ML19241C194
| ML19241C194 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 06/11/1979 |
| From: | Stello V Office of Nuclear Reactor Regulation |
| To: | Frank L OREGON, STATE OF |
| Shared Package | |
| ML19241C195 | List: |
| References | |
| TAC-11299, NUDOCS 7907300121 | |
| Download: ML19241C194 (11) | |
Text
(4N J
n nec UNITED STATES 4} ).s.q y,%j NUCLEAR REGULATORY COMMISSION
- .47/.- l W ASHINGTO N, D. C. 20555 S. t4Wa/}~
%' % 9 JUN 111973 Docket No. 50-344 Mr. Lynn Frank, Director Oregon Department,of Energy Labor and Industries Building Room 111 Salem, Oregon 97310
Dear Mr. Frank:
This is in response to the Oregon Department of Energy's letter of March 1, 1979, wherein several questions were raised concerning generic safety issues and the seismic, environmental, and radiation qualification of electrical equipment at Trojan Nuclear Plant.
Taking the questions in the order that you posed them, the following five Westinghouse Topical Reports were referenced in the Trojan FSAR and contained infomation intended to support the issuance of the cperating license for Trojan:
WCAP-7821, Seismic Testing of Electrical and Control Equipment (High Seismic Plants), December 1971 WCAP-7744, Environmental Testing of Engineered Safety Features Related Equipment (NSSS - Standard Scope), August 1971 WCAP-7672, Solid State Logic Protection System Description, June 1971 WCAP-7819 Revision 1, Test Report, Nuclear Instrumentation System Isolation Amplifier, January 1972 WCAP-7705, Engineered Safeguards Final Device or Actuator Testing, March 1973 The present generic review status and the Trojan review status of these WCAPs is as folicws:
WCAP-7821 Generically, this report is acceptable provided that (1) justification of the Eagle Signal timer used is proviaed and (2) all output relays in the Solid-State Protection System for all high seismic plants are replaced with the qualified relays.
.A,6284 2907300\\M
Mr. Lynn Frank.
The Trojan design does not use a Westinghouse-supplied Eagle Signal timer. Therefore, proviso (1) is not applicable to Trojan.
PGE has replaced all output relays in the Solid-State Protection System with the qualified rotary-type relays. Therefore the Trojan design meets provis~o (2) and is acceptable.
WCAP 7744*
Generically, this report was found to be acceptable provided that Westinghouse supplied satisfactory answers to the folicwing four items:
(1) Verification that the deviations in accuracy and time of failure noted in the test results are within the specified time and accuracy required in the accident analysis for each specific plant.
(2)
Identification of those instruments inside containment required to follow the course of Condition III and IV**
events and verification of the capability of each instrument so identified, together with recommendations for a replacement instrument model for those not capable of long-tem monitoring.
(3) Westinghouse has indicated that additiona,1 instrumentation located outside of containment is available to the operator to follow the course of Condition III a :J IV events.
Westinghouse has been requested to identify this instrumentation and its capability for each specific event and plant referencing this WCAP.
(4) Confirmation that the differential pressure transmitters are temperature-compensated and that deviations are within that required for each specific application.
Westinghouse was requested to provide this infonnation in a letter dated January 15, 1979.
Their response is enclosed. Since Westinghouse did not supply answers to these questions and has withdrawn its reliance en WCAP-7410L, we have requested that PGE meet with us to provide answers to these questions as they relate to Trojan and discuss the qualification of eplacement transmitters which we understand they have ordered. We expect that this matter will be resolved at this meeting or shortly thereafter.
- N o n-p rop ri et a ry.
The proprietary version of.his docunent is designated WCAP-7410L.
- These are defined in Trojan FSAR, pc.15.3-1 and 15.4-1.
c,.
m 2SG
Mr. Lynn Frank Additional background regarding environmental qualification of electrical equipment is contained in the enclosed memorandum to the Commissioners dated March 15, 1979, and NUREG-0413, " Staff Report on the Environmental Qualif.ication of Safety-Related Electrical Equipment", February 1978.
WCAP-7672 and WCAP-7819 Generically, the staff has found these topical reports acceptable.
However, these reports do not address environmental and seismic quali-fication of the Solid-State Protection System and the nuclear instru-mentation system isolation amplifier.
These equipments have been seismically qualified under WCAP-7821 (see status of WCAP-7821 above).
In regard to environmental qualification, Westinghouse has documented that these equipments have been qualified to operate at 120 F and 95". humidity.
For Trojan, the Solid-State Protection System and nuclear instrumentation system cabinets are located in the control room.
The maximum temperature and humidity of the control room are 110 F and 80% humidity respectively.
These values are well within the qualified design limits for these equipments.
Therefore, these topical reports are acceptable for Trojan.
WCAP-7705 Generically, this report was fcund unacceptable.
Mcwever, this report is not applicable to the Trojan Plant. As stated in the Trojan Safety Evaluation Report (SER), Sect. ions 7.3.1 and 7.3.2 dated October 1974, the design of the ESF actuation system was found acceptable.
This conclusion was reached without reliance on Topical Report WCAP-7705,
" Engineered Safeguard Final Device or Actuator Testing", which was referenced in the FSAR but was found unacceptable by the staff as a basis for establishing confomance to safety criteria pertaining to ESF final actuator testability. However, our evaluatio of the ESF actuation system was based on our review of the Trojan r$AR, the schematic drawings of the circuitry used to initiate :;peration of ESF components, and on our prior review (on the D. C. Cook docket) of identical ESF actuation logic.
Therefore, the fact that WCAP-7705 was found unacceptable does nct constitute an unreviewed safety issue for Troj an.
For the balance-of-plant equipment, we find no indication of any Class 1E equipment that is not environmentally qualified for Trojan.
- Hcwever, PGE did identify uncualified electrical splices fcund on three pressurizer level transmitters (LER #78-027).
These splices were replaced with bliERS6
F 3
a Mr. Lynn Frank.
qualified components. This licensee event report was submitted in response to IE Circular 78-08 dated May 31, 1978, which directed all licensees to examine all installed safety-related electrical equipment and determine
~
that proper documentation exists which provides assurance that the equipment will function under postulated accident conditions.
