A*'
f were discussed, and il technica) papers were presented.
The following conclusions are quoted from the proceedingst No serious cable problems currently entst. Fositi and nuclear plant cables abould provide continued reliable perfernance throughout the normal Lifettme of the plant, Conventional monitoring techniques traditionally used for cable troubleshooting (i.e., in s'tu high pot, power 4ctor, insulation properties), as well as lah ratory testing of electrical pnd mecha alng the gradual degradation of resistance aging cables.
The benefits of extending and maxim ting plant and cable system lifetimes are an incentive for developing improved monitoring techniques.
Improved techniques for monh,oring the condition of fossil and nuclear balance of= plant enW elneedonly.shtrasterletheability of the cab)es to provide re hable tarvice setnts'tahly n a normal operating environment, improved Itfe assessment techniques could guide refurbissment programs and avoid the high costs of unplanned outages.
As the plant ages, increasingly detailed assessments could be made.
Im> roved monitoring techniques for inside containment nuclear plant caales must characterite the ability of aging cables to perform safety-related functions in harsh accident environments.
Design and testing to industry standards provide assurance for the usual 40-year qualified Itfe of nuclear grade cablet that are used in harsh environments however, improved monitoring techntgues may be needed to assure qualification beyond the normal 40 year Itcensed term of nuclear plants.
Monitortnq techniques should allow for in situ testing without 1ticonnec',ing and damafitng the cables.
The techniques should also apply so the large amounts so low voltage, unthleided power and control calles in nuclear and fossil fueled generating stations. An tonited gas surrounding the test ~ cable can provide a low-tmpedance path to ground.
Monitoring should include cable terstnations.
In addition to meeting life'entension needsi an insreved in situ monitoringtechniquecouldbeusedfortroubleshooting(faultdetection) and for acceptance testing of cable suspected of bein defective.
for exsaple, personne) at the TVA lequoyah lant suspected a Recently, been damaged during installation, and in-si u hi-pot testing cable had proved to be an unreltable indicator of entle condttion.
With further development, advanced diaenestic methods such as time =
domain reflectometry, time domain spectrometry, and insulation
'ndentation could prove to be useful, it is unlikely that one all= encompassing test method or procedure could e
measure the global condition of insulat on material and also pinpoint a local defect or degradation in insulation.
Plant managers must also weigh the costs and benefits of applying a technique against rewiring
- costs, l.-14
~
i s
1 Collecting, maintaining, and ustne informatie3 en cable installation, environment, performance, and to { results would generate a useful database to support cable life-extension decisions.
The following recomendettens were made for future short tem research projects:
Compilattenofutilitypraetteelandespertensewithepble(failure expertence trouble table typtai types it, use, and repistement totta) =
highpriorIty.
Utility applicatten of time demain reflettemetry = high to medium priertly.
Utility appitcation of time domain opettrometry - high to medium priority.
Qutdelines for in situ acceptance test of cable (including statistical sampilng approaches) - high to medium priority.
Correlatten of Jacket tenditten to insulatten sentitlen = medlMm
- priority, ReviewofAC/DChl=petles",technelety(bonqfits, damage)a limits, etut nt availability. AC/DC equivalence, testing Lneused med um priority.
Vulnerability to installation and handling (cables, splices, terminations) low to medium priority.
nothodi of environmentii asioiiment (ohnie espacity..rreietten, ambient temperature radiation levels, surge voltage survey, tell tale tags)-notprierttIsod.
Techniques to locate hetYspets a not prierttiged.
Quidelinesforutilityactivities(baseline,measureenvironment, archiving of samples, ev419ttien of fat)ures) not prierttired.
The followIng recomendettent were made for future long term research projectsi Improved cable technolo materiall, shleiding) =gy for new plants (teltability, layout, e
high priority.
Techniques for providing ground plane for unshielded cable high priority.
Definition of the end of life for cable (damage modes, acceptance criteria, failure propagation) high prhority.
Ietter models for radiatten and thermal aging of polymerit materiall
- medium priority.
I.= tl
s g
s
.c Apn1tuattenofartiftsklitntelittsnes-techniquestotablesystems1ow l
prier'ty.
l Evaluation of instrument table ' low priority.
Reivvenation techniques for damsged esble 19W priority.
Correlatten of different aging methodelegies by different manufseturars i
l not prierttlaed.
Thermal fingerprinting of cable not priorittled.
The following information 18 a summary of some of the more notable weirkshop presentationsi A,
- 0varvlaw af fen' Aettw'.itaa in Nuelaar slant and f ahla Life tatsmalan.t
)
l Banend F l tar a 'the t netrie pawar Raamareh nalltuta l
This paper provided backereund information en the issue of cable life I
extension. Although 'EPRI studies as early as lift...eencluded that-there are no techn, cal barriers to operating nuclear pl6nts beyond the
' cable issues that were more or-less arbitrary 40 year itsensed tom,d uncertainttes 16 identified relativo tot 1' fe ex1,enslen int' ude environments, asse era",et vs. 'n. plant as ne cenpervat,taa of sosign correlatten of matertaViseradatten and lipilt'gal nortering WLth L.kA perfemance, and unavailabilhty of ertelna, table stettment er documentation, tutsting table research that was cited in this I
presentation included cable indenter aging montter (franklin?. i,). plant in-studyofnaturalvs.trVfttialaging(UniversityofConnectitu tontted gas shield feat ht11ty (t I), and,antipxident study (UnLysrsity of Virginia).
(PRI research that was in the planning phase sneluded condition monitoring of local degradation !$andla
'ncluding activation assessment of industry practice to assure cable enercy mespurements nervteability(tC0iCH),lamagestudy(TBO),andcableItfeextension thema) fingerprinting (University of Tennessee) insta11attenf utt11tyguldeltnas(fl0),
8,
'Catta Candillan Manitarina in a pressurtrad Water Ranctor f avironment.',
i T.t. Al Wuaantni af Duhn sawar famaanv An on going cable conditten monitoring pretram that was initiated at Oconee by the Duke Power Company in order to verify the condition of reactor butiding table over the 11fe of the slant was described.
At the time when cables won insta11ed at osense 11H51970 enytrennental cuallfication requirements were nonexistene, and no na) lural cable aging
-cata w.re available. Verleus table snares wero installed with the intention of removing ttble samples at 5 year < ntervals for examination and testine.
Test results for the first e and 10 year intervals indicated that tables are aging in the resetpr environment "as-expectedi ' spettfit dettlls of the test Peluits were proV dedi.
'cahta canditian Nantin' ine prearam at Parro NuelaLP pnwar D' ant.'
C, e
- 5. Ennny Matturt af MOS and Ituart Litchfia' d of flave' and F actric
@minat' na Canaany l
L t4 l
b
l N
A description was provided of the cable condition monitoring progras that was comitted to for the Perry nu tr power plant during plant Itcensing.
Representative cable sampi tre installed as spares in various locations for the purpose of ch aoing monitoring and evaluation at 5-year intervalsi spectfic details of the cable monitoring progras j
were discussed.
The first set of cable samples were scheduled to be i
examined sometime in 1993, i
1 0.
" Physical Deoradation Assessment of cenerator Station Cabl,ea." D. L
)
Stonkus of Ontario Hydro 4
l The results of cable degradalton assessments and plans for future i
rermarch by Ontario Hydro were summarized.
Preliminary studles indicated that the low voltage PVC insulated cables were in rielatively good condition after nearly 15 years of service. However, cables -
l Insulated with styrene butadiene rubber were embrittled.'.
e i
...,p.
- y.. <,.
i]
Techniques being considered for; evaluating uble degradation included: **
oxidationinductiontime(0!T).thermogravinetricanalysts(TGA),
I attenuation total reflectance spectrometry (ATA?. and density 6nd i
modulus profiling.
Results obtained using OIT undicated that physical j
preparties of cables can be misleading in predicting the remaining life of cable insulation, a
l Long-term reliability assurance and plant Ilfe extension studies were j
being actively pursued by ontario Hydro and the oldest operating Canadian nuclear plant, Pickering NGS "A" (operating since the early 1970s), was among the plants being studied. Also, the decommiissioning j
of two small nuclear plants which were placed in operation in 1962 and 1967 wou'd provide an opportunity for further evalu Q on of field aged
)
4 I
- cables, f
f,.
" Cable and Electrical Anoaratus Monitorino Procram at San Snofre Nuclear Generatina $ tat' en ($0NGS) Unit 1.* R. a 5t. Onae oJ tout sern California (dison Comaany Southern Callfornia Edise,n Company (SCE) had estabitshed a program for monitnring certain characterisites of cable and electric equipment.
