ML19260C353
| ML19260C353 | |
| Person / Time | |
|---|---|
| Issue date: | 12/05/1979 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | Mattson R Office of Nuclear Reactor Regulation |
| References | |
| IEB-79-01, IEB-79-1, NUDOCS 7912260235 | |
| Download: ML19260C353 (49) | |
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DEC 5 1979 -
MEMORANDUM FOR: Roger Mattson, Director, Division of Systems Safety s
FROM:
D. Eisenhut, Acting Director, Division of Operating Reactors
SUBJECT:
OPERATING EXPERIENCE MEMORANDUM NO. 22 - ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT Problen On February 8,1979, the NRC Office of Inspection and Enforcement issued IE Bulletin 79-01 entitled, " Environmental Qualification of Class IE Equipment." As a result of the licensees' reviews of equipment qualifi-cation initiated by the bulletin, the Division of Operating Reactors has received 28 notifications from licensees of safety related electrical equipment that they have concluded is unqualified for operation in accident conditions. Enclosure 1 is a listing by plant of the equipment that has been reported.
Safety Significance The safety functions which could be adversely affected by a failure of an unqualified component cover a wide range of safety significance including containment isolation, containment pressure relief, containment purge, main steam isolation valve position input to the RPS, valve position indication, process isolation, containt:ent cooling and fission product removal, ECCS pump bypass isolation, stean generator level transmitters, and ventilation cooling for safety-related equipment.
DOR Action In each instance where an item of equipment was detemined to be unqualified, DOP. has evaluated the impact on the health and safety of the public and the adequacy of the remedial steps to be taken by the licensees.
In some cases, the licensees elected to replace the unqualified equipment imediately; in others a basis for continued operation pending corrective action at a later date was provided.
In those cases where the licensees proposed to 1617 079 7912260
8
. continue to operate the plant for a period of time before shutting down and re;, lacing the affected equipment,' the following factors were considered in DOR's evaluation of whether the plant could continue to be operated safely: (1) redundant / diverse components available to perfom the required safety functions; (2) locking the affected component in its safety position; (3) administrative actions and revised operating proce-dures; (4) additional operability tests and inspections; (5) post accident mitigating actions avarilable; and (6) fail safe design features.
In all cases where centinued operation was requested by the licensees based on
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a plant specific safety evaluation, 00R has concluded (contingent upon additional staff requirements being satisfied in some cases) that the plants could continue to be operated safely.
In addition to its reviews of specific qualification problems identified by the licensees, 00R is working with the Office of Inspection and Enforcement to determine the adequacy of Class IE electrical equipment qualification in all operating reactors. These reviews will be conducted using guidelines prepared by DOR (see Enclosure 2). The objective of the evaluations using these guidelines will be to identify Class IE equipment whose documentation does not provide reasonable assurance of environmental qualification. All such equipment identified will then be subjected to a plant application specific evaluation to determine whether it should be requalified or replaced with a component whose qualification has been adequately verified.
DSS Action Concurrent with DOR's review of electrical equipment qualification for operating reactors, DSS is reviewing several applications for operating licenses for new plants under construction which are nearing completion.
We understand that these reviews will be based on criteria developed by DSS and set forth in NUREG-0588, " Interim Staff Position on Environ-mental Qualification of Class IE Electrical Equipment."
Recommendation In view of the problems with electrical equipment qualification which have been identified, D0R and DSS must coordinate their efforts to assure that future plants licensed (in particular the near term OL plants) have electrical equipment that has been demonstrated to be qualified to at least the same confidence level as that which is being required of the currently operating reactors. We believe that the D0R and DSS equipment qualification review programs currently underway provide the necessary frame work to accomplish this. Specifically, since the 00R guidelines are written in the general character of an SRP (standard review plan),
DOR reconuends that the DSS reviews of the near term OL plants begin with a comparison of the plant equipment qualification to the DOR 1617 080
s
. guidelines. Once equipment whose qualific,ation is in question his been identified using the DOR guidelines designed specifically for this purpose, the component specific issues raised can be resolved in accordance with the criteria and positions in NUREG-0588. This would be consistent with the 00R program because our goal is to use NUREG-0588 whenever possible to resolve open issues from our initial reviews using the 00R guidelines. This proc.edure should assure the necessary consistermy oetween the near term DL reviews and reviews of operating reactors.
In addition, we have determined that many of the operating reactor licensees are referencing topical reports still under review by DSS as the basis for qualification of electrical equipment in their plants.
In order to expedite our reviews currently underway, we request that DSS complete its topical report reviews as soon as possible. These reviews should be conducted using the same procedure recomended above for the individual case reviews. During the course of our reviews we will be contacting DSS regularly to detemine the status of the topical report reviews and to determine the extent to which the reports currently under review and previously reviewed meet the DOR guidelines.
.\\
A'cting ect r Division o Operating Reactors
Enclosures:
As stated
Contact:
E. Butcher X-27900 cc w/ enclosures:
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H. Denton E. Case D. Vassallo D. Muller DSS ads DOR ads D0R BCs 00R SLs E. Butcher J. Sniezek L. Nichols W. Stiegelman T. Dunning A. Szukiewicz V. Thomas D. Tondi C. Heltemes
ENCLOSURE 1 J
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STATUS OF 00R REVIEWS Iti CONNECTTDF WITII I D ULLETIN 79-01 24 TiDUR NOTIFICATIONS i
DATE OF 24 hr DATE OF 14 DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DEIAILED REPORT STATUS OF D0R REVIEW 1.
Indian Point, ilAMCO Model EA-700 stem nounted 2/14/79 2/16/79 Complete - continued operation is Unit 3 limit switches on valves in the justified on the basis of redun-following systems:
dant position indication, redun-dant equipment not exposed to the a,
liigh llead Safety injection accident environment and admini-System - llot Leg injection st/ative procedures in effect (Indication) until the switches can be replaced The switches will be replaced as b.