In essence, the intent of that circular was to highlight to all licensees important lessons learned fraa environmental qualification deficiencies reported by individual licensees.
Further, by IE Bulletin No. 79-01, all licensees with operating power reactors including PGE, have been requested to expedite completion of the re-review program described in IE Circular 78-08.
This was done because inspections conducted by NRC with respect to responses to Circular 78-08 had identified components which licensees either have found to be unqualified for service within the LOCA environment or which do not have documentation of such qualification.
The effect of Bulletin 79-01 was to raise the threshold of IE Circular 78-08 to the level of a Bulletin, which would require licensee response.
Bulletin 79-01 in part requires licensees of all operating power reactor facilities to provide written evidence of the qualification of electrical equipment required to function under accident conditions and to complete the re-review program described in Circular 78-08.
PGE's response to this Bulletin is expected in June 1979.
This response should resolve any concerns regarding environmental qualification of all instrumentation and electrical equipment for Trojan.
With regard to compliance of the Trojan environmental qualification program with IEEE-323-1974 as endorsed and modified by Regulatory Guide 1.89, the acceptance criteria for the qualification of Class IE equip-ment for Trojan did not include the aging consideration specified in IEEE Standard 323-1974 and Regulatory Guide 1.89 (which endorses IEEE-323-1974).
In 1979, during the deliberations of the NRC's Regulatory Requirements Review Conmittee on the implementation of Regulatory Guide 1.89, con-sideration was given to the incremental improvements to safety it afforded in comparison of the then-current staff review practice. The Committee recommended that the guide be applied only to future CP applications; i.e., it should not be backfitted.
The decision was based on the Staff's judgment that the incremental imorovements were not significant to safety and that full implementation of IEEE-323-1974 required the further development of other ancillary standards to provide guidance on specific safety-related equipment and com enents.
Subsecuent public canments and review by the ACRS did not alter the recormenda: ion concerning implementation of Regulatory Guide !.39.
iAM7
Mr. Lynn Frank We recognize that additional guidance is needed in the area of accelerated aging techniques used to establish a qualified life for electrical equipment and assemblies. Our Category A technical activity on equipment qual,ification (Task Action Plan A-24) and an NRC extensive research program being carried out at Sandia Laboratories are intended to provide additional guidance for the development of test methods and licensing review procedures on aging. These programs will also allow us to make infomal judgments regarding the effects of aging.
In addition, as part of the Staff's Systematic Evaluation Program (SEP), the staff is assessing the surveillance and maintenance records for equipment inside and outside of containment of eleven selected older plants.
Since this equipment has been effectively " aged", the assessment of these records should provide additional infomation on the effects of aging.
Following completion of these ongoing activities -- the Task Action Plan A-24, the NRC research program, and the SEP effor* -- we will reconsider our position on the need for backfitting the aging requirements.
At that time, should we deem it necessary, we will take appropriate steps to ensure that aging effects are considered in assessing the adequacy of Class IE equipment used in the Trojan Plant.
It is our judgment that the natural aging that the Class IE equipment will undergo in the period to this reassessment will have little effect on its dnvironmental or seismic capability.
Regarding your question about the asymmetric blowdown load generic issue, acceptance criteria are currently being develc::ed as part of generic Task A-2, " Asymmetric Blowdown Loads on Reactor Primary Coolant System".
The program for resolution of this task (Task Action Plan) is contained in NUREG-0371, " Task Action Plans for Generic Activities".
Acceptance criteria are scheduled to be established by early summer. The NRC staff has developed guidelines for load combinations.
These are described in NUREG-0484, " Methodology for Combining Dynamic Responses".
We are currently evaluating break area and break opening time assumptions.
The Trojan analysis (PGE-1014) will be reviewed with respect to these generic criteria, and we expect that an evaluation for Trojan will be made later this summer. The basis for continued operation of affected plants is contained in Section 3 of the Task Action Plan.
M 268
Mr. Lynn Frank The fiRC staff recommendations for ATWS resolution are described in Volu. e n
3 of fiUREG-0460.
It is expected that a proposed rule on ATWS will be presented to the Commission for approval this summer after completion of ACRS review. We expect the rule to be effective early next The recommended modifications for Trojan class of plants year.
(operating Westinghouse plants) are changes necessary to provide A",,'S mitigating system actuation circuitry satisfying the criteria in A::pendix C of fiUREG-0460 Volume 3.
The rulemaking process would ultimately detemine the need for these plant changes and provide an implementation plan.
With respect to fire protection, PGE has proposed to provide a safe shutdown capability independent of the cable spreading room and control room. This shutdown capability does not involve a new system.
It consists of design modifications to existing safe shutdcwn systems which allow their use to safely shut down the plant even if cables now located in the control room or cable spreading room were lost in a fire. The independent safe shutdown capability was not a requirement of the fiRC; it was chosen by the licensee as an alternative to up-grading the existing fire suppression system in the cable spreading room and the control room to provide the level of defense which the staff considered acceptable for these areas to preserve shutdown capability.
The present fire protection is provided by (1) administrative controls over ignition sources and combustibles, (2J automatic sprinkler system (cable spreading roon), (3) snoke detection and manual fire suppression with hoses and portable extinguishers, and (4) physical separation and marinite barriers between divisions of redundant safe shutdown cabling.
For most fires, any one of these levels of defense is adequate to prevent loss of safe shutdown capability.
However, for added conservatism we postulate that both (1) and (2) fail, and therefore, the fire may become larger, in which case, (3) must be adequate to prevent loss of redundant safe shutdown equipment.
For Trojan, the staff had initially required (for the cable spreading room) a more sophisticated prompt-acting automatic system; i.e., directed water spray with open head at intermediate levels to cover all trcys and flame retardant coatings and fire barriers, and an automatic halon system for-control room cabinets containing redundant safe shutdown systems. As an alternative to upgrading O MS
Mr. Lynn Frank the sprinkler system the licensee proposed, in addition to the existing protection and additional detection in control room cabinets, to provide an alternate shutdown capability independent of these areas. The criteria for these existing systems as modified are the same as that prior to their modi fi cation.