SCE i
used the computer based ECAD. System 1000'(Electronte Charactertration' and Diagnostic System) for gathering, storing and trending the information.
SCE concluded that the systematic, accurats collection of electrical characteristics is the first stes in the ability to predict l
and trend cabis and apparatus integrity.
Tie analysik of this data i
allows for assessment of cable insuhtien resistance probleets, equipment failures, and degrading conditions of an electrical nature.
For j
example, moisture problems are manifested as a change in capacitance, and the location of the moistura can be deterstned by time domain reflectometry (all within the capabilities of the ECAD System 1000).
l i
F.
" Predictive Life Measurements of Naval Aircraft Wirinn
- A. M. Bruning cf Ligfromechanical Deston Comoany The study of premature failure of poljamide insulated aircraft wiring (Kapton) performed by the Naval Research Laboratary was summarized.
The l
l l
.L-27 g
y
(
probles tras evaluated on a molecular level and a syllogism was constructed of '.he deterioration scenario for polyamide. Other insulation materials are expected to have different deterioration scenarios that need to be determined (it is thought that the testing of field wiring will be more reliable if it is based upon insight into the theory of the agtnp process?,^ Physical phenomena that was cited for diagnosinginsulatLondeterioration'includedgasdetectionlcwavest electromagnetic emission - visible light, infrared,' or rad l
audio noiset electrochemical indication of leakage currentst co.ona i
voltage wave form; change in 1ocal distributed constant electrical.
~~
]
properties detectable by reflectometry. power factor, etc.
l G.
"A Review of Candidate Methods for benactino Incinisat Defects due to j
imino of installe: cables in Nueleur lower Pl antt.
- remnesit D.
j Tartzloff of the i ntional Rureau ao12andarna A sumary of research that was underway at the National Bureau of Standards (NBS) to evaluate in-situ methods of detet. ting cable
)
insulation degradation was presented.
The NB$ presentation indicated that a very impartant concept for in-situ insulation evaluation by i
nondestructive test methods is that trends in test results yield much more informattrn than absolute values obtained from single tests.
l Spectfic test methods that were discussed included time-domain reflectometry (TDR), low-energy pulse testing (i.EPT), partial discharge detection, mechanical properties Le.g., tenslle strength, elongation, stress / strain characteristics). dhelectric tests,le technique for and time-dasain spectrometry (705). While a universally acceptab i
l detecting degraded insulation has not been established, the NBS thought' that TOR, i.tPT. partial discharge detection, 708, and stress / strain
)
measurement (indenter) offered some interesting possibilities.
In i
particular, the Nps thought that TOS could.poteatially provide a very i
powerful detection method for the overall' aging of cable insulation if a correlation could be established between TDI measurements and insulation l
integrity.
,,7 l
Note:
It seems thht s lot of effort has gone inte developing techniques i
for monitoring yingt ih,rhaps this information could be useful in l
establishing EQ surveillance requirements.
NUREG/CR-5313. "Eautoment Qualification (E0) - Risk Scenino Study.'
l January 1989 The EQ Risk Scoping Study used probabilistic risk annessment (PRA) techniques to assess the insact of EQ on reactor risk. A list of candidate harsh environment, rist significant equipment was developed; the list was narrowed by screening out equipment whose operational reliabilities would most likely not be <ffected by the %rsh environment; and the equipment remaining on the list was then examined in more detail.
The study included a review of several FRA and safety ranking studies to determine what insights could be gained relative to equipment performance in elevated environmental conditions.
Although many obstacles to straightforward interpretation of EQ risk significance via PRA methods were identified by the study, several general conclusions and recommendations were made based on PRA insights, including:
+
(a) EQ issues associated with'long' term accident equipment operability are not risk significant and equipment qualtftcation should focus on ensuring equipment operability for the first few days of the accident exposure; (b) demonstratingequipmentoperabilityduringtemperaturel lit /duringradiation pressure, and steam conditions la more important than demonstrating operab conditions, and further Investigation of whether an instan",aneous release sourcr term adversely impacts risk was warranted; (c) harsh environment I
equipment reliabilities could increase core damage frequency estimates for PWR transient-induced and small-break LOCA sequences and for BWR TW sequences; and (d) the worth of plant status instrumentation and currently installed accident manag uent equipment was difficult to measure using current pRA techniques, i
and PRA techetques to assess the risk importance of plant status equipment and currently installed accident management equipment should be exploreo, Also, asaresultofthisstudyl(thefollowingconclusionsregardingtherisk sig.lficance of historica Q 1: sues (as discussed in MtEQ/CR-4301) were stated:
PRAs calculate that equipment function only has high risk significance
+
1f the equipment operation occurs during the first few days after accident initiation. Given this time frame, the issue of whether correct accident acceleration techniques have been used is not risk significant. Other EQ issues that'are not risk significant on this basis as well include the importance of performing simultaneous versus sequential accident simulations of radiation, steam, and chemical spray environments; the importance of including crygen within the qualification test chamber during accident simulations; and the l
importance of appropriately simulating the beta radiation dose.
i Note:
If substantiated, this information could drastically influence the restructuring of EQ prograa requirements.
)
PRAs do not calculate substantial in-containment radiation Conditions 1
until core melt has occurred.
Based on pRA calculations, core melt i
rarely starts within the first half-hour of an accident sequence and may not occur until several hours after accident initiation.
PRAs also l
Indicate that risk significant containment failures will occur within a few hours to a few days. Therefore, from a PRA perspective, much of the safety-related e: 1pment inside containment need not function in an acc1dont radiat1on environment and the remaining equipment needs to
' unction during radiation conditions only for a few hours to a few days.
Additional issues impacted by this conclusion include the importance of simultaneous versus sequential accident simulation techniques: the importance of post-accident dose rate effects; and the significance of beta versus gamma effectiveness in producing equipment damage.
Note:
If this information is valid, perhaps it could be used as a basis for relaxing to some extent scae of the EQ requirements that currently exist.
Two sets of sequences where degraded harsh environment equipment
+
reliabilities may impact risk perceptions were identified by the study.
In the case of a PWR experienring a transient-induced or small break LOCA, both the high pressure :njection and auxiliary feedwater responu L-29
and the feed and bleed operation (using high pressure injection and pilot-operated reitef valves) were itsted.
For a SWR, the trentient initiated sequsnce (commonly called the TW sequence) where there is a loss of suppression pool cooling was listed. These sequences typicaily involve salenoid operators, motor operators and safety relief valves.
The study noted that in both the PWR and the BWR sequences, equipment reliability data for steam, temperature, and pressure concittons, but not for radiation accident conditions, was desired.
Several accident scenarios not historically part of the design basis are risk significant (e.g., station blackout for Gnnd Gulf) operation of equipment-that is not included on IQ macter 1<sta is important for some sequences (e.g., feed and bleed'uset in containo6nt solenoid and motor.
i operators that historically were not tutilfted and also lacked redundancy)! and PRA assume equipment functlpnt) tty for environnants that exceed the design basis (e.g., safety-relief valves in BWRs are assumed available to function after drywell rupture even thoujih the environments that would cause a.drywell to rupture substantia ly exceeds the design basis).
Issue:
Based on PRA studies, should additional equipment be added to the EQ Master List?
The EQ process does not provide coverage for all environments that PRAs calculate as risk significant. Consequently, further research and/or testing was recommended to:
(a) assess the reliability of BWR high pressure core s pray lHPC$) pums: to handle high temperature working fluids (appitca)1e to rtation plackcut sequence propabilities); (b) assess the reltabilities of safety relief valve $RV solenoid operators. main steam isolation valve {MllV) sole (noid) operators, and H51V bypass valve actor operators for TW sequences where rising i
i suppression pool temperatures increate the severity of the in-containment accident conditions, and the reliability of SRV solenoid operators sheul1 also L:e assessed for operation after eccident conditione that would cause rupturtne er vefitine of the drywelli (c) assess the reliabilities for BliR Mark i and Mark !! Iow pressure l
injection equipment for the steam conditions that might follow a l
centainment rupture or venting operation; and (d) the reliabilities of-I the PORV solenoids and the bloi:k valve' motor operators should be l
assessed for transient induced and small break LOCA environments (these components currently lack redundancy and were not historically included on EQ-easter lists).