Containment Pressure Relief soon as a replacement is available (ContrrJ and Indication) at an outage of sufficient dura-tion.
c.
Containment Purge (Control,
and Indication) 2.
Vermont Yankee f1AMC0 SL-3CH-7L and SL-3C-Ill 2/14/79,4/4/79 2/28/79 Complete - the }imit switches will stem mounted limit switches be replaced at the next scheduled' on main steam isolation valves outage. Continued operation is which provide position indica-justified on the basis that the tion and input to the RPS.
RPS input is not required in LCCA conditions and the limit switches are not required for containment isolation.
3.
Cooper NAMCO Model 3L3f1L stem 2/21/79 2/21/79 Complete - continued operation nounted limit switches on justified on the same basis as nain steam isolation valves Vermont Yankee. The limit which provide position switches will be replaced at the indication and input to the aext refueling outage scheduled HPS.
for April 1979.
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DATE OF 24 hr DATE OF 14 DAY
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PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED HEPORT STATUS OF D0R REVIEW 4
Browns Ferry, NAMCO Model SL3-B2W stem nounted 2/23/79
'3/9/79 Complete - unqualified switches Units 1/2/3 limit switches on main steam 7/31/79 will be replaced at next refual-isolation valves which provide ing shutdown. Unit 3 is ccmplete, position indication and input to Unit I will be shutdown in Jan.
the RPS.
1980, and Unit 3 will be shutdown in April 1980. Continued operation is justified on the same basis as Vermont Yankee (see item 2 above).
e 5.
Indian Point, NAMC0 Hodel EA-170 stem mounted 2/27/79 rior required Co,nplete - continued operation is Unit 2 limit switches on valsas in the an rarlier justified on the same basis as containnent pressure relief and rep (rt was Indian Point Unit 3.
the limit containment purge systems
- provided, switches will be replaced during (Control and Indication).
3/23/79 the refueling outoge scheduled for the sunner of 1979.
6.
Fort Calhoun NAMCO Hodel D2400X limit 3/16/79 3/28/79 LER Complete - the limit switches will switches on 18 different 3/29/79 Letter be replaced duri.ng the next two containnent isolation valves 6/6/79 Letter refueling outages (per LER #79-7).
(position indication only).
Continued operation. is justified Also later informed in letter since the switches are not used dated 3/29/79 that the in any safety-related initiation followin9 additional switches
- circuits, were unqualified:
NAHC0 D1200G 13x2 - 26 switches Fisher Type 304 14x2 - 28 switches Ch CD CD U
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DATE OF 24 hr DATE OF 14 DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF D0R REVIEW 7.
Salem Units 1/2 The licensee reported several 3/6/79 None Complete - since only the limit types of equipment without switches were declared unquallfles 3
qualification documentation continued operation is justified i
(NAMCO limit switches, on the same basis as Indian Point Instrument Panels, Air Unit 3.
All of the equipment operated solenoids, and identified including the unquali-i Transmitters). Only the fled limit switches will be I
HAMCO Hodel D2400X limit replaced at the first refueling switches on containment shutdown the first week in April, isolation valves were declared 1979.
unquali fied. These limit switches are in control and position indication circuits.
8.
H111 stone 2 The licensee has identified 3/23/79 4/6/79 Complete - the limit switches for 26 valves with stem-mounted all but four valves have been limit switches that they say replaced. The balance will be "may not be suitable for replaced at the first shutdown service in the LOCA environment."
after 9/15/79. The basis for The function of these switches continued operation in the interh was not identified.
is in the SER for License Amend.
No. 52 dated S/12/79.
9.
Browns Ferry Hotor operators for main 3/13/79 3/27/79 Complete - all motors will be Units 1/2/3 steam line drain and RilR rewound with the proper class head spray isolation valves.
insulation. Unit 2 is complete.
Unit 3 will be done by September 1979 and Unit I will be done by December 1979. In the interim the valves will be lockea in the safety position (closed).
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DATE OF 24 hr DATE OF 14 DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF D0R REVIEW
- 13. Surry I ASCO solenoid pilot valve model 5/23/79 N/A Complete - the unqualified No. 8320A102 used for inside valves will be replaced before containment isolation valves on restarting from the current steam generator blowdown lines, outage.
- 14. Brunswick, AMP Special Industries control 5/25/79 Verb.
6/6/79 Ccmplete
'icensee has verified Units 1/2 and instrumentation cable 5/29/79 Letter that termination will not fall i
terminal lugs.
even if the lugs becoam totally uninsulated in an accident en0tronment. Continued operation is justified on this basis. A fully qualified rethod of insulat-ing the terminations will be implemented at next refueling.
- 15. Hatch 1 Stem nounted limit switches 6/13/79 Verb.
6/18/79 LER on MSIV providing input to
- 79-037-01-T-0 Complete - LER clarified app 11-the RPS.
cation of the switches. They do not provide input to the RPS, only position indication.
Switches will be replaced at the next scheduled ccId shutdown.
The containment isolation function is assured in the interim by the-fall safe design ar.d a redundant valve outside.
- 16. Dresden, Limit switches on testable LER 79-9/01T-0 N/A Complete - the unqualified Units 2/3 check valves in the core 5/24/79 switches are only associated with the test actuator. No failure spray system. NAMC0 SL2-C-11.
mode of the switches could impair the safety function of the check Cys valve. No corrective measures are required.
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DATE OF 24 hr DATE OF le. DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF D0R REVIEW
- 17. Calvert Cilffs. ASCO pilot solenoid valves for 6/13/79 Letter N/A Complete - licensee states that the unqualified valves will be Units 1/2 air operated containment purge '
replaced and that in the interim isolation valves.
continued operation is justified because power leads to the unqual-ified valves have been removed and containment purge will be prohibited during power operation.