The modifications to these existing systems are not to degrade the original design but simply to allow the flexibility of operation of at least one shutdown capability independent of damage in these areas.
In response to the second part of the request as to why interim operation is acceptable without the alternate capability, the following basis is provided.
In the report of the Special Review Group on the Browns Ferry Fire (NUREG-0050) dated February 1976, consideration of the safety of cperation of all operating nuclear power plants pending the completion of our detailed fire protection evaluation was presented.
The folicwing quotations from the report summarize the basis for the Special Review Group's conclusion that the operation of these facilities need not be restricted for public safety.
" Fires occur rather frequently, however, fires involving equipment unavailability comparable to the Brokns Ferry Fire are quite infrequent (see Secticn 3.3 of NUREG-0050).
The Review Group believes that steps already taken since March 1978 (see Section 3.3.2) have reduced this frequency significantly.
" Based on its review of the events transpiring before, during and after the Browns Ferry Fire, the Review Group concluded that the probability of disruptive fires of the magnitude of the Browns Ferry event is small and that there is no need to restrict operation of nuclear power plants for public safety. However, it is clear that much can and should be done to reduce even further the likelihood of disabling fires and to improve assurance of i
rapid extinguishment of fires that occur.
Consideration
[
should be given also to features that would increase further
~
the ability of nuclear facilities to withstand large fires without loss of important functions should such fires occur."
=
E6200
Mr. Lynn Frank We recognize that,the " Risk Assessment Review Group Report to the U. S.
Nuclear Regulatory Commission" NUREG/CR-0400 (The Lewis Committee Report) states that this Review Group is unconvinced of the correctness of the WASH-1400 conclusion that fires contribute negligibly to the overall risk of nuclear plant operation.
It is our conclusion that the operation of the Trojan facility, ?ending the implementation of all facility modifications including the alternate shutdown capability, does not present an undue risk to the health and safety of the public. This is based on our concurrence with the Browns Ferry Special Review Group's conclusions identified above as well as the significant improvements in fire protection already made at the Trojan facility since the Browns Ferry fire.
These include estblishment of administrative controls over combustible materials and use of ignition sources; training and staffing of a fire brigade; issuance of technical specifications to provide limiting conditions for operation and surveillance requirements on fire protection systems; and the existing detection, automatic suppression systems ard the manual fire suppression means for all areas including those for which an alternate capability is being provided.
At the time of the Fire Protection SER issuance (March 9,1978) the Trojan operating license was amended to require the implementation of the alternate shutdown capability during the refueling outage prior to return to power for Cycle 3 operation.
It was estimated at that time that return to power for Cycle 3 operation would be about June 1979.
Due to unanticipated outages at Trojan, the actual refueling outage will be later than June 1979.
This refueling outage is presently scheduled for Spring 1980.
The modifications for alternate shutdown capability will be implemented at that time.
For the reasons stated above, we continue to hold that this implementation schedule is acceptable.
Unresolved safety issues are identified in NUREG-0510, " Identification of Unresolved Safety Issues Relating to Nuclear Power Plants", Report to Congress, January 1979. Justification for continued plant operation is discussed in this report and in NUREG-0371.
[.5@d293.
Mr. Lynn Frank.
Copies of all NUREG documents mentioned (except for encicsure 3 below) in this letter have previously been made availble to ycu.
i nce rely, k.v' E ! ji ictor Stel4,o(,' dr{., U1 rector j'I D0ivision of Operating Reactors
Enclosures:
1.
W response to Question I.3 2.
Remo to the Commissioners dated March 15,1979 3.
9
cc:
Mr. H. H. Phillips Portland General Electric Company 121 S.W. Salmon Street Portland, Oregon 97204 Warren Hastings, Esquire Counsel for Portland General Electric Company 121 S.W. Salmon Street Portlanc, Oregon 97204 Mr. J. L. Frewing, Manager Generation Licensing and Analysis Portland General Electric Company 121 S.W. Salmon Street Portl and, Oregon 97204 Columbia County Courthouse Law Library, Circuit Court Rocni St. Helens, Oregon 97501 Director, Oregon Department of Enercy Labor and Industries Building, Room 111 Salem, Oregen 97310 Richard M. Sandvik, Esquire Counsel for Oregon Energy Facility Siting Counsel and Oregon Department of Energy 500 Pacific Building 520 S.W. Yamhill Portland, Oregon 97204 Michael Malmrose U. S. Nuclear Regulatory Commission Trojan Nuclear Plant P. O. Box 0 Rainier, Oregon 97048 566293'
t
-id-(Conti nued)
RESPONSE
Beginning in mid-1976,' Westinghcuse initiated a retesting program, tor
~
the instrument transmitters utili~ zed by tee plants in the program (Table
- 1) at the request cf the NRC, to demonstrate their capability to perf orm their required functions, either trip or icng-term monitoring, under more severe environmental conditions than previously employed. The retesting et the original models of transmitters preved unsuccessf ul f or long-term monitoring f unctions and as a consequence, between Aucust and Septemb: ~ 1977, Westinghcuse issued detailed, transmitter replacemen:
recccnendations to those custcmers within the scope ni this pregram (Table 1).
On mere result license acplications (O. C. Ccck and North Anna) the statt has introduced the additional requirements that sequential testing be employed and that a minimum et one hour cperability be demcnstrated for transmitters employed f er shcrt-term reactor trip and/cr saf ety injection autcmatig prctective functions in a high energy line break Pr'or to this change in NRC Requirements Westinghouse had i
environment.
considered that transmitters qualified f or shorter periods than one hour were capable et performing short-term'f unctions.
As a consequence of these additicnal 'statt recuirements, the information contained in WCAP-741CL fustifying the qualiticatien of Barton, Foxboro and Fisher-Perter pressure and ditterential pressure transmitters ter short-term saf ety related hign energy line break applications is hereby withdrawn frcm consideration within the Sucplemental Program. Westing-
[
house will isshe additicnal transmic:ar replacement recc=endations, as acpropriate, to the plants within this prcgram (Table 1) tc reflect this change at position.