Note:
Again, this would indicate that additional equipment may need to be added to the EQ Master List based on PRA studies that have been completed; this also suggests that the single failure criteria may not always be sat sfied.
l Uses of'in-containment instrumentation (other than trip functions) are
+
not well moda11ed in PRAs and therefore, a quantified risk significance is difficult or impossible to assess.
The study recomnwnded that research prograas assess which instrumentation is truly needed and in particular, which instrumentation is needed after the occurrence of
'Y
..L-30 l
e 1c'
~
.u,tn1 i i W u tu o i ln im i rtfi s.il n it w in18 s ni a
no.1<
i, non witni.m..
tun unnRnt nn.I.
.,inini i
i oui n n
iwer i
.nmnntnani nn n 4i, v.l.h.a
& n un o.in. nil.,,wl,I nnl,iIl n..ni u. u w
i 4
n ai a
i in c
i not n.mitat ui, ria ni,
- R!Ji'l!"Imil'lietuf'""" '" " '"'""
m
!!.nin.ini.n,en.
in.
noa
.,n.i n ini,
f yg %/,$jh1DL An' M. ' f It u pia tan at.i va m i teat s' nhas sa' walafin fand'
'an Mantuarina, and Lusamefafan' an Las' dant QM WAT ra 1
i M M Tunu1L R
ann,.nn.nr..it.nnoni.e4 liritui!!'J"fMfid"""""';nnh..i.u.hn i'.cong or.
..ir,. ui., a inn p" ". g' n gi"an, one.uwoi mcv
- sto, l
!%nl.lt'h j'ai'i tih"5:4u..! it?.ti, <
12i :f.v41 m n,l'tl1;f.n
't,
T'tW "' "
'tiP I,it,, n,i a
n.n. ir.tr, 6.L..
en n,
i n
l i
n o u,. id;hrlen. I.n.
n..(h.c.,,w,n n.u,.o,. I n,u..n. fin..n t.,n r.,
J n,
i.e..
n r
i in.i o
- n. pi, z n i, i a
u m n i9 ui.u a tilj!N"in ap'jituMd! T "!
.n. hit!T,itM.1h"fpI nun !.n{'It U.!diT y
d
.1 lt 4'
I L..n sn..
1 u u.,anoir.ornnt,uul.lm.r..nu.t.a6.n..
ur an f.ru.n,.Iin n an..
o;un.nu,..au.tunno,nin..,
. e.on,.
inu iu ru nio. o a.o.a ur
,.a i
tr...u i.n n,.
g g,,g;;a.o n...g a..atu..it..r.i
.k+... nn,..
m o, cir n a.a. iJn.u.n lu o.nv n tun,nn. lot.,io,. ur i
,, eon mdui<uitn J.,
i iian am o
inwung, u niran.n.J. ar o
run m wt i,,
ownn o a. un ni au vu,u".nn r.., n..o..ts n, w,,, a.
a,. o r., ig ;p... n,,..n.nunou r
,u.o
...n l
8.n,r.4.ri,a.s.d... no.iuo PliNh.M!W"""giun}n..om.M,on,1 an v in in.n)nnu..r.g nje imus i i n
u,s
' '""gc un.i4' " ""
yi' ""fl t il i
nivr a
iI i
l
" " ""O "
or a.mu3wr..nu i.n.c
.4 i.iu
.u..n..i.n. n.eth.n
.. n.,a, u
A thovah a qued m.hhnd, u.ri a
. inn.
kt f.uit i..itobilih pa.i M.
.en.H.lu soAsi n.enal.nt)
.rLa isai,n..iniiul. (thu h nn r1, ioni.nn3n>.v a wi uo.
n
}
I, u
re atning. Compressive modulus and density could 41so be somewhat effective for monitoring residusi life, although acceptance criteria l
would be much more difficult to estabitsh for these measurements bscause l
extensive testing has not.been performed to demonstrate that modulus and density respond consistently for varied test conditions. The results of the cable indenter modulus test did not correlate well with aging of XLP0 insulation, and electrical measurements were not effective for monitoring residual life.
Ncte This seems to indicate that condition monitoring may be a viable option for insulation materials.
The accident performance fin terms of electrical properties) of the XLP0-insulation did not ddffer substantially from the accident performance of XLPO-insulation aged at more highly accelerated (both conditions in past Sandia tests, as well as sequential and simultaneous)fyf inpastindustrytests,/g The IEEE 38344 post-l.0CINindrel~ band test en the cables that had been aged for 9 months induced crackini of three conductors of one cable i
type.
The high notential test dic not induce any cable failures (assuming the ca)1e did not crack during the mandrel band test), even after bands that were significantly more severs than the IEEE requirement. Thus, for XLP0-insulation, the most severe part of the post-accident exposure appsUrid to be the hand test.
I Issue:
The post-mandrel bend test may be too severe for EQ purposes.
j 1
The accide'nt performance of XLPO-insulation aged to the three different Itfetimes was not significantly different. Thus, for XLP0-insulation 4
exposed to environments less severe than those that were simulated, these tests did not indicate the need for additional qualification requirements as cables age beyond their current qualified life.
This conclusion is based on the technical finding that, with one exception, l
the cables tested did not fail. However, the testing perform 6d did not prove or disprove the adequacy of current qualification practices and requirements.
Note:
Apparently, Sandia has not attempted to correlate natural.
aging test results with accelerated aging test results to materials. qualified. life' projections for insulation validate '
HUREG/CR-5772. 'Aoine. Condition Monitorino, and less-of-Coolant Accident (LOCA) "ests of Class IE Electric Cab lee
- Volume 2 - Ethylene )rooviene Rubber nsulation. November 1992 reoort described the results of aging, enndition monitoring, and accident testing of electrir. cables with ethylene prosylene rubber (EPR) insulation.
This work was performed under contract for tie NRC by Sandia National Laboratories (f!N A1818).
The objectives and conduct of this research were the samt as what was l
.g.
~
s
described in Volume I of NUR[Q/CR lf?! except the tests were performed on s l e t t r i c. c ables wit h [PR tr.ul at ina b a t ert nn t he t e t t ing t ha t wa s pe r f ormed, the following conslusions were reached' Most
- properly Installed' LPR insulated cables should be able to
~
survive an accident after 60 years for tot.al aging doses on the order of 150-200 kGy and for moderate ambient temperatures on the l
orJer of 45'C to $5'C (potentially higher or lower, depending on l
spectftc activation energies of the materials and total radiation doses)
Dy 100 LGy, the residual elongation of the LPR materials that had a bonced chlorosulfonated polyethylens (C5PE) Jacket approached 0%.
However, some of the [PR insulation with essentially 01 residual elongatton remaining at the end of aging did go on to perform acceptably l
In subsequent LOC A t est s.
Of the measurements tested, elongation was the best condition conttoring method Although a quantitattve generic acceptance criteria is difficult to establish based on these tests, a reasonable range (that is I
fairly conserv at ive ) would be about 50-1001 absolute elongation I
remaining.
Compressive modulus (i.e., cable indenter test) and density could also be effective for monitoring residual life for some materials, although additional testing would be needed to establish Acceptance I
i r i t e r I a bec avse e s t eas t ve t e s t Ing ha s not been perf ormed t o demons t rat e
{
that modulus and density respond consistently for varied test l
conditions.
The electrical measurements were not effective for monitoring residual life.
Both elongation and Indenter modulus generally correlated well with aging under the conditions of this test.
[longation was usuall y a tie t t et measure of aging at lower tot al doses, while modulus bat ame a bet t er measure of aging at the higher 'atal doses, pas t icularly when the elongation fell to near 0%.
Note:
This information indir.ates that it may be possible to establish conservative criteria for monitority the condition l
of Insulation materials.
facept for Boston losulated Wire (BlW) cables, as long at, the insulattnn did not fall, the accident performance (as measured by it.sulation resistance during accident testing) of the EPR insulation was reasonably similar to the performance of similar insulation under more accclarated test conditions The accioent restitances of the BlW Insulation.
<v lower in this test than i n s t en t i a r, but more highly accelerated te :s 1
For [PR insulation '. hat had not failed during the accident test of l
cables aged for 9 months, the subsequent IEEE 303-74 post-LOCA mandrei bend and high potential testing induced failures of all EPR-insulated
)
cables that had bonded CSPE Jackets, except the Blw multiconductors.
The conductors that survived the mandrel band and dielectric testing went on to survive mandrel bends somewhat more severe than the IEEE 383 requirement.