Further, the outside containment purge isolation valve is not ex% sed to an accident environment Unit 2 valves will be replaced before restart following fall 1979 outage. Unit I valves will be replaced during the Spring 1980 outage. See also item 22.
- 18. Geonee, Aluminum limit switch housings 5/23/79 Verb.
6/5/79 Complete - the aluminum limt Units 1/2 on 8 containment isolation switch housings will be replaced valves for Unit I and 6 valves at the next cold. shutdown for for Unit 2.
Systems involved each unit. Continued operation included: letdown cooler A and is justified in the interim B, reactor coolant pump seal because: 1) the limit switches return, normal decay heit did function properly for the removal, quench tank vent, first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a qualification component cooling water test and they perform their return.
containment isolation function within 45 sec. of a DBA; 2) the valve could be repaired during the post accident period if necessary cys by making mods. outside contain-
~~J ment; and 3) redundant containment isolation valves are located c:)
outside containment.
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DATE OF 24 hr DATE DE 14 DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF DOR REVIEW
- 22. Calvert Cliffs. ASCO pilot solenoid valves for 6/ 29/ 79 N/A Complete - the licensee will Units 1/2 the air operators on the replace all non-qualified valves, Unit 2 Fall 1979, Unit 1 Spring following valves: Aux. spray, (see also item IT Si tanks fill and drainline, 1980. Continued operation is above)
SI tanks check valve leak, justified in the interim because:
SI recirc. return line drain,
- 1) the containnent purge isolation and containment purge. Also valves will be maintained closed ASCO solenoid valve on in Mode 1, 2, 3 and 4 and Containment Instrument air electrical power and air to the
- header, valyes will be locked out (licensee will confirm this in his.14 day response). Further, the isolation valve outside containment will be closed; 2) for aux. spray a redundant path for vessel flushing for boren precipitation is available; 3) the SI tank valves are not containment isolation valves. All valves feed a common header which has manually ophated normally closed isolation valves outside containment; 4) the Instru. air header valve is used only for separation of non-seismic and seismic section of piping. All valves supplied by seismic piping fall in safe condition.
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DATE OF 24 hr DATE OF 14 DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF D0R REVIEW
- 23. Crystal River, Containment isolation valve 6/17/79 LER #79-062/01T Complete - the motor operators will Unit 3 electric motor operatocafor
-0 be replaced by Jan. 31, 1980.
pressurizer steam, pressurizer 7/17/79 Continued operation in the interin water, and two steam generator is justified on the following sample valves (CAV-1, 3, 4, 5).
basis: 1) valves are normally The Ilmit switch gear frame and closed; 2) redundant valves located outside containment; 3) covers are nude of Al and notor dose to date is less than 1500 cagles are only qualified for Rad; and 4) the containment 10 rads.
isolation function is complete within 1 min of an isolation signal and the licensee concludes that the valve will function for at least this length of time.
- 24. Beaver Valley
- Inside containment the following 6/14/79 Complete - all unqualified Nuclear Power did not have environmental 7/24/79 equipment will be replaced by the Station qual. docs.: 1) ASCO solenoid 7/30/79 next refueling outage. The level valves (44 in number ); 2) 7/31/79 transmitters will be mounted at Meter device tenninal blocks; 8/1/79 a higher elevation. The bases and 3) Interconnecting wires 8/28/79 provided by the licensees for (Continental Co. SIS type).
continued operation for 6-7 weeks The 4 level transmitters asso-were found acceptable. (See ciated with the steam generator minutes of meeting d=3s4 t/s,p3),
were not quallfled for submer-gence.
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DATE OF 24 hr DATE OF 14 DAY PLANT NAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF DDR REVIEW 25.
Indian Point, ASCO solenoids - 43 solenoids 8/1/79 Complete - by letter dated Unit 2 inside containment (info.
7/18/79 8/30/79 the licensee submitted obtained by telephone call 8/30/?9 list of valves and bases for with licensee).
continued operation. Unit shutdown for refueling will start-up during second week of September 1979. Bases for continued operation:
a) Jnstrument air to SV's is isolated in response to a SI signal allowing valves to go to their safe position, b) Failure mode (electrical) will cause the valve to go to safe position.
c) Degradation of internals would not allow enough air pressure to build in the actuator to open the-valve.
Os The licensee has sa stated that some of the SV's will be replaced prior to start-up.
ca The remaining will be replaced s4) during the next refueling outage.
DATE OF 24 hr DATE OF 14 DAY PLANT HAME UNQUALIFIED EQUIPMENT REPORTED NOTIFICATION DETAILED REPORT STATUS OF D0R REVIEW
- 26. Pilgrim 1 ASCO Solenoid Valves associated 8/28/79 8/28/79 Letter Complete - the unqualified valves with drywell flow and equipment LER 79-032/
79-171 will be replaced at the next drain sump effluent line isola-OIT-0 9/11/79 LER scheduled plant outage following tion valves. These valves are followup receipt of the replacenent valves located outside primary contain-79-032/01T-0 on order. Continued plant opera-ment in the torus conpartnent.
tion is justified by the licensee i
on the basis that alternate cont-rois and isolation equipment are readily accessible to adequately isolate drywell sump effluent lind,s. By telecon (confirmed by the 14 day report from licensee dated 9/11/79) the licensee has
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stated that the emergency operating procedures will be revised to remove instrument air from the SV's associated with one of the efflu-ent lines, and to manually isolate the second effluent line by using valve in Raddaste Building.
27.
D.C. Cook 1/2 ASCO solenoid in each unit (Model 9/7/79 Complete - the ASCO valves will No. 831654V) in Containnent isola-be replaced during first outage of tion valves in the following sufficient duration following sys tens : 1) Ice condenser refrig.
receipt of SV's.