Since the transmitter qualitication test results contained in WCAP-7*1CL will not be relied ucen ter instrument qualiti-caticn ter the plants in this program, Question I.3, in total, requires no further resconse.
btd5296
3 March 15,1979 SECY-79-ll2A p:;
x CM:
The Comissioners b::
THRU:
Lee V. Gossick, Executive Director for Operations ',#'-
flf -
FROM:
Harold R. Denton, Director, Office of Nuclear Reactor Regulation SUBJECT-UNION OF CONCERNED SCIENTISTS' PETITION FOR
[.:.:.
RECONSIDERATIOM - ENVIRONMENTAL QUALIFICATION lW OF ELECTRICAL EQUIPMENT g
v;
.0 In a memorandum dated December 12, 1978, B. Snyder of the Office
[d of Policy Evaluation requested that we respond to several questions
[
regarding environmental qualification of electrical equipment in f
operating plants.
Our response is enclosed.
iz
/
?
H. R. Denton, Director
^
h:.:
Office of Nuclear Reactor Regulation p
V::
Enclosure:
2:
fI.
Response to B. Snyder's memorandum dated December 12, 1978
[.
I cc:
NRC POR p[T Union of Concerned Scientists SECY DISTRIBUTION:
Commissioners Cor lission Staff Offices Exec. Dir. for Opers.
F...
ACRS If Secretariat
[~
L:
3 505295
-wD t
b b
[D y 0o930
.a RESPONSE TO B. SNYDER'S MEMORANDUM OF DECEMBER 12, 1978
=tt
-Ouestion 1 SIE
==
"To what degree has the staff relied on probability analysis when it
==
states on page 36, Appendix 8 to the December 15, 1977 sta ff memo,
==.r that one of the reasons that no immediate action is required is that-E'_..
"the likelihood of a major accident requiring the perfonnance of this pg equipment is very low."
[..
h Resconse The staff has not conducted an analysis of the probability of a major p
ac-ident as part of its consideration of the adequacy of environmental f
e-qualification of safety related equipment. However, in reaching judgments E
in this issue as to the type of action (i.e., ir:nediate or otheraise) that ir V
should be taken to ensure no undue risk to the public, the level of y
protection provided in the facilities to prevent sudden pipe breaks
!E
[=
(" major accident requiring the performance of this equipment") is considered.
Experience in comercial power reactors alone is sufficient to demonstrate b
F that the likelihood of such events is low.
In addition, data developed p
from similar piping system designs in other industries is in agreement with
{!
this experience.
b E
[.
As initially outlined in the staff memoranda of December 15,1977 and i
March 1978 and subsequently in Item 11 of Enclosure 1 of the July 6,1978 response, the scope and timing of staff programs to provide additional ilL confidence that adequate envirormental qualification of equipment exists are based on several factors, including the likelihood of a major accident requiring the performance of this equipment.
Stni29G
/
E
- =
2
.m
_ -. - - _r..g..
The degree to which this factor has shaped the staff's actions is difficult
...l..~ ~ ~
E..
to quantify. In cases in which the licensing s taff had insufficient Wr" confidence that equipment important to safety would function in a major s;;;;;;; -
accident, the plants were required to shut'down and remedy the t:
problem (e.g. D. C. Cook Unit 1 and Pilgrim Unit 1). These decisions were reached with little or no consideration as to the likelihood of such a
" major accident."
In other cases, the staff judgment was that the equipment could be de=cnstrated to be qualified and additional time was F
I allowed for such demonstration in part because the likelihood of an
[
accident environmentally challenging this equipment during the time recuired to confirm or furtner document its qualification was low.
E:
t-
{-
In continuina to reccmmend that no immediate action need be taken, the
- f.._-.
staff does not rely solely en the icw likelihood of a major accident, but
{'.:.
{.'
rather is guided primarily by its judcment, as discussed in the response to L.t_.
I~
question 6 below, that equipment required for safety will not fail before
[.l performing its safety function when exposed to design basis accident h!_.
EE condi tions.
[f=.
m- =
O I'
l-7{
i:
tg 5
r t
i-
s Question 2
=
Item 17 (page 40) of Enclosure 1 of the staff's August 31, 1978 memo does not appear to address the UCS statements as requested.
.Zl.I It appears from the staff's response that D.C. Cook was permitted to operate prior to complete demonstratica for all envirorr. ental r==
qualification.
If this is so, on what basis was this decision
[@s made? Was such operation without complete environmental justifica-p=
tion consistent with the Commission's regulations?
Is the follow-ing statement in the April 13 Order on page 25, footnote 25 correct:
"As a pre-condition for initial operation, the staff p
required the licensee to document adequate environ-mental qualification of numerous electrical components,.
=
including connectors and terminal blocks."
Have other plants been permitted to initiate operation prior to y$
full environmental qualification?
If so, please identify those h
plants and. explain. Also, has full environmental qualification p
now been provided?
p l:.(
t?
RESPONSE
E g
A.
Summary Item 17 of Enclosure 1 of cur August 31 submittal summarized the actions taken by us in licensing D. C. Cook Unit 2 to begin operation.
(:.:_..:
We described the bases for our conclusion that safety related e
. electrical equipment was adequately qualified at the time the decision was made to allow Unit 2 to operate at significant power levels. That decision was made by us on the basis of data supplied by the licensees to support their statement that safety related electric-i ecuipment woulgf perform its safety function if a design basis evert were to occur.
However, we also required at that time that the applicant i
l perform in the near future certain confirmatory tests to provide r
ined2S8 L
.w.
w.
~
additional information. Most of those tests have been ccepleted and,
- n.,
~
for the most part, support our original judgement.
Therefore we
']
believe that the footnote at the bottom of page 25 of the April 13
,. =. =.
Order is correct.
b c
fG...
F, The UC5 petition, together with the results of the Sandia tests referenced in that petition, identified a need to intensify our review
. of environmental qualification information for some equipment.
Since that time, staff review in this area has been intensified.