Thus, the IEEE 383 mandrel bend and dielectric testing were very severe for many of the EPR-insulated cables, inducing the failure of otherwise functional cables.
I 1
L-33 e
)
m
.I
l 1
1 i
bte:
Perhaps the need and severity of candrei b:nd testing should l
be evaluated l
the accident performance of cables aged to the three dif ferent lifetimes a
l was not lignificantly different.
Thus, for (PR insulated cables exposed to enytronments less severe than those that were simulated, the s tests l
did not indicate the need for additional qualification requirements as l
c ables age beyor$d their current Quallfled life.
However, the testing l
performed did not prove or disprove the adeguacy of cureent 1
l Quali f it at ion pr er t ices and requ:rement 5.
4 l
MhM/CH-5772. 'Actna. Condition Monitorino. and Loss of-CoC ant Accident iLQQi f ests of Class Il flectric Cables." Volume 3 - Miscel'aneous Cable j
1ru ul at ion bstL _finnbtL.1392 1
j N i report no u r ibe.1 t he resulii nf soing conaltion monitorino, and arrident i
s. i t o,,; o f. i e. i, o i anle n.iin miu eilaneo,us iniut at inn a ype s the situal t si ipesimeni outuded multiple samples of aoble produs(s from lintkbestch l
l I,o t h tilltone rubber and toes tal cables), Kerite (Kerite f R insulated), and
)
(hamplain (polyamide, or Kapton insulated).
Ihis work was performeo under contract for the NRC by Sandia National taboratories (FIN A1818).
j lhe ph j o( t 1,e g and (ondg(t of thig research were the same as what was j
deu rlbed in Volume i of NURIG/(R l??!, except the test s wers performed on j
elec tric (ables with miscellaneous Insulation types as discussed above Based the testing that was performed, the following conclusions were reached
- on Most
- properly installed
- miscellaneous cable products that were tested l
l should De able to survive an accident after 60 years for 101.1 agtny do us of A; least ISO kCy or higher (depending on the matertal) and for j
moderate ambient temperatures on the order of 45"t to 55'l (potenttally higher or lower. depending on speciflC activation energies of the l
materials and total radiation doses).
Although the silicone rubt.er j
cables sttil survived accident testing, by 200 kGy the residual elongation of the elscell.neous insulation types was approaching 0%.
Of the me:surements tested, elongation was the best condition monitor'ing j
method.
Although a quantitative generic acceptance criteria is difficult to establish based on these tests, a reasonable range (that is fairly conservative) would be about 50-100% absolute elongation remaining-Compressive modulus (i.e., cable indenter test) and density j
could also be effective for monttoring retidual life for some materials,
}
although additional testing would be nseded to establish atteplante i
triteria because extensive testing has not been performed to demonstrate j
that modulus and density respond Consistently for varied test conditions.
The electrical measurements were not effective for monitoring residual life.
The modulus (i e, cable indenter test) showed good correlation with aging for the Rockbestos silicone rubber and the Kerite Jacket materials, but not f or t he Rot ht>e s t os c oa x l al jat ket mat ernal As long as the cable did not fall, the accident performance (as mea,ured by insulation resistance during act' dent testing) of the miscellanecus L-34
_- ~,
-.. ~. - -
I i
l Insulation types was reasonably stellar to the performance of similar cables under more 4" e'cre ed fett rendit tons The miscellaneous tr.sulation types that did not fall during accident
+
j testing after 9 months of aging also did not fall the subsequent
![L[ 383-74 post LOCA mandrel bend and high potential testing.
Except for the Rockbestos s' icone rubber insulation, all of the others passed l
l mandrel band and dielectric withstand testing that was more severe than the l(([ 38) #equirement.
for the allCellaneoul inlulation tyDes. the 1(([ 383 mandrel band and dielectric testing did not induce any failures j
of otherwise functional cables.
The accident performance of cables aged to the three dif ferent lifetimes was not 5tgnificantly different.
Thus, for the miscellaneous insulation types that are esposed to environments less seve o than those that were simulated, these tests did not Indicate the need for additional qualtftcation requirements af cables age beyond their current quallfled l
1ifetime Howsver, the testing performed did not prove or disprove the j
acequac y of current qual titcat ion pract ices and requirement s.
l 1ALA. lt thal u l l vm in a e _ fu i en p u ca. _.1L.P rgut t i.19 r.lu I a w Up g r a di ng l
end M4}ntaininu Luulumant Quallf b at tun )*. bs.Phil19.M. Ilullman. uf S ItthDuluur and biluultt).. int.. Junt.19h j
the Internallonel Atomtr inorgy Agency (l AI A) held a t ec hnic al commit tes meeting. *Hevtowing. Upgrading and halntaining (quipment Qualification.* from j
June 21 25. 1993, in vienna, Austria to discuss equipment qualification j
issues This paper was presented at the IA[A meeting, i
I Most of the Inf urret ton that was presented was entracted f' om an (leCtrlC Power Fossarch Institute ([PRI) buok on equipment qualtf tcation, Nuclear Power Plant rautoment Oualification Reference Manual.
The En Reference Manual provides a broad overview as well as the technical details of equipment qualification as practiced in the United States.
The following points were tsken from Mr Hollman's paper:
i in the older plants, some equipment types could be quellflod without a
considering aging effects ur subjecting the same piece of equipment to all acc ident conditions.
The [Q rule defli.
(0 requiremer t s for all plants, but permits the older aperatte; plants to use less st-
. gent [Q criterta The most s i g a t f i c a a ', c riteria dif ference between the older and newer plants j
relates to aging.
The older plants were allowe.1 to adoress aging using j
analysts, cartial test data, and operating expertence whereas agtr1 i
effects for the newer plants (and for replacement equipment in the older plants) was required to be addressed by preaging equipment during the qualtficatton testing programs.
for areas outstde contsinment, qualification is often required for relatively mild pipe brtak conditions (e.g., 65'C).
Issue:
The need for rigorous qualification of equipment located outside containment thould be attested.
New technical inforsation and research efforts continue to test the acceptabilify of older practices O' ten, new information suggests that significant limitations exist on performance and life predictions based on laboratory tests of prototype equipment.
Note:
Once again, the ability to truly determine a ' qualified li f e" is questioned.
PRAs indicate that demonstrat'ng eculpeent operation beyond the first few hours or day 5 af ter a LOCa is not risk 5?gnificant.
Yet, uti '!tles continue to empend *aluable resources qualifying equipment fer ocst accident operating tines of up to a year
$ 1 g r. i f i c a n t resources could be s a v e d by I t m i t i ng t he ma x i mum ope r a t i ng t i me requ i r eme n t s t o a mo re realistic value such 45 2 to a weeks.
Issue:
The need and/or ability to establish l i fica t i on beyonJ a two t o f our weet peri od shoul d be a s s,
J The LOC A acc ident radiation d.se values presently used for (0 are representative of a severe accident rather than tne values resulting from proper safety system response to a LOCA Note:
ignin. this suggests that the radiation dose required for equipment qualification should be reassessed.
When materials experience the low dose rates associated with normal clant operation, oxygen dif f usion limited degradation will not occur j
.wrrent [0 requirements do 'ot recognize t hi s di f'erence between accident dose rates and normal operating dose eates Issue The need to cons ider oxy 9en di f f us ion effects on aging d ur i ng no rma l pl a nt o pe r a t i on ma y not be wa r r a r t ed.
w:th regard to E0 of mechanic al eaut poent. little add'tiona. [0 efforts have been required for older plants.
F or newer pl ant s. mec hanic a l [0 has been addres sed by evaluat ing the e f f ec t s c' aging and ac c ident condittons on non-metallic materials.
Issue:
The adequacy of EQ of other safety-related equipment should be a s ses s ed.
Note The information presented in this report indicates that the i
current practices and methodologies for establishing and ma19taining equipment qualification for electrical equipment (and
.other safety-related equipment) should be reassessed.
EPRI TR-102399. 'Proceedinas: 1993 EPRI Workshon on Power Plant Cable
+
Condition Monitorino.* June 1993 lhe [lectric Power Research Instit.te ([PRI) sponsored a workshop on ocwer pl ant c able condition monitoring, which was held in San F ranc isco. C al '.ra i a on february 9-11, 1993, this was the second workshop on cable condition j
monitoring (the first was in 1988), and it offered industry professtonals an l 36
opportunity to continue the dialogue on operating plae*, experiences and hands f
in electrical cable degradat'09, icsclonmerts in the area of cab'e condition monitoring, and the results of ongoing research activities.