Continued opera-supply and return; 2) Upper contain-tion is justified in the following ment purge supply & exhaust; 3) bases: 1) each valve has an associ Iower containnent purge supply &
ated redundant containment isola-exhaust; and 4) Instrur.ent room tion valve located outside contain-c7s ventilation supply & exhaust.
nent; 2) based on analysis the SV's ASCO valve contains Buna N are capable of operation to expo-
'd elastomer and acetal materials.
sures of a c..kawb..s of 400,000 Rad and 180*F. Based on the time C23 of initiation of the isolation
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valves, analysis has shown that the valves see less than 44,000 Rad - thus the ability of the SV's to perform their function under LOCA will not be impaired by radia-
- i. ion. For the ifalting temperature
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LP 92
ENCLOSURE 2 GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EQUIPMENT IN OPERATING REACTORS 1.0 Introduction 2.0 Discussion 3.0 Identificatiow of Class IE Equipment 4.0 Service Conditions 4.1 Service Conditions Inside Containment for a loss of Coolant Accident LOCA 1.
Temperature and Pressure Steam conditions 2.
Radiation J.
Submeroence 4.
Chemical Sorays 4.2 Service Conditions for a PWR Main Stean Line Break (MSLB)
Inside Containment 1.
Temoerature and Pressure Steam Conditions 2.
Radiation 3.
Subrcergence 4
Chemical Sorays 4.3 Service Conditions Outside Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a H1gn Enerov Line Break (HEL3 4.3.2 Areas Where Fluids are Recirculated From Inside Containment to Accomollsn Lono-Term Emercency Core Coolino Followino a LOCA 1.
Temoerature, Pressure and Relative Humidity 2.
Radiation 3.
Subm,'rcence 1617 09Ei 4
Chemical Sorays
. 4.3.3 Areas Normally Mat tained at Room Conditions 5.0 Qualification Methods 5.1 Selection of Qualification Method 5.2 Qualification by Tyoe Testing 1.
Simulated Service Conditions and Test Duration 2.
Test Soecimen 3.
Test Secuence 4.
Test Specimen Aging 5.
Functional Testing and Failure Criteria 6.
Installation Interfaces 5.3 Oualification by a Combination of Methods (Test, Evaluation, Analysis 6.0 Marcin 7.0 Aoina 8.0 Documentation Appendix A - Typical Equipment / Functions Needed for Mitigation' of a LOCA or MSLB Accident Appendix B - Guidelines for Evaluating Radiation Service Conditions Inside Containment for a LOCA and MSLB Accident Appendix C - Thermal and Radiation Aging Degradation of Selected Materials 1617 096
GUIDELINES FOR EVALUATING ENVIRONMENTAL QUALIFICATION OF CLASS IE ELECTRICAL EOUIPNENT IN OPERATING REACTOR (
1.0 INTRODUCTION
On February 8, 1979, the NRC Office of Inspection and Enforcement issued IE Bulletin 79-01, entitled, " Environmental Qualification of Class IE Equipment." This bulletin requested that licensees for operating power reactors complete within 120 days their reviews of equipment qualification begun earlier in connection with IE Circular 78-08. The objective of IE Circular 78-08 was to initiate a review by the licensees to determine whether proper documentation existed to verify that all Class IE electrical equipment would function as required in the hostfie environment which could result from design basis events.
The licensees' reviews are now essentially complete and the NRC staff has begun to evaluate the results. This document sets forth guidelines for the NRC staff to use in its evaluations of the licensees' responses'to IE Sulletin 79-01 and selected associated qualification documentation. The objective of the evaluations using these guidelines is to identify Class IE equipment whose documentation does not provide reasonable assurance of environ-mental qualification. All such equipment identified will then be subjected to a plant application soecific evaluation to determine whether it should be requalified or reolaced with a component whose cualification has been adequately verified.
These guicelines are intended to be used by the NRC staff to evalaa:e the qualifica 4cn metnods used for existing equipment in a parcicular class of plants, i.e., currently operating reactors including SEP plants.
1617 097
. Equipma'it in other classes of plants not yet licensed to operate, o-replacement equipment for operating reactors, may be subject to different requirements such as those set forth in NUREG-0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment.
In addition to its reviews in connection with IE Bulletin 79-01 the staff is engaged in other generic reviews that include aspects of the equipment qualification issue. INI-2 lessons learned and the effects of failures of non-Class IE control and indication equipment are examples of these generic reviews. In seme cases these guidelines may be applicable, however, this detennination will be made as part of that related generic review.
2.0 DISCUSSION 1
IEEE Std. 323-1974 is the current industry standard for environmental qualification of safety-related electrical equipment. This standard was first issued as a trail use standard, IEEE Std. 323-1971, in 19~1 and later after substantial revision, the current version was issued in 1974 Both versions of the standard set forth generic requirements for equipment quali-fication but the 1974 standard includes specific requirements for aging, margins, and maintaining documentation records that were not included in the 1971 trial use standard.
The intent of this document is not to provide guidelines for implementing either version of IEEE Std. 323 for operating reactors.
In fact most of the operating reactors are not comitted to comply with any particular industry standard for electrical equipment qualification. However, all of the operating r'eactors are required to comply with tne General Design Criteria I IEEE Std. 323-1974, "IEEE Standard for Qualifying Class IE Eouipment for Nuclear Power Generating Stations."
I6L7 098
3-specified in Appendix A of 10 CFR 50. General Design Criterion 4 states in part that " structures, systems and components important to safety shall be designed to accomodate the arfects of and to be compatible with the environmental conditions associated with nomal operation, maintenance, testing and postulate'd accidents, including loss-of-coolant accidents."
The intent of these guidelines is to provide a basis for judgements required to confirm that operating r'eactors are in compliance with General Design Criterion 4.
3.0 IDENTIFICATION OF CLASS IE EOUIPMENT Class IE equipment includes all electrical equipment needed to achieve emergency reactor shutdown, containment isolation, reactor core cooling, containment and reactor heat removal, and prevention of significant release of radioactive material to the environment, Typical systes included in pressurized and boiling water reactor designs to perfom these functions for the most severe postulated lon of coolant accident (LOCA) and main steamline break accident (MSLS) are listed in Appendix A.