Only four plants -- North Anna Unit 1, Cook Unit 2, Hatch Unit 2 and Arkansas a
Nuclear One,. Unit 2 -- have been licensed to initiate operation since
{
the petition was filed. The staff found the environmental qualification i:
program for both the Hatch and Arkaatas plants to be acceptable. The
[1.
ps
~ ~ ~
North Anna and D. C. Cook plants were also acceptable with One exception.
The exception concerned a need for confirmatory data on several kinds of pressure transmitters used in both plants.
For both Cook Unit 2 and hL North Anna Unit 1, we permitted operation based on our judgement that the
[}v transmitters would perform their safety function if called upon, but R
t we required that additional tests be per' formed to confirm that judgement.
The response to question 6 addresses the status of equipment qualification p
[
for plants licensed prior to D. C. Cook and North Anna.
flk.;
5 E
[
As noted above, almost all confirmatory tests in connecticn with the c--
Cock Uni: 2 OL have now been ccmpleted.
There are still some unresolved questions regarding tests performed on Barton pressure transmitters which were one of severai transmitters requiring confirmatory testing. We wil.1
[
recuire that additional tests be performed to resclve the questions remaining f
udNN/
...Y
-3
=. g on the Barton transmitters. With that exception, all equipment qualification
~7im
~' k':7 issues have been resolved for D. C. Cook Unit 2 and North Anna Unit 1.
=
More details on each of these points are presented below.
This informa-tion is intended to supplement the response provided in our August 31, 1978 0
submittal.
B.
Sicnificance of Adecuate Oualification Some clarification of our position on the adequacy of equipment qualifi-iE b(:--
cation may be helpful.
The UCS statements refer to " complete" demonstra-tion of equipment qualification. We believe that the word " complete"
[.{!
p connotas an achievement of perfection that is neither possible in an p
engineering sense nor required by the Commission's regulations.
The L.L staff conclusions with regard to the adequacy of environmental qualification I3 are a matter of engineering judgement based on a technical evaluation
~
il of the available experimental and analytical data.
Further, the details
[n of exactly what information.is needed to reach such a finding are continuing
((]f iMs?
to evolve as we acquire and analyze more data.
Today we believe there jggi cm..
is sufficient technical information available to make judgements that EEE equipment is adequately qualified, but there is always the possibility
[
i that, as we expand our equipment qualification data base, new information
[{
may ce developed that would require reassessment of scme equipment..
b i
.I
-.m..
/E!=
N[s p..-
=.
506300 3.4 g=
k e
J
I 4
C.
Licensina of D. C. Cook, Unit 2 On March 7,1978, near the end of our review for licensing Cook Unit 2 initial operation, we met with the licensees. They described both test results and test procedures for transmitters, terminations, cables, and
....g penetrations to demonstrate that this equipment was adequately qualified.
,.We concluded that most of this information, once documented, would be sufficient to ensure that all electrical equipment would perform it.s safety function if a design basis event were to occur. However, the licensees justified the operabili'ty of Barton and Foxboro pressure transmitters on the basis of it=
separate effects test.
In separate effects tests, different pieces of g
p --
equipment or components are exposed separately to the different environmental
$p conditions (e.g., radiation, temperature, humidity) produced by design basis j
- =
events.
Engineering judgement and analysis are used to assess the effects of ccmbining these conditions. While we found the results of these separate b
effects tests acceptable, we believed that 'ull sequential tests should be
- .a performed to confirm our judg. ment regarding tr5nsmitter operability in the
((:..c EK."-
event of a design basis accident.
E pga; Fi z_
Following receipt of the required documentation and other review matters Q
being satisfactorily completed, k.endment.2 of the D. C. Cook Unit 2 operating license, which authori:ed operation at or below 5% power, was issued on March 8,1978. This amendment included four license conditions 3j::
f=;_
at;6301 g
h--
I L
.=
related to eauipment qualification. The conditions are summarized as
. = =
"=
follows:
Condition 3.8.1 required that the licensees provide within 90 days (June 8,1978) results of full sequen-
+
tial qualification tests, in accordance with IEEE-323-1971, to demonstrate the qualification of Foxboro and Barton transmitters and, within 2 weeks, provide a basis for continued operation of the facility during the time required to complete these tests.
Condition 3.8.2 required that the licensees provide, f
within two weeks, qualificaticn test procedures and results for all electrical cable connections and ter-
}-
minations in safety related circuits within containment.
=
r Condition 3.8.3 required that the licensee 5 provide f
within two weeks the doccmantation of test results and analysis that demonstrate the environmental qualifica-
{F tion of safety related cables.
Eik-Condition 3.B.4 required that the licensees provide within two weeks documentation demonstrating compara-M=
bility of electrical penetrations installed at D.C.
4 Cook Unit 2 to prctotype penetrations which had under-gone testing under steamline break environmental conditions.
We recuired that these conditions be resolved cricr to coeration above 20% of rated cover. This requirement was selected somewhat R
arbitrarily, but was derived from the staff's ceneral kno.ledae that fG accidents at this power level would result in much less severe y
environmental stress on safety equipment than accidents at higher power l evel s.
Because the conditions called for the licensees to sucoly.
n' most of the documentation within two weeks, the oower limit of 20%
StA302 h
l' t
i.
J
- --6:: : :
was expected practically to apply only to the transmitters.