The report summary concluded that viable cable test methods need to be brought from the laboratory to the utility operating environment, and reliable correlations nave been established for assessifig the remaining life of cables based on
]
known polymer aging characteristics and artificially aged cables.
Technical i
pe esent at luns were t entered around the following worlishop lessions.
1 5ession 1 Plant Cable [xpertence j
Session ?
Cable Condition Monitoring Techniques in Use and j
Under Development I
Senton 3 Aging (valuation and Character 12ation Session 4 License Renewsl
?he fofIo-ing informalion i5 a summary of some of t he more not able wtie k shop presentations.
]
lytriltu 9! Woruhus. and_Qn291ng._LPRLittd t Londillan_. Monitor ing i
^
h our ams. L J. luman uf Ogdtrt.LnrJronmental and Lntrus ;try ttes wwana and J. J. Larts andMitin_9L LP.fu 4
An overview of the workshop and its purpose was presented and summaries i
were provided of cable condttton monitoring projects being spon:,ored by IPPl. including j
i l
{
Iingerorint Ing the Thermal History of Polymeric Materiali A tecni.lque has been developed by Dr. Paul Phillsps of the University of Tennessee to detennine the temperatures to which a t able hes been exposed when excursions occur in the vic inity of a 4
iable Specific details are contained in (PRI Report TR-10l?05,
)
- fingerprinting the thermal History of Polymeric Materials,*
5eptember 1992 i
Using an lon12able Gas to Troubleshoot Nonshielded llectric Cables Research was conduc ted at the University of Connecticut to evaluate the use of tonitable gas in electrical high-potenttai testing of nonshielded icw voltage cable located in conduits
w results of this resear.h indit.ated that a practital method for di scriminat ing between daa'.4tjed and undamaged c abl es may be possible Specific detatis are containnd in (PRI Report TR-10l??), 'Using an lontrable Gas to troubleshoot honshtelded Ilectric Cables," 0(tober 199?
Review of Polyamide Insulated Wire in Nuclear Power Plants Research and studies on polyamide insulation (Kapton'") identified a number of tonterns applleable ta nucleer power plant t-37
appil6situn6, i n61 oil i ng.
Innulpitun that in unir a few mitt thi&h may lie more tuni e rl il'I s in hanillin0 ebuts than heavier wall innulsi tonn that ere 1Heenanty unstli pelyest4s It tuni spiIlile ie aging and t raf king if lightly hent and tuhjef ted in onlit senillilann 41 sievaled Ismeeratursii anti suppture la a few megat aili ut raillallHn may toute the lefien glus Ihat osait the isoolallon i,6 tem la f all, leaving only the met hanis 41 nest of the owesiopping Iape Iayes n (nuh6eiluent malling i dulil i aune slei ti le al to s ehdown) 16tus Have asel9n vulnetaktillion heen adequalent adh eened hv t he kM f r....i r i a n n.'i s. I t. s e t.. I n o... e Iable Op i abiI 6 4, l
t rit i ile ve loped a e spoe l t hat i nvee n table irelem elentun.
inalsiletlon. nputshtiIly laines, in noevtee perfoemanae,
..... i e t i.i n e ps posisee an1 iseting anal avIng management the met tiene.1 he iPftl i Huileat Maintensruo Appllialinn Ionlet e s pi.. I INMA ) se Nr 'ent, 'rowee rlant Ps a l li e n t o inauee Iable up.i sl. i l i e s sul, lest M a i n t e n a ru e atul Ansiriit of \\lat tur ( sblot IMA%I )
trHI han spientur est r e nsart h t o delsreine i f ( nnvent ional t eni n e en he ennen ed to identify letal and everall tendillan the program avaluated phylltal and elsstrital lette en lahnratory ausil sabien anil t hose removed f rom operat ing plant s and alt rophy g li al t ail tec hniques were evaluated in which very small samples of insulation and Jacket materials were used.
The program determined which tests are most useful in evaluating the age of XLPE, Pf.
(PR. SBR, and PVC The study indic ated that low frequenty i
dielerirIc spor t rosc opy (ould be usef ul in evaluatiny the og t ro) of l
rables The study also confirmed that partial dischar ge tesiing tan deleil phygt(al damage at all lotatlont un the sur late of unthtelded muIll conductor i ables using surrounding conduc t or s a s the ground plane for the conductor under test the MA5( researth is continuing and as such a report of the results has not yet been i.s ed
[PRI (able life Program l
On going ac t ivities under this program include oxidation indut t ion time testing, pre-lontzed gas high potential testing, natural versus artificial aging of nuclear power plant components. and cable life database.
l 1
VildALjaa Induction i me festing:
This is a means of measuring l
the relative amount oi anti oxidants in a cable Insulation.
l l
Measurement of relative anti-oxidant levels provides an Indication l
l of the degree of aging of a cable material.
Currently, the University of Virginia (under contract to [PRI) is attempting to develop acceptance (rlteria and a standard methodology for testinq nuclear powee plant insulations L-38
t dilu L & Lv A. Ar_11 [1 v1 ALAu !ttu u !.hutJ e tr..f uita r.r.l AtiL.L umaunir lhe Univernity n fonne<ttvut it pe r 'twm ing re s s et t h t n inmpare natur al auing of polymerit materielt in nuc lear power plants with the aging that occurt under at t elerat ed laborat ory condit ions lhli prog am. diu us ted in l PNl Interim Nepori NP 4997 (December 1911 t. ), enfompattet tablet at well an other components.
Ihs retults of the first 6 yeart of the program are contained in (PRI Report IL 100/45, " Natural vertus Artificial Aging of Nuclear Power Plant Componentt,' January 1997 Stept are being taken by trHi to enhante the t able life program by adding a tontrolled, j
long term, low level oven aging program (allltLilL@<h4Ag A database It being developed whlrh will (ontain table informatlon related to aging, tetting, rensarch, and opes at ing esperlent e the rurrent af(ort Ii evalualIng ihe typet of infurmetlon available, iti reletIve importance, and Ihe moti appropriate format f or Inf ormat ion storage in a det aliate l
l Nnte Induttry progrant appear to be aggrettively purtuing toses of the rurrent (Q programatic innuet I
It
- llettrital Nunihuring and Analista Prugram at the Davis 8: 13e Nutlur J
runst H a yun,'
gr 3 iragnet vf u ledu hdlaun Lumuant and utquoi 4
40 Arid L'av id yar du uf,LA>uirlituriufLM ethnoluuivi tviour ativn An ove'. low of the Davi t liente Peeditttve Maintenante Program, with spot Ift fin un ami diis ut a lon on Itt i l ot I r 4 al MonttneIng and AnalytIn i
i l' e n,p a m ilHAP) was presented the IMAP wat built eenund the ( ( AD j
i s,ttom 1000 which provides the tapability to monitor important plant l
elei t r ti al tomponenti for degradatton by obtanning data that ather i
lla n t i llette predittive mainlananie programt 4annot the l( All n y t t om c an tie wied in *cature (or 4 ali ulat o) A( and D( voltage, A( arul (H iesistan.e i apai it ant e, di t t ipallun f at t or, Induttanse, quality fattor, insulation retIitante, pol.Ilatnon ratto, and a time domain i
refleitometer (10R) trace for the circuit under test.
Problems and trpet of degradation that can be detected by (CAD include changet to dielegtric mat erials, deteriorat ion of (Ircuit Insulation, hlyh eettnIaruo eonnettlont, thurt ( Iri yl t 6, upon Ctrtuit6 mu l t t ut e intrusion, ttrault notte, improper ground and/or thleld innnetttont, and development of thunt conducting paths, tri livLjp4ffurtt
([I E v e and 1,l t Lil.1L {4 } u!h!!!4L illy Lumuint. (L o )linL 2glJmer. Cluftd$,
Qtt.fruur Am" Lur i MLlau lt e o f ffit pfstentatlon was a folin= on to an eselier piesentatinn that., 4 4 viven un the tama subjett at the 1980 1 PHI warnshop on Power P1 ant IabIe
( oruli t inn Mon t i or ing, in \\an f rant t u o, talifornIa, Iobruary i t, in.