More detailed descriptions of the Class IE equipment installed at specific plants can be obtained from FSARs, Technica! specifications, and emergency procedures, Although variation in nomenclature may exist at the various plants, environmental qualification of those systems which perfom the functions identified in Appendix A should be evaluated against the appropriate service conditions (Section 4.0).
The guidelines in this documer.t are applicable to all components necessary for operation of the systems listed in Appendix A including but not limited to valves, motors, cables, connectors, relays, switches, transmitters and valve position indicators, 1617 099
~
4-4.0 SERVICE CONDITION 3 In order to deter mine the adequacy of the' qualification of equipment it is necessary to specify the environment the equipment is exposed to during normal and accident conditions with a requirement to remain functional, These environments ar,e referred to as the " service conditions."
The approved service conditions specified in the FSAR or other licensee submittals are acceptable, unless otherwise noted in the guidelines discussued below.
4.1 Service Conditions Inside Containment for a loss of Coolant Accident (LOCA) 1.
Temperature and Pressure Steam Conditions. In general, the containment temperature and pressure conditit.4s as a function of time should be based on the analyses in the FSAR. In the specific case of pressure suppression type containments, the follcwing. minimum high tempeature conditions should be used: (1) BWR Drywells - 3400F for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; and (21 PWR Ice Condenser Lower Compartments 3400F for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, 2.
Radiation - When specifying radiation service conditions fo'r equipment exposed to radiation during nonnal operating and accident conditions, the nonnal operating dose should be added to the dose received during the course of an accident. Guidelines for evaluating beta and garmna radiation service conditions for general areas inside containment are provided below, Radiation service conditions for equipment located directly above the containment sump, in the vicinity of filters, or sutrnerged in contaminated liquids must be evaluated on a case by case basis, Guidelines for these evaluations are not provided in this
- document, 1617 100
.5 Gama Radiation Doses - A total gama dose radiation service condition 7
of 2 x 10 RADS is acceptable for Class IE equipm..it located in general areas inside containment for PWRs with dry type containments, Where a dose less than this value has been specified, an application specific s
evaluation must be perfomed to determint if the dose specified is acceptable. Procedures for evaluating radiation service conditions in such cases are provided in Appendix B.
The procedures in Appendix B are based on the calculation for a typical PWR reported in Appendix 1
0 of NUREG4588.
Gama dose radiation service conditions for BWRs and PWRs with ice condenser containments must be evaluated on a case by case basis, Since the precedures in Appendix B are based on a calculation for a typical PWR with a dry type containment, they are not directly applicable to BWRs and other containment types. However, doses for these other plant configurations may be evaluated using similar pmcedures with conservative dose assumptions and adjustment factors developed on a case by case basis s Beta Radiation Doses - Beta radiation doses generally are less significant than gama radiation doses for equipment qualification. This is due to the icw penetrating power of beta particles in comparison to gama rays of equivalent energy, Of the general classes of electrical equipment in a plant (e.g., cables, instrument transmitters, valve coerators, contairment penetrations), electrical cable is considered the most I NUREG-0588, Interim Staff Positier. on Environmental Qualification of Safety Related Electrical Equipment.
1617 101
6-vulnerable to damage from beta radiation. Assuming a TID 14844 source term, the average maximum beta energy and isotopic abundance will vary as a function of time following an accident. If these parameters are considered in a detailed calculation, the conservative 8
beta surface dose of 1.40 x x 10 RADS reported in Appendix 0 of NUREG 0588 would be reduced by approximately a factor of ten within 30 mils of the surface of electrical cable insulation of unit density. An additional 40 mils of insulation (total of 70 mils) results in another factor of 10 reduction in dose. Any structures or other equipment in the vicinity of the equipment of interest would act as shielding to further reduce beta doses.
If it can be shown, by assuming a conserva-8 RADS and considering tive unshielded surface beta dose of 2.0 x 10 the shielding factors discussed here, that the beta dose to radiation sensitive equipment internals would be less than or equal to 10% of the total gamma dose to which an item of equipment has been qualified, then that equipment may be considered qualified for the total radiation environment (gamma plus beta). If this criterion is not satisfied the radiation service condition should be detennined by the sum of the ganna and beta doses.
3.
Submercence - The' preferred method of protection against the effects of submergency is to locate equipment above the water flooding level.
Specifying saturated steam as a service condition during type testing of equipment -that will become flooded in service is not an acceptable alternative for actually flooding the equipment during the test.
1617 102
Containment Sorays - Equipment exposed to chemical sprays should be 4.
qualified for the most severe chemical environment {actdic or
~
basic) which could exist. Demineraliz'ed water sprays should nec be exempt from consideration as a potentially adverse service condition.
4.2 Service Conditions for a PWR Main Steam Line Break (MSLB) Inside Containment
~
Equipment required to function in a steam line break environment must be qualified for the high temperature and pressure that could result.
In same cases the environmental stress on exposed equipment may be higher than that resulting from a LOCA, in others it may be no more severe than for a LOCA due to the automatic operation of a containment spray system.
1.
Temeerature and Pressure Steam conditions - Equipment qualified for a LOCA environment is considered qualified for a MSLB accident environ-ment in plants with automatic spray systems not subject to disabling single component failures. This position is based on the "Best Estimate" calculation of a typical plant peak temperature and pressure and a thermal analysis of typical components inside containment.E The final acceptability of this approach, i.e., use of the "Best Estimate",
is pending the completion of Task Action Plan A-21, Main Steamline Break Inside Containment.
Class IE equipment installed in plants without automatic spray systems or plants with gpray systems subject to disabling single failures or delayed initiation should be qualified for a MSLB accident environment determined by a plant specific analysis. Accaptable methods I See NUDEG 0458, Short Tem Safety Assessment on the Environmental Qualification of Safety-Related Electrical Eauipment of SEP Operating Reactors, for a more detailed discussion of the best estimate calculation.