As notec
..}; iii above, it was our belief that these transmitters would perform their jf_] s!M_ safety function if required and it was expected that the confirmatory sequential tests would confirm that judgement, The licensee believed =- in March that these tests could be completed by the time the plant was expected to reach 20% power, although we also considered the fact that, up until that point, the fission product inventory would L fjl. ~ be inconsecuential even if the transmitters were to fail and saf~ety t.- eauioment did not function as designed. [i L Amendment 3, issued on P. arch 16, 1978, allcwed pcwer operation above 5% and up to 20% of full rated power. None of the qualification related i conditions added to the license as a consequence of Amendment 2 were i altered by Amendment 3. [] g.... i-On March 20, 1978, the licensees met with the staff and provided in-IL _.. p formation which satisfied Co.nditions 3.S.3 and 3.B.4 of that Amendmen* E F They also provided information in acccrdance with Condition 3.8.1 to V justify their request for permission to operate above 20% power before the tests on the Foxboro and Barton transmitters were completed. This information included additional results of testing performed Ip on these transmitters. We concluded that these results provided 1._ additional assurance that these transmitters would perform their function in the unlikely event of a desgin basis accident. This Ni additional informaticn was olaced on the Cook docket at that time. i i s,6303 7.- 6 i
K At the March 20th meeting, the licensee also provided test data for safety-related connections and terminations used in D. C. Cook Unit 2 ~.T =; (Condition 3.B.2). Nearly all of these connections and terminations \\?h?" [E"W successfully passed tests which, for the most part, simulated main R~~ ry. - l:: steam line break accident conditions. Power cable terminal blocks i t-failed the tests and were replaced with qualified splices. Based on p. this information, we concluded that the safety-related terminations I t:.. and connections would perform as designed in the event of a design jcf basis accident and thus that Condition 3.B.2 was satisfied. 'owever, y H .( we concluded that full sequential confirmatory tests on these connections i and terminations should be performed to confirm that conclusion. t ~-. [.. Amendment 4 to the license, which authorized operation at 50'; of P-6 rated power, was issued on March 24, 1978. It called for the full i"y sequential confirmatory tests, for both the Barton and Foxboro trans- [p mitters and for the connections and tenninations, to be completed by 't June 6,1978. (The 50% power limit was imposed for reasons not E= related to equipment cualification). The decision to allow operation .1- ~ - n it power levels above 20% prior to receiving results of full sequential [ tests on the transmitters (instead of limiting power to 20% as stated f in Amendment 2) was based on the additional information we received at b the March 20 meeting and our judgement, based on that information, that the [:.y -g.. ecuipment would perform its safety function. Also the lice. sees indicated i= that the confirmatory, sequential tests were to begin shortly cnd that we p. would be made aware imediately of any results which indicated t.'at the j; I transmitters were inadecuately qualified to perform their safety functions. I t. }. St43C4 h
s . _3: l = m Subsequent to the issuance of Amendment 4, the licensees met with the .b 5:- staff in April and May and provided additional information on both the h~g transmitters and safety related terminations inside containment. { ,::..i; ihr-Foxboro Transmitters E Additional documentation of separate effects tests was provided for staff review. The licensees askedthe staff to reconsider the needfor sequential tests. However, the staff's requirement for full sequential testskas based en the limitations of separate effects testing. Therefore, this cocumentation was not sufficient to change our judgement that additional confirnatory s-tests were necessary. Rather than pursue these confirmatory = tests, the licensees committed to replacing these transmitters f ~ C with other qualified instruments prior to startup following the first regularly scheduled refueling. We found this proposal acceptable based on cur judgement that the Foxboro instruments were qualified to perforn their p function if required. (As noted in our August 31, 1978 pi submittal to the Co. mission, one model of the Foxboro 5_ transmitters survived the separate effects test while hi__ the other model failed after 4 minutes. It was our E_ E~~~ judgement that 4 minutes was adequate to ensure that the transmitter would perfom its safety function. E; However, it could not be assured that it would perform E.. its post accident monitoring function. The applicant E provided an evaluation which demonstrated that, by modifying E. 2.. plant operating procedures, proper operator action during !(__ %;h.3 b the post accident period could be assured. These modified procedures were to be utili;:ed only until the instruments = were replaced.) - m -- r;= Barton Transmitters We reviewed the program proposed by the licenseef to sequentially test the Barton Transmitters and con-k cluded that it was consistant with the conditions speci-E fied in Amendment 4. However, we were told that the " ~ ' test program could not be complete until July 1978, with a formal report to de submitted by October 1973. ~~ It was our judgement that conti".ued operation of Unit 2 until October 1978 was acceptable because we viewed = these tests as being confir atory in nature and because we would be informed prom::tly of any deficiencies revealed {l { by the test in. July. .= L66305 p e:. t:: t . :~."
i l X i .g. tz:". j Safety-Related Terminations Inside Containment The licensees provided documentation which satisfied i~l. the condition 3.B.2 specified in Amendment 4. . =. _ = j Safety-Related Cables i e_ she applicant also informed the staff that one type -- - 5-of cable previously found acceptable in accordance E with condition 3.b.3 of Amendment 3 had failed further E. ~ tests being performed by the applicant. These additional tests were conservative in that the test conditions were selected to envelop both the high radiation conditions p associated with a postulated LCCA and the high tempera. E tures associated with a steam line break accident. p Rather than retest to more realistic conditions, the [ applicant elected to replace this cable with cable that EM-had survived qualification tests. [ii f Amendment 6, issued en June'16,1978, reflected our decisions regarding p:. the Foxboro and Barton transmitters and called for completion of h v. sequential testing of the Earton transmitters by October 1978. As I::7 noted in Section belew, those tests have been completed and the results have been reviewed by the stcff. p 5.= F In ummary, there was an i'ntensive review of environmer.tal cualifications by { the staff in the process of grantinc D.C. Cook Unit 2 cermission to coerate. I h[ r was our judgement prior to granting pennission to allow cperation at reduced power that equipment was adquately qualified. To con-firm that judgement, we required that the licensee provide additional [ documentation and, in some instances, the results of additier.a1 != tests. That process took about seven months to comolete but, for 5 the most part, it confirmed that cur initial judgements were correct. As noted in cur August 31 submittal, in two instances (for certa n terminal i biccks and one type of cable) the tests 'did not show that ecuiment was adecuately cualified and the a;:plican replaced it. ot>S3CG =