19dit the prin eedinut of thin eselter workshop ase doiumented in IPHI It e po e 1 il/NP!t\\ f.914 \\H, July 19 till l
Du.Ing the lit.enting protest, the Perry Nuclear Power Plant (PNPP) e s t ablished a program t o monit or temperature and radiation effetts un s a f e t y r e l a t ed t. abl e s in verlous plant environments the PHPP table i 10
monitoring program Intluded tracking of temperature and redlition (onditions in tolertett loaattant in the plant, remnval of rable samplet at five year intervals for tetting and compartton of the efforts,of art ifit tally aged t able to natural,ly agsd (able lhlt presentation i
provided the table lett results following the first flve year interval 4
of plant operation.
]
In general, no signific ant polymer degradation was identified following the first five yeart of plant operation Also, no conclusionn 4
i nnt erning t he e f f ort s of natural aging vertut art if trial agtng of r able insulatton iould be made Nonethelett, the following major observatinnt were made i
Notkhettot inntrol iable enhibited dela ti at les dueIng ietilog I
I the intulatinn remained pltable and showed no vitliile tlynt of i
ilogr ad a t t un, but t he more t evere env i ronment a l saputure (theemal and rad'allon) may have faused more degradation than antiaIpated j
..., a
,,.e
, e a, p s., i> a v se t al tons in tensile tirength and alongal lon based on the ininr
{
.ie s atile type wan i
.. f the iniwlellon inilopendent of manuf ai t ue s ident i t ied i
l litue.
The color of insulation material may influence the rate of degradation of insulatten material.
I i
t)ue t o delit tent iet in the methodology that was Inillally estabitthed at the ontet of this program, it turned out that the artificially aged cables were actually aged more than 15 times l
what was actually needed to timulate 40 years of cable life j
i on t eepson t l y, a valid t omparlton of t he met'han ti al proport le t of i
the 4,ttflitally aged aeul nat urally aged tablet tould not be inade l
l D
'In Plant.indtntar Unt4lcamocatalth td11onllin112.JL 1 Jaman uf Quden invircnmar)t4l and knpruy Servity Lum 0Anr. Int, and $. Iluntader and O. Petsf1 uf 1,gmmynwealth [dilun I,gepany the resulti uf in plant trials of the Ogden/IPNI Indenter Polymer Aging Monitor weg protonted the indenter it a r!'at hine t ha t de t e rmine t the iomprenstwe modulut of ( able insulatlon and jatket matertal the modulut is determined by pretting a probe of known thape agalntt the well of the table at a (Isod veloilly while mea sur ing t he in* i s, a nil t hen di v iding t he t hattge in foeie by the thange In put it ton due leiy Ihe inward motion of the probe the indenter modulut can t 9ted for frs'hIng aging of materiali thet rhangs hardnett th proportlon to Ihe tumulatt.ve effortt of heat and radiation.
Such materials include ethylene propylene rubber, thlorotulfonated polyethylene (l5Pt),
polyvinyl thlorlds, neoprene, butyl rubber, and tillcone rubber the indent er i.an not be used to trac k the aging of materialt such at trost-linked polyethylene (XlPI), where material hardnett doet not change in proportion to age, the Indenter was used to test a verlety of cable intulations at Dresden 3, lion 2, and LaSalle 2 the results were promiting based on i 40
the data that wat obtained, and the following conclusiont were pfatented, The tuerent indenter 16 practical for une in plant applications.
1 Nu g lyn t f it ant m(ult f li at iont in the indenter are needed for aontworttal une lhe indentef setulth tu date indliate that the tablet that were teited have not aged nIgnifteantly and appear 1o be aging at a slow rate
'he si.pleested aging teilt indtiate that useful s'implanie iLiesla ate a t h l e v alil e
'ho indenter it a useful tool for evaluating cable matertalt that manner.
these #natorialt Intlude an orderl{he indenter will not be useful for h e r iten ce soften) in I P 68, t\\ll, neoprene, and PVL I
i it. : iables unlett a jaihet with t rendatile proport les i t used 1
t u e t t.e r evaluatton of jethet aging proportist and temperature effeits is noLettery 1
"LULA.11aling._ol.DaaAata lablai " Au. A__Y tal.L.9L.Rlactt 10d Litulnieritiu Ai}9Llataa.And MAtA.A datoDua.and LurLli f. NeIson.of 1&lllj l 4 ft & L lulle l Labufaluflal i
I
'he results of esperiments conducted at $andla to assets die lei t e li j
withstand voltage testing on iablet and to attest the sur viv abilIt r of
{
49eil and damaged (ablet unif e r lost of toolant aiiIdent f l O( A ) t ond ' t t orn
=st presented Of parttiular interent, all of the Otonite satlet (with l
l e peeiondiitoned iIaw) failed the iOfA tenting thermal aging and the i
5
.. s e of a b o n d e.1 in. k e t appeseed to lie the two majoe f ei t oe i that mar l
have iort.ihuted to the une a pei t eil f ailui e of t he se n atile s the testing l
that was performed on the Okonite tablet simulated aging for 40 yeart at about IS*(
while the Cablet were actually rated for a 40 year life at 90 '(
two Hot hbe tt on t ables (wit h a prof ondit ioned flaw) sito failed
.tus ing L tH A l e t t ing whit h may have been due to the timulated theemal auin9 tonditiunt litus:
The qualification of bondad Jackets needs to be assessed.
- DeufadallunMupl}uringlettfurPulramidtlfoulallun.'A.M.Uruningof I
LEL(l umethaliltal. Utilun (urf$4!)r l
{
lhlt paper it a f ollow.on t o one t hat wat presented at the 190fl ( a b l e Deyt ada}"lon wor k thup c onc erning degr adat ion o f poi yamide insulation (Kapton ) in haval alltieft fl y evaluating Il temples from the /n milot of wiring on a Naval alrtrall, dettsinnt were made at ta whethoe ior!ciIIvo aitlong war e nestled
~lhe ptutett that wat developed foe I
l 4i
resolving thIt. problem was based on an understanding of the chemical methanismt involved in ibe aging prorets, what ionstitutes the end of 1i f e i ondit ton, and t he it at itt it al met hanit t hasit for evaluatlon.
An equation wat derived (numswhat analogout to Arrhenius) for calculating the remaining life of the polyamide insulation for the allumed op> rating (onditions (such at temperature, relative humidity, and end ol-life strain sondition.
It was tu ge6ted that a similar approach may be possible for uti Ity applicat on for various insulation typot.
i Ant gny)Lf.gwI,h nrtsf r.
n&Lyr.g jlgi and 1 g g at'on for the fvAl s
ti.h2 RLDomain
).a,actric W ee tr cht sev'
(
w h. Mann i
the Univgritty of fennett ucut and h i c h an'
.abow o o n t o' idatag a E n (EDiaJ lhe result s of rensarth nn the appilr ation of time domain dielet trit i
spectruu opy ('01) for trarking cable insulation aging routed by heat, f
I muisture, and tont ut with mineral oil was presented.
Preliminary research Indicated that characterlitic changen occurred in the 105 surves of the table insulations tested, and that these changen depended well a n. on t he t yr.e o f on the type and neverity of the aging ntrent at insulation it was observed that aging in oil or wet shvironmenti produced more signiftrant changet in the 105 than in the elongation, wh t( h suggastt that the elettvtial proportlet of the eable malestalt are affeited tu a greater eatent than their mei hanis al pe upse t io n when agloy is peeformed under these sundtt6unt Pratently, a eesaarth pru.joit Is undeiwa, to develop a systemalla 10% signature file un 1 c omrnonl y used low.oltage powet plant iables and to devolup a portable i n s i e unie n t iapable of generating IDS data on installed power plant tables drn..
mandat d. raun.to) ofA..herra.
har atte laatig of A e Latjo Lbit haulation. Lo tri H
- f. s c f.I a and ihrt L
sh Ja2 urn. y alsn and Stonkwi s. Lar-Several physital diagnostic techniques requiring micro temples for monitoring the condillon of nutleer generating stattt)n tables were studied the types of insulation that were evaluated uting thets techniquet inc luded polyeth l PI), trott linked polyethylene (ILPl),
ethylene propylene rubber (y ene (
IPR), butyl rubber, styrene butadiene rubber t sllH ), and polyvinyl chloride (PVL).