1617 103
for performing such an analysis for operating reactors are provided in Section 1.2 for Category II plants in NUREG-0538. Interim Staff Position on Environmental Qualification of Safety-Related Electrical
~
Equipment.
2.
Radiation - Same as Section 4.1 above except that a conservative 6
gamma dose of 2 x 10 RADS is acceptable.
3.
Submeroence - Same as Section 4.1 above.
4.
Chemical Sprays - Same as Section 4.1 above.
4.3 Service Conditions Outside of Containment 4.3.1 Areas Subject to a Severe Environment as a Result of a Hioh Eneroy Line Break (HELB)
Service conditions for areas outside containment exposed to a HELB were 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 HELB. The equipment required to mitigate the event was also identified..This equipment should be qualified for the service conditions reviewed and approved 'n t9e +i.S Sa'ety Evaluation Report for each scecific plant.
4.3.2 Areas Where Fluids are Recirculated from Inside Centainment to Accomolish Leno-Term Core Coolino Followino a LOCA 1.
Temeerature and Relative Humidity - One hundred cercent relative humidity shouTd be established as a service condition in confined spaces. The temperature and pressure as a function of time should be based on the plant unique analysis reported in the FSAR.
bh hOk
.g.
2.
Radiation - Due to differences in equipment arrangement within these areas and the significant effect of this factor on doses, radiation service conditions must be evaluated on a case by case 6 RADS would be basis. In general, a dose of at least 4 x 10 expected.
3.
Submergence - Not applicable.
4.
Chemical Sorays - Not applicable.
4.3.3 Areas Normally Maintained at Room Conditions Class IE equipment located in these areas does not experience significant stress due to a change in service conditions during a design basis event.
This equipment was designed and installed using standard engineering practices and industry codes and standards (e.g., ANSI, NEMA, National
. Electric Code). Based on these factors, failures of equipment in these areas during a design basis event are expected to be random except to the extent that they may be due to aging or failures of air conditioning or ventilation systems..Therefore, no special consideration need be given to the environmental qualification of Class IE equipment in these areas provided the aging requirements discussed in Section 7.0 below are satisfied and the areas are maintained at room conditions by redundant air conditioning or ventilation systems served by the onsite emergency electrical power system.
Equipment located in areas not served by redundant systems powered from onsite emeroency sources should be qualified for the environmental extremes which could result from a failure of the systems as determined from a plant specific analysis.
5.0 OUAt.IFICATION WE HCDS 16J7 105
=
5.1 Selection of Qualification Method The choice of qualification mathod employed for a particular application of equipment is largely a matter of technical judgement based on such factors as: (1) the severity of the service conditions; (2) the structural and material complexity of the equipment; and (3) the degree of certainty required in the qualification procedure (i.e., the safety importance of the equipment function). Based on these considerations, type testing is the preferred method of qualification for electrical equipment located inside containment required to mitigate the consequences of design basis events, i.e., Class IE equipment (see Section 3.0 above). As a minimum, the qualification for severe temperature, pressure, and steam service conditions for Class IE equipment should be based on type testing.
Qualification for other service conditions such as radiation and chemical sprays may be by analysis (evaluation) supported by test data (see Section 5.3 below). Exceptions to these general guidelines must be justified on a
~
case by case basis.
5.2 Oualification by Tyee Testina The evaluation of test plans and results should include consideration of the following factors:
1.
Simuiated Service Conditions and Test Duration - The environment in the test chamber should be established and maintained so that it envelopes the service conditions defined in accordance with Section 4.0 above.
The time duration of the test should be at least as long as the period from the initiation of the accident until the temperature and pressure service ccnditions return to essentially the same levels that existed before the postulated accident. A shorter test duration may be acceptable 1617 106
. if specific analyses are provided to demonstrate that the materials involved v 11 not experie.sce significant accelerated trannal aging during the period not tested.
2.
Test Specimen - The test specimen should be the same model as the equipment being qutlified. The type test shoul'd only be considered valid for equipment identical in design and material construction to the test specimen. Any deviations should be evaluated as part of the qualifica-tion documentation (see also Section 8.0 below).
3 Test Sequence - The component being tested should be exposed to a steam / air environment at elevated temperature, and pressure in the sequence defined for its service conditions. Where radiation is a service condition which is to be considered as part of a type test, it may be applied at any time during the test sequence provided the component does not contain any materials which are known to be susceptible to significant radiation damage at the service condition levels or materials whose susceptibility to radiation damage is not known (see Appendix C).
If the component contains any such materials, the radiation dose should be acclied prior to or concurrent with exposure to the elevated temperature and pressure steam / air environment. The same test specimen should be used throughou' the test sequence for all service conditions the equipment is to be qualified for by type testing. The type test should only be considered valid for the service conditions applied to the same test specimen in the appropriate sequence.
4.
Test Sceciren Acino - Tests which were successful using test specimens whicn had not been creased may be considered acceptable crovided the component dces not contain materials which are known to be susceatible 1617 107
. to significant degradation due to thermal and radiation agin (see Section 7.0).
If the component contains such materials a qualified life for the component must be established on a case by case basis. Arrhenius techniques are generally considered acceptable for thermal aging.
5.
Functional Testino and Failure Criteria - Operational modes tested should be representative of the actual application requirements (e.g., components which operate normally energized in the plant should be nonna11y energized during the tests, motor and electrical cable loading during the test should be representative of actual operating conditions). Failure criteria should include instrument accuracy requirements based on the maximum error assumed in the plant safety analyses.
If a component fails at any time during the test, even in a so called " fail safe" mode, the test should be considered inconclusive with regard to demonstrating the ability of the component to function for the e-tire period prior to the failure.
6.
Installation Interfaces - The couipment mounting and electrical or mechanical seals used during the type test should be representative of the actual installation for the test to be considered conclusive.