- u. '
D. Licensino of North Anna, Unit 1 The staff licensed the operation of North Anna Unit 1 at about the same time that D. C. Cook Unit 2 was undergoing licensing review. On April 1,1978 North Anna was licensed to begin power operation conditioned, in part, on the future perfomance of full sequential confirmatory tests on Barton and Foxboro transmitters installed in the plant. These transmitters were of the same design as those used in D. C. Cook. These tests were to be completed within 90 days. l [q j.: Subsequent to that date the licensee decided that, rather than test both the Foxboro and Barton transmitters, the Foxboro transmitters would be replaced with Bartons. ( Amendment 7 to the North Anna Unit i license, dated July 3,1978, [-; _- 1 reflected that change. It also. allowed the licensee an additional t-3 months (from July 1 to October 1,1978) to complete the confirmatory [. sequential tests on the Barton instruments. These were the same tests which were to be used to 'confim the adequacy of the qualification 5 of the D. C. Cook ins truments. This delay was found acceptable for North Anna for the same reason as for D. C. Cook. f Bc . u :. d Sc6307 l z. n. b
l a m.; * " " *
- f...--
a E. Current St.:tus of the D. C. Cook / North Ann, Environmental-Oualification = rs = Review E]; r=s We noted in our August 31, 1978, memorandum to the Commission that "-7.7.l.5 = problems had developed during the sequential tests performed on the Barton transmitters and that we had requested additional infomation. These sequential tests were performed by Westinghouse'and, on the 29th of Sep'. ember, they provided the staff with a report describing all tests performed and the results obtained. This report sho'wed that b:) i the instruments performed properly during the time they would be -~ required to perform their safety function of initiating protective actions. However, at a later stage in the test, which corresponded' to the post accident monit:, ing phase, the transmitters exhibited larger L t :,:,._.. than expected errors that fluctuated with time. At the time the test was completed, all transmitters were again functioning properly. !?- y_y n The environmental conditicos for these sequential tests were established 5.. by selecting a conservative radiation / steam / temperature / pressure pro-f[ ism = F... file that bounded those conditions expected for both the steamline ~~ break (SLB) and loss of coolant accidents (LOCA). When the accuracy 2. problems were detected, additional tests were performed to help iden-tify the cause. From their analysis of there tests, Westinghouse l deduced that the problem was due to the combined effects of high ![l_ b radiation and hi h temperature produced by attempting to test for f g i L F non3C8 J.. I
. ~ w both. LOCA and SLB conditions simultaneously. This 'tpproach was employed z.y. for testing efficiency - it was not required for meeting MRC licensing ~~ requirements. In addition, Westinghouse believed that these additional = tests demonstrated that the transmitters would perform properly when subjected to the less conservative conditions (but accepcable under NRC requirements) for LOCA and SLB accidents consider d separately. I We have concluded that the tests demonstrate that the instruments .-.z. [.:; will perform their safety function (i.e. initiate protective action) h-t.. in the early stages of either a LOCA or SLB. Vc iso agree with if t-Westinghouse that the errors in instrument 'eadings wr:ich occurred Ii during the post accident monitoring pbu e of the sequent::' tesi; f i:' may be due to the unusual way the seg iential test was performed. (cc=bining LOCA ar.d SLB). However ..t nian to request that full g. p~:i~. sequential confirmatory tests be completed to confirm cur judgement. [ h7 e ki_. i. [?;i. ~ SG6309 g L I [x -.
b ...=;- Question 3 =- "Please furnish an updated response to Iten 2 of Enclosure 1 of the staff's N August 31, letter addressing the further results of the staff's ongoing .= review. Can the staff now fully respond to the question?" p SUN Resoonse 7;-;;... i-The question responded to in Item 2 of Enclosure 1 of the staff's August 31,1978 ] memorandum for the Comissioners was as follows: = "Are there any plants which cannot demonstrate envirorcental qualification for electrical connectors, splices, penetrations, and terminal blocks with full docamentation such that they - meet the minimum requirements of the Comission's regulations? f" If there are, please explain the legal and regulatory basis for permitting their operation." In its response to this question, the staff was able to verify that adequate documentation existed for connectors, penetrations, and unprotected L .~ ~ ~ ~ terminal blocks. With regard to protected terminal biccks, splices, and other electrical equipment, the staff is still unable to fully respond. [ L: g~- In the August 31, 1978 response the staff stated that it would b-continue to pursue the ;uestion of environmental qualification and documentation in two principal ways: (1) the Systematic Evaluation Program (SEP); and (2) IE inspections of. licensee acticns in response to IE Circular 78-08. The preliminary results of the SEP were previously reported in NUREG-0458 and centinuing SEP efforts are ciscussed in de response to Question 5 belcw. IE inspections related to Circular 7E-08 g are centinuing. The results of this effort since the August 31, 1973 respcnse and subsequent followup acti.cns are summarizec in the folicwing paragrapns. [ atb310 [ifi: b:
l =33 As a result of IE inspections conducted to date, items of safety related T f: equipment located inside containment have been identified where documentation hEi to demonstrate environmental qualification was not available at the time ~ of the inspection. Licensee efforts to produce documentation for this a equipment are continuing. f-In addition to the apparent documentation problems identified, at least'ane plant has been identified which had unqualified splices in safety related circuits (see Trojan LER No. 78-027). These splices have been re'placed I and the licensee has expanded its inspection program to check for similar i problems in other electrical circuits. i i The only other ccmponents that have been identified as being unqualified 7t. I are certain specific models of valve position indication limit switches. E These limit switches are different from those identified in an earlier i j IE Bulletin (IEB 78-04). A failure of these switches would not directly I result in a loss of valve function, only the position indication would be p r affected. The switches are being replaced with qualified components at h; iG each of the plants where their incorrect application has been identified. i I i. The above findings are preliminary since many of the licensees' reviews of qualification documentation initiated by the Circular are only partially completed. However, based on these preliminary findings, the staff . 7_. concluded that IE Circular 78-03 is not receiving the level.cf attention ~ frcm all licensees that the staff believes is warranted. To excedits ccm-pletion of the licensee's re-review prograri the staff nas issued IE sulletin 79-01 (see attacnment). This bulletin requires that the licensees' SGS311 i J