Intulation degradation was astetted by c onvent ional alongat ion measurement, dif f erent ial u anning (atorImeter (05C) onIdatton Induttton time (0ll). DSL oslaatton I
trulut t ton t emperature (under h tyh ps ygen pro t sure), infrared 6atbonyI alaor pt ion, pl a s t it lier ( ont ent. af*d gol tontent/ swelling meatuiements Ihe tentittwitles of the diagnn'. tit techniquet to measure onIdat ton and embrit t lement were c ompared wit h the elongation results, and a artteston l
f or monit oring t able degradat ion was developed Hanget of values ot>l a ined ut ing Ihe se teiholiuos -uve peetonted in a table =b li h. uu l.t 1+c
.. sed tu assett the deusse o do y e ad a t ion o f s urvenon l y in s e d insulating nia t e r t a i s the f o l l ow ing obt ro v a l long and tuntlutions were stated in the report As t able*.ge, there is a gradual embrittlement of the insulatton material, leading ultimately to trathing l
l l 42 I
l 1
I l
l Dif f erent physical properttet must be measured to atlets deteriorotton affe.11. sly.
liongation valust were not a good Indicator of the intulation rondit ion f or XLPf, IPR PI and $5R because of a long induction 4
period.
l Note.
Ihlt teest to be temewh6L contrary to teme of the other studiet that have boon completed.
1he tensile measurement was the best Indicator of aging for batyl 4
+
r ulitie r and for PVC An 011 me a surettient wa s a tensitive technique for measuring the remaining osidative stability in IL Pl. Pl. and l PR Materiale nearing embrittlement showed 011 valust of long than one minute i
Well stabillied materials (at retelved and further aged) eihibited Ulf values in estent of fine minutet.
No control nample It e silu t s ed t o attent the inndtilon of field aged satilet I
the unidallon Indut t iun t emperat ure mee turament wa s a tens it ive indit ator of osidatIvo stability and aging in butyl rubber and
\\llH Intulations that have undergone considerable aging nuidited at 14 -(
lower than the unaged tpstIment-A referents temple for aumparItun it required for making a definitive attentment of field aged cables.
Solubility and twelling measurements were good indicators of a hangen ai rompanying t he ou tdat Ivo trott linking c hain tr in t ion of et en n anil (t hee a f or e) of the I ondillan of buty t ulitie t and l'VI intulated iabler, ilQntfitantly aged FVL thowed ieduted talubility tidt rahydroh"an, and but vi rubber showed increated iniubility 5e toluene int Pi, ( PR, $llR, and Il Pl wi t h hlyh inillsI gel enoteht, the twelling meanurementt may show ilel ei t alit e diffetsntet only when the insulationi near embrlitlement j
lhase was good spreement lielween psattlalter lost and doi s sa te In
'l e
A minim m level of plastitltet u
wat e squit eil to alunuallon of PVL prevent embrittlement of pvt and a plantt(Iter level of lett than s '.1 li y mast tould indleals s lyn t f it ant aging
\\leinluial thanget at i r iliut afile t o aging i ould be del se mined - 6 I h
=
fourter transform infrared Spectrotropy (Ilik)
Aged insulallont a shlbit ed a c arbony l absorpt ion t and at IMO Im ' be f ore thangen were sletected by elongation measurementt, listaute additiven tuih at antlosidantt Ian interfere with the antignment of the (artionyl peak, comparison with a control sample could indicate whether the table han been oxidlied.
Innventinnel electrital diagnoltle measurementt (dr Insulation ret it t anc e and polaritat ion inden) proved completely Inter,t it ive to the thermal aging of a broad range of cable Insulat ton matertalt l
l 4)
s Thelowfrequency(<0.01Hiptan4measurementswerefoundto change markedly with therma aging (and measurement temperature) while no significant change was recorded at power f requency.
The low frequency tan 6 measurement may have applicability to a rance of intolat ion tyttemt beyond table telling I
- In. lltull&ungillL flgnjluting.W1_l'bntL fj ani L&D1 & 'J.,J. )(ullAua.
Az111in.andK.
ditamina dn1E_1sJLratD L.L.itt0 Lith H
\\everal diagr.ut t ic proc edur e s were us ed t o a t t e s t the jatket degradation of thermally aged (abies.
The diagnotttt procedures inc luded the t rule n t e r ag i ng mon i t or, tonic velo (It y letter, and near inf r ared f
esfieslanie (NIN) analytit.
It was thought that atletament of the jatket (undltion would provide (in most caten) a conservative stilmate of the insulat ion condit ion.
the results obtained from each technique were discutted in thlt report Although eath of the techniquel sahibited tome potent ial, the NIR analytit method proved very Interesting; the concept it based on molecular changet that occur in the jatket material at it ages (at it the Late with fil$ analyllt dlicutted in (H) above).
for the NIR tec hnique to he Imslemented In monitoring cable at ng, a t allbrat ion i
t urve munt he est a allahed using temples of known c ogradat inn During thlt study, the NIR analjtll provided a prediction of 28 38 days and !?0141 days of aging for two test samples which were ac tually aged for 40 Jays and !?4 days, respectively Note.
thlt teset imprettival J
"Aulliu.alW.LunQlllunjiunilufJ0uufi,lan) ll tablt)." *laf A J JaLWDun uf h Afld11JiallQnal..LADWI ALQLita A brlef summery of ihe retultt of an NRC sponsored tett prog,am on aging, t undit ion monit oring, and alt ident te s t ing of C la n t 11 ( ali t e t was presented the following conclusions were stated with regard to condition monitoring of cable Insulation:
Of the parametern lett ed, alongallon at break had the tient a
(orrelation with aging for the most table types.
Note; This toest contrary to some of the other studiet that were performed (see H above).
Hardnell and indenter modulut measurements both incr.. sed with aging for some of the matertalt, especially the Jacket materials, and the modulus measursments were significantly more sentttive to aging than the hardnett measurements.
Wher e indenter modulut usi ions 11Ivo tu aging, 11 was mott tensitive in the later staget of aging, after the alunuation h4J reathed nearly 01.
!)en t i t y generally Ini s saico alth aQlno f ue mutt matettalt, but inme thangea were int ont I t t stil i 44 i
J
5 tfith only a few estoptions, tantile strength and a numlier of diffsrent eleitei.si " a..orments dt.1 not i nrr elat e well wit h I
aging L
t
'l'ulant A.ltthntaut toft Destruttire lyl lustjof).yf L4 Die K
l'itarla 3.'.Asnittih.h luf i(len.tnd R uur L. L
.sugh 05.)andj a Nallont!
.. Gl Lapur alG[itLa()Q JatnSL &llL tun.ADi[ tQL.dLtD$ttiLS M}t,Piparlmthl 9f l
l'Ql.:mtL.1tthn01 DEL fluA; fitt11 ult _.9L._KhnQ19U Lutden)
I l
Iwo techniquel based on dentity change 6 were pretented at potential non-dottruttive evaluation methodt which could be ute for monitoring the mechanical condition of cable matertall in nuclear power plant environments.
The first technique wat direct measurement of dentity ahanget unIng small thavingt removed f rom Ihe surf at e of ( alite jet keln.
I the te(ond te(hnique wat dentity meaturement by (omputed a ray tomugraphy til). alto referred to at tamputer attitled tomography (LAI) the (l method wat par t l(ular b y atIrattive tiet aute i t ( ould moni t or ile n t t t, ihanymt in lio t h intulatton and jaatottety malesfalt the intital tasulIt of I I IeeIthg wete psomtiIng and iut(hoe woe b wee antisIpated I
Note A lot of information was presented at this werkthop that may be useful in ottabilthing condition monttnrtng methods itVKLk Ll* V141...*Wuf uhup}9ft.l'.!sulam. f or llimintLlun of KeQuirements Marwinal 19hafety." Septemtier it j At past of the Nutirar Regulatory L cmi t t ion's pr ogram t o elimina te re @ irements that are marginal to safety, the NRC conducted a public workthop on April ?? ? fl. 1993, in nethenda, Maryland.