The equipment qualification program should include an as-built inspection in the field to verify that equipment was installed as it was tested. Particular emphasis should be placed on comon pmblems such as pmtective enclosures installed upside down with drain holes at the top and penetrations in equipment housings for electrical connections being left unseated or susceptible to moisture incursion through stranded conductors.
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. 5.3 Qualification by a Combination of Methods (Test, Evaluation, Analysis As discussed in Section 5.1 above, an item of Class IE equipment may be shown to be qualified for a complete spectrun of service conditions even though it was only type tested for high temperature, pressure and steam. The qualification for service conditions such as radiation and chemical sprays may be demonstrated by analysis (evaluation).
In such cases the overall qualification is said to be by a combination of methods. Following are two specific examples of procedures that are considered acceptable. Other similar procedures may also be reviewed and found acceptable on a case by case basis.
1.
Radiation Oualification - Some rf the enrifer tvoa tests perfomed for operating reactors did not include radiation as a service condition. In these cases the equipment may be shown to be radiation qualified by perfoming a calculation of the dose expected, taking into account the time the equipment'is required to remain functional and its location using the methods described in Appendix B, and analyzing the effect of the calculated dose on the materials used in the equipment (see Appendix C). As a general rule, the time required to remain functional assur.1ed for dose calculations should be at least I hour.
2.
Chemical Sorav Oualification - Components enclosed entirely in corrosion resistant cases (e.g., stainless steel) may be shown to be qualified for a chemical environment by an analysis of the effects of the particular che.f cals on che particular enclo-sure materials. The effects of chemical sprays on the pressure integrity of any gaskets or seals present should be considered in the analysis.
1617 109
- 14 6.0 Marcin IEEE Scd. 323-1974 e ines margin as the difference between the most severe specified service conditions of the plant and the conditions used in type testing to account for nonnal variations in comercial production of equipment and reasonable errors in defining satisfactory performance.
~
Section 6.3.1.5 of the standard provides suggested factors to be applied to the service conditions to assure adequate margins. The factor applied to the time equipment is required to remain functional is the most significant in tenns of the additional confidence in qualification that is achieved by adding margins to service corditions when establishing test environments. For this reason, special consideration was given to the time required to remain functional when the guidelines for Functional Testing and Failure Criteria in Section 5.2 above were established. In addition, all of the guidelines in Section 4.0 for establishing service conditions include conservatisms which assure margins betweerf the service conditions specified and the actual conditions which could realistically be expected in a design basis event. Therefore, if the guidelines in Section 4.0 and 5.2 are satisfied no separate margin factors are required 3
to be addid to the service conditions when specifying test conditions.
7.0 A91n9 Implicit in the staff position in Regulatory Guide 1.89 with regard to backfitting IEEE Std. 323-1974 is the staff's conclusion that the incremental improvement in safety from arbitrarily requiring that a specific qualified life be demonstrated for all Class IE equipment is net sufficient to justify the expense for plants already constructed and ocerating. This position dces not, however, exclude equiement 1617 110
. using materials that have been identified as being susceptible to significant degradation due to thermal and radiation aging. Component maintenance or replacement schedules should include consideratiors of the specific aging characteristics of the component materials. Cegoing programs should exist at the plant to review surveillance and maintenance records to assure that equipment which is exhibiting age related degrada-tion will be identified and replaced as necessary. Appendix C contains a listing of materials which may be found in nuclear power plants along with an indication of the material susceptability to themal ar" radiation aging.
8.0 Documentation Complete and auditable records must be available for qu:lification by any of the methods described in Section 5.0 above to be considered valid.
These records should describe the qualification method in sufficient detail to verify that all of the guidelines have been satisfied. A simple vendor certification of compliance with a design specification should not be considered adequate.
1617 111
APPENDIX _A
, TYPICAL EQUIPMENT / FUNCTIONS NEEDED FOR
~
MITIGATION OF A LOCA OR_MSLB ACCIDENT _
Engineered Safeguards Actuation Reactor Protection Containment Isolation Steamline Isolation Main Feedwater Shutdown and Isolation Emergency Power I
Emergency Core Cooling Containment Heat Removal Containment Fission Product Removal Centainment Combustible Gas Control Auxiliary Feedwater Containment Ventilation Containment Radiation Monitoring Control Room Habitability Systems (e.g., HVAC, Radiation Filters)
Ventilation for Areas Containing Safety Equipment Component cooling Service Water 2
Emergency Shutdown 3
Post Accident Sampling and Monitoring 3
Radiation Monitoring 3
Safety Related Display Instrumentation 1617 112
2-I These systems will differ for PWRs and FWRs, and for older and newer plants.
In each case the system features which allow fo transfer to recirculation cooling mode and establishment of long term cooling with boron precipitation control are to be considered as part of the system to be evaluated.
2Emergency shutdown systems include those 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. Examples of such systems and equipment are the RHR system, PORVs, RCIC, pressurizer sprays, chemical and volume control system, and steam dump systems.
3More specific identification of these types of equipment can be found in the plant emergency procedures.
I617 113 e
APPENDIX B PROCEDURES FOR EVALUATING GA!HA RADIATION SERVICE CONDPTIONS Introduction and Discussion The adequacy of gama radiation service conditions specified for inside containment during a LOCA or MSLB accident can be verified by assuming a conservative dofe at' the containment centerline and adjusting the dose
~
according the plant specific parameters, The purpose of this appendix is to identify those parameters whose effect cn the total gama dose is easy to quantify with a high degree of confidence and describe procedures which may be used to take these effects into consideration.
The bases for the procedures and restrictions for their use are as follows:
7 RADS (11 A conservative dose at the containment centerline of 2 x 10 for a LOCA and 2 x 106 RADS for a MSLB accident has been assumed.
This assumption and all the dose rates used in the procedure out-lined belo,. are Based on the methods and sample calculation described in Appendix D of NUREG.0588, " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equip-ment." Therefore, all the limitations listed in Appendix D of NUREG-0588 apply to these procedures.
(2) The sample calculation in Appendix 0 of NUREG 0588 is for a 4,000 6 fg3 contain-MWth pressurized water reactor housed in a 2.52 x 10 ment with an i' dine scrubbing spray system. A similar calculation o
without iodine scrubbing sprays would increase the dose to equipment 7
approximately 15"..
The conservative dose of 2 x 10 RADS assumed 1617 114
, in the n ocedure below includes sufficient conservatism to I
account for this factor. Therefore, the pro.edure is also applicable to plants without an iodine scrubbing spray system.
(3) Shielding calculations are based on an average gama energy of 1 MEV derivect from TID 14844 (4) These procedures are not applicable to equipment located directly above the containment sump, submerged in contaminated liquids, or near filters. Doses specified for equipment located in these areas must be evaluated on a case by case basis.
(5) Since the dose adjustment factors used in these procedures are based on a calculation fcr a typical pressurized water reactor with a dry type containment, they are not directly applicable to boiling water reactors or other cuntainment types. However, doses for these other plant configurations may t,a evaluated using similar procedures with conservative dose assumptions and adjustment factors developed on a case by case basis.
Procedure Figures 1 through 4 provide factors to be applied to the conservative dose to correct the dose for the following plant specific parameters:
(1) reactor power level; (2) containment volume; (3) shielding; (4) compartment volume; and (5) time equipment is required to remain functional.
1617 115
3 The procedure for using the figures is best illustrated by an example.
Consider the following case., The radiation service condition for a 6
particular item of equipment has been specified as 2 x 10 RADS. The application specific parameters are:
~
Reacted power level - 3,000 MWth 3
Containment volume - 2.5 x 106 ft Compartment Volume - 8,000 ft3 Thickness of compartment shield wall (concrete) - 24" Time equipment is required to remain functional - 1 hr.
The problem is to make a reasonable estimate of the dose that the equipment could be expected to receive in order to evaluate the adequacy of the radiation service condition specification.
Steo 1 Enter the nomogram in Figure 1 at 3,000 MWth reactor power level and 6
2.5 x 10 ft3 containment volume and read a 30-day integratid dose of 7
1.5 x 10 RADS.
Steo 2 7
Enter Figure 2 at a dose of 1.5 x 10 RA05 and 24" of concrete shielding 4
for the compartment the equipment is located in and read 4.5 x 10 RADS.
This is the dose the equipment receives from sources outside the compart-ment. To this must be added the dose from sources inside the compartment (Step 3).
Steo 3 3
Enter Figure 3 at 8,000 ft and read a correction factor of 0.13.
The 7
dose due to sources inside the compartment would then be 0.13 (1.5 x 10 )
6 RADS. The sums of the doses from steps 2 and 3 equals:
= 1.95 x 10 7
6 RADS 4 RADS + 0.13 (1.5 x 10 ) RADS = 2.0 x 10 4.5 x 10 1617 116
. Steo 4 Enter Figure 4 at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and read a correction factor of 0.15. Apply this factor to the sum of the doses detennined from steps 2 and 3 to correct the 30 day total dose to the equipment inside the compart:nent to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
6 5 RADS 0.15 (2.0 x 10 ) = 3 x 10 6
In this particular example the service condition of 2 x 10 RADS specified is conservative with respect to the estimated dose of 3 x 5
10 RADS calculated in steps 1 through 4 and is, therefore, acceptable.
1617 117
FIGURE 1 NOMOGRAM FOR CONTAINMENT VOLUME AND REACTOR POWER LOCA DOSE CORRECTIONS
- CONTAINMENT -
VOLUME (ft3) 3 x 108 2 x 10s 30 DAY Mwm INTEGRATED yDOSE 4000 1 x 108 7
3000 4 x 10 2000 3 x 107 1000 5 x 108 7
4 x 108 2 x 10 3 x 108 200 1 x 107 2 x 108 1 x 108 5 x los 4 x 108 3 x 108 2.5 x 10s 2.0 x 108 1 x 108 i
- MSL3 ACCIDENT DOSES SHOULD BE READ AS A FACTOR OF 10 LESS l
1617 118
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APPENDIX C THERMAL AND RADIATION AGING DEGRADATION OF SELECTED MATERIALS Table C-1 is a partial list of materials which may be found in a nuclear power plant along with an indication of the material susceptibility to
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radiation and thema'1 aging.
0 Susceptibility to significar!t thennal aging in a 45 C environment and nonnal atmosphere for 10 or 40 years is indicated by an (*) in the appro-priate column. Significant aging degradation is defined as that amount of degradation that would place in substantial doubt the ability of typical equipment using these materials to function in a hostile environment.
. Susceptibility to radiation damage is indicated by the dose level and the observed effect identified in the column headed BASIS. The meaning of the tems used to characterize the dose effect is as follows:
Threshold - Refers to damage threshold, which is the radiation e
exposure required to change at least one physical property of the material.
Percent Change of Property - Refers to the radiation exposure e
required to change the physical property noted by the percent.
e Allowable - Refers to the radiation which can be absorbed before serious degradation occurs.
The infomation in this appendix is based on a li.terature search of sources including the National Technical Infomation Service (NTI3), the National Aeronautics and Space Administration's Scientific and Technical Aerospace Report (STAR), NTIS Government Report Announcements and Index (GRA), and l617 122 i
. various manufacturers d.a reports. The materials list is not to be considered all int.iusive neither is it to be used as a basis for specifying materials to be used for specific applications within a nuclear plant. The list is solely intended for use by the NRC staff in making judgements as to the possibility of a particular material in a particular application being susceptible to significant degradation due to radiation or thermal' aging.
The data base for thennal and radiation aging in engineering materials is rapidly expanding at this time. As additional infonnation becomes available Table C-1 will be updated accordingly.
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