Ihlt NURlG/(P delalled the pe ni eed ing s of Ihat woebthop anil, of partliutar toterent, \\estton 6 of the NUHit.tle contained the Indus t ry ' t viewn on environmental que'.tftration regt rements that have been imposed the f ollowing coment s and tuggest ions perteining Io lQ requirementt were taten from the NURIQ/IP-IQ programt require signifl(ant resourret to att ablish anti maintain the epaa l I f is a t lun of hundredt of ploiet of eles ir le al equipment Iheoughout the plant these retourtet, lh many ratet, go well lierond what is appropriate with respect to the r elat ive s ink attot lat ed with the l
per f ormanc e of t ha t equipment Ut11itlet are current 1y ( neen i t t e d In pnti at1ident opreabittty t l ene s that range from JO dart to one saio that su cedt a year the v ist tignifitant eventt that impai I c'psipment typtially ni t or wilbin the ftett few dayt following ihe ai i t ilun t W61h iew e ni opt lons. long t es m post at t ident operability tuntelliolet minimally to elsk thlt it dit<unned in NU41C/CH 6311 atol ineiont PRA i'inf l e a support Ihlt i uni futlun With r#Qard to pQuipmt-*. that 11 exposed to radiation harsh (onditions:
a) t het c 1s no rocogn'.lon of allowant a made f or equipment qualIf le at non 1 4%
I i
where the postulated iidletten levelt are insignificant with resport to the relevant materialt o ' rontirurtion, and b) the regulation requiren equipment to be quellfleii to environmenti that are lignif frantly enre sevare than what the equ'pment would reallatically esperiente polla r
actIdent.
the name qual titrat ion r agor (I s., pre aging, acc idsnt timulation, qualtfled Ilfo, replacement, etc.) 16 required regardlest of whether the equipment may be esposed to the lean severe, short term pipe be sak tonditionn that are postulated to occur outside containment or to the more to. ors (ondit ions that are post ulat ed to o(t ur ins ide s ont a ientient dutIng a 10(A 1 van ihuugh qualifled iife ettImatlong are far from ensi(, ou fleilbility it allowed in scheduling end of life maintenants.
A grate period on the order of 161 thould be allowed consistent with the standard that has been tot f or lei hnic al 5pec if tt'at ion surveillance sequie montt Only on very rare ott allonn and very reluctantly han the NRC staf f allowed ll entset the opportunity to demonstrate that there 14 no need partisular piece of Regulatory Guide 1.9? squipment.
to qualify a
the NHL staff hat not allowed the applicallen of leak before break in the structuring of IQ requirements even though thlt iontideraltons philutophy hanboonallowedinotherregulatoryappll(ations.
The current IO regulations and guidelinen discourage the une of anything but full sequent ial lett ing to att ablish the qual t f trat ion of new aint rept ar sment romponents in general, the NRL ttaff will not arrept analytit in llou of lett data
(.iven t he 14 plut yese t of indutter e spe s ioni e in tumulative (Q testing, estearth, opetalino espetlenie, and lettons leavned, more ententive une of engineering analytti should lie allowed.
litue Indust ry representativen have presented reasonable and rompelling argumenti strongly suggent ing that the methodology and prart ti st for estabilthing and maintaining [Q thould be roastatted.
Argonnt hattonti La7 oratory.Drafy.Kaport rt; ljN AIJJ0.yll1A A)nlunment No, b.
- ltinA 1mDatt uf [Q ituulttment!J UI Upstallnu ht&Llyr!.. Murembel V. 1V93 Aegonne hattunal Iahoratory ( ANL ), under contract to the NRL, perf ormed a study to attent the risk significante of environmental Qualification requirements the work performed by ANL was broken into the following two tubtask6 Subtank No. I Collection and (valuation of Rollability Data for (loctrical [quipment Performance in a Harsh
[nvironment A Llterature Review tubtask Nu. /
l dent If it st lon. ( herat t erIf at lon, and I va tual lon of Hlik. impor t ant Au lelent %: enat ion Holst ed t o 10 liguet t th t
- " + -
--e
,4.
pJ
.a a
A i
o l
Under tubtalk I, ANL concluded that reltability data for erformance of j
electrical componenti in harth environments wat entential y nonealitent, llowever, bated on the review that was performed. the following qualitative insights were recogntied:
the IMl f telt data Indicated that moisture intrusion at the electrical penetratton to a device is a predominant mechantam of failure.
1ha Sandla teatt of ' undamaged" cables thow that the observed failurst alturred dart after the harth environment wat estabilthed.
i
'Predamaged" cablet seem to be more likely to fall in a harsh I
envIrnnment than the ' undamaged' rablet.
]
Inntallation and maintenance practicet may af fect the reliability of elettrical componentt in a harth environment.
Synergittlt effects do not appear to be algnificant for cable testt.
the availalite slet a doen nni provide adequate informat ton on the une of i nnute t able t ent s to predict f ailure of mult it anduc turn.
- lnsulation resistante* may not be an adequate measure for the i
a per f ormans e of an ales t ris al component i
4 there are no tests that addrets the differences among the variout (Q e
requirementt.
fallure of other electrical component 6 luch 48 electrical penetrations a
i and connector attemblist may be more important than the fatture of elettriral tablet q
litue:
This information thould be conaldered further by the staff to enture that qualtftcation of thte equipment it adequate.
lhe three eier trics) penetration attemblist tested at landla failed at 6 hoort, 1) hourt, and at 4 dayn.
Intulatlon retittance wan used at l
the f ailure criterton.
llated on the Sandia lett s of elec t ric al penet rat ion at temhllet, li isomed that insulat ion ret int ani s by i tsett may not alwart lie a goint indst ator 'af electrical performant e, and the rate of v otitt ent e degradation depended more on the type of cablet and the loads than on the particular penetration detton being tested A sev' low of letti performed on tolenold operated valvet 00V6) Inditated that the telt procedures and at,teptance criterta may not be directly applicable to nome SOVs that are potentially important to the prevention i
of fore damage.
the important 50Vn being referred to were those used for the PORV6 in PWRt and those uted for A05 In the BWRs.
lete 10Vs are normally doenerglied and the failure mode of interent It I"re to 1
energits and to remain energlied.
1-41
a Issue Adequacy of EQ testing of solenold valves that are used in power reactors thould be atletted by the staff.
11mulation of thermal aging is typically based on the Arrhenius equation, which it very senlltive to uncertainties in the activation energy allt,med and the component ambient temperature.
This poses quattlons on how reallitit the calculated simulation conditions are, Notei Once ag6In, the ability to areurstely determine a quillflod life in quantioned, One Sandle report quattioned whether the testing..tvironments specified by ll(( 323-74 were tevere enough to simulate actus) accident (onditions.
Ior the PWRt that were analysed in subtask I (turry and Secuoyah), it teamed that the sequencel with potential increate in Core camage frequency due to harsh environment are initiated by a small or very imall LOCA, therefore from a risk perspective, the tetting environment speilflod by liff 17).t14 may he conservative.
Notes Perhaps this information enn be uteful in focusing (Q requirements, Previout work performed by the NRC llaff and by landla National Laboratorien (NURl0/CHet313) han shown that the impact of harsh environment on the reIIabit tty of in-contalnment electrical componenta may have a algntf trant 4
Impact on the core damage frequency (C0f) as presently quantified in contemporary probabillatic safety analysen (PLAs). Under subtask I, ANL environment on the (h charactertle and quantify the Impact of harth stlempted to identif building upon the findings of previous studies.
Since reliabilit y dat a for elec t rical rompnnenti in harsh environments la non f
esintent, tantillvit y analytet were perf ormed based on the Nultf 0-Ill0 PkAt for i
t wo PWit t ($urry and 5equoyah) and one AWR (Peach Bottom),
the ANL ttudy only considered accident toquences that could be i..pacted by electrical equipment failurst inside containment Ahl reached the following ranslutiont-i for hurry and hequoyah, the impat', of harsh environment was dominated by a
sequences initiated by small or very small LOCAs.
Note.
Again, thlt informatten may be uteful in fecullnj IQ requirements, The important in Containment electrical componenti that were affected by a
harth environment for Surry and lequoyah were the PORVt (and/or block valves) and the steam generator level control / detection equipment Ihe impact of harsh environment upon CDf for PWRi 16 plant dependent-the sequences initiated by a medium or large break LOCA followed by a e
failure of hot le0 recirculation could be significant if the recirculation motor operated valven are all located inside containment.
f or Surry and Sequoyas, however, this was not the cane.
Although the main steam line break Instdo containment scenario was not tperifically addretted at pas t of thin study, ANL recognised that the succent criteria for thlt initiator was basically the same at for the small break LOCA6.
In the 8ene of l'each Gottom, ANL concluded that the Intermediate LOCA i
toquence (Intermediate break LOCA followed by failure of itPCI and the j
and the lmall LOCA lequence (small break (%A l
depretturitation tyttem)W HPCI, RCIC and the depressurtsation system {
l followed by failure of f rould putentiall,r contribute ligniftsantir to the total C0f because of I
harth environmenL effects on the 10Vs of ;he depressurtsation tystem.
Note:
Although equipment reliability numbers are not available for performance in harth environments, mWeh of the informatten presented in this report indlettet that PLA may be useful in focusing (Q requiremente, l
l l
L 49 J