3 complete their re-reviews within 130 days and report to :he staff, in writing, the documented basis for qualification of electrical equip-g ment required to function under accident conditions. NRC inspections of licensees' programs for review of coconent qualifi-cation dodumentation will continue in conjunction with IE Circular 78-08 and IE Bulletin 79-01. The basis for continued reactor oceration while the staff and licensees continue to pursue the question of I environmental qualificatien and documentation remains as sta:ed in the p.e August 31, 1978 res ponse. IP+ L p != _e_;~+ k I c. L c E'.. g .N;:.- I:.- g- ~ =.a. {:g.-+ E 0$,Q 4 ...:._.T '^.~.2 ~~ ^
g Atta:hment To A Resp:nse No. 3 _+ UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE 0.F INSPECTION AND ENFORCEMENT WASHINGTON, D.C. 20555 { En February 8,1979 R IE Bulletin No. 79-0I { EINIRONMENTAL QUALIFICATICN OF CLASS IE EQUIPMENT ~ i Description of Circumstances: l The intent of IE Circular 78-02 was to highlight to all licensees important lessons learned from environantal qualification deficiencies reported by individual licensees. In this regard, licensees were i reque-ted to examine installed safety-related electrical equipment [ and determine that proper documentation existed which provided assurance that this equipment would function under postulated accident conditions. The scope of IE Circular 78-08 was much broader than other previously [ ~ issued Bulletins and Circulars (su.ch as IES 78-04 and IEB 78-02) which F addressed specific comptnent failures. The intent of this Bulletin is to raise the threshold of IE Circular 78-08 to the level of a Bulletin; I-i.e., action requiring a licensee response. [ e Inspections conducted to date by the NRC of licensees' activities in response to IE Circular 78-08 have identified one component which licensees l have found to be unqualified for. service within the LOCA environment. L" Specificially, unqualified stem moented limit switches (SMLS), other than those identified in previously issued IE Bulletin 78-04, were found to be i installed on safety-related valves inside containment at both Duane Arnold t-l and Quad Citie 1 and 2 Nuclear Generating Stations. The unqualified switches '?~ are identified as NAMCO Models SL2-C-11, S3CML, sal-31, SAI-32, D1200j, b i EA-700 and EA-770 switches. According to the manufacturer, these switches 5= are designed only for general purpose applications and are not considered I suitable devices for service in the LOCA environment. Consequently, l-switches are being replaced at the above power plants with qualified F components. ~ [- i Also, NRC inspection of component qualification has identified equipment
- ~
which does not have documentation indicating it is qualified for the LOCA environment. The inspecticns have also identified that the,. licensees' re-review and resolution of problem areas are not receiving the level of attention from all licensees which the NRC believes is warranted. Because of the protracted schedule for completion of the rs-review, we are new requesting the power reactor facilities with operating licenses to expedite c:mpletion of their re-review program originally requested by IE Circular 78-08 dated May 31, 1978. s. Page 1 of 2 W 313 ^ _ - ;- y -
1. IE Bulletin No. 79-01' February 8,1979 I 1._ Action to Be Taken By Licensees of All Power Reactor Facilities ~' JExcect Tnose 11 SEP Plants Listed on Enclosure 3 With An Operatino License: ~" 1. Complete the re-review program described in IE Circular 78-08 within 120 days of receipt of this Bulletin. .p 2. Determine if the types of stem mounted limit switches described above are being used or planned for use on safety-related valves which are [ located inside containment at your facility. If so, provide a [ written report to the NRC within the tirre frame specified and to -the address specified in Item 4 below. 3. Provide written evidence of the qualification of electrical { equipment required to function under accident conditions.* For j those items not having complete qualification data available for i. review, identify your plans for determining qualification, either by testing or engineering analysis, or combination of these, or by i replacement with qualified equipment. Include your schedule for completing these actions and your justification for continued operation. ~ Submit this' ir1formtion to th'e Director Division of Reactor Opera-tions Inspection, Office of Inspection and Enforcement, Nuclear Regulatory Comission, Washington, D.C. 20555 with a copy to the i appropriate NRC Regional Office within 120 days of receipt of this [ Bulletin. 4. Report any items which are identified as not meeting qualification requirements for service intended to-tte Director, Division of ~ Operating Reactors, Office of Nuclear Reactor Regulation, Nuclear-. _ i Regulatory Comis:icn, Washington, D.C. 20555 with copy to the [. appropriate NRC Regional Office within 24 hours of identification. l -~ If plant operation is to continue following identirication, provide [ justification for such operation. Provide a detailed writte.1 report within 14 days of identification to NRR, with a copy to the cppro-priate NP,C Regional Office. No additional written response to this IE Bulletin is required other than those responses described above. NRC inspectors will continue to monitor the licensees' progress in completing the requested action described above. If additional information is required, contact the Director of the appropriate NRC Regional Office.
- This written evidence should include: 1) compenent description;
- 2) description of the accident environment; 3) the envircament to whicn the ccmponent or equipment is qualified; 4) the manner of cualification which should include test methods such as secuential, synergistic, etc., and 5) identification of the specific suppor;.ing qualification documentation.
t Page 2 of 2 566314 =
s IE Bulletin No. 79-01 L=h February 8; 1979 3 = y=--q-- LISTING OF IE BULLETINS ~ ISSUED IN U.ST TWELVE MONTHS ~~ g...
- ..Z..'
Bulletin Subject Date Issued Issued To .No, {:) L.1 78-03 Potential Explosive 2/8/78 All BWR Power Gas Mixture Accumula-Reactor Facilities tions Associated with with an OL or CP BWR Offgas System
- =
l Operations h. p 78-04 Environmental Quali-2/21/78 All Power Reactor h fication of Certain Facilities with an i.L i Stem Mounted Limit OL or CP i-Switches Inside F Reactor Containment {rg 78-05 Malfonctioning of 4/14/73 All Power Reactor Circuit Breaker Facilities with an [_.. _ Auxiliary Contact OL or CP g Mechanism-General s Model CR105X k 78-06 Defective Cutler-5/31/78 All Power Reactor I l Hamer, Type M Relays - Facilities with an - L: With DC Coils OL or CP.-- p(-- l 78-07 . Protection afforded 6/12/78 All Power Reactor i by Air-Line Respirators Facilities with an and Supplied-Air Hoods OL, all class F and F l Research Reactors with n I an OL, all Fuel Cycle { Facilities with an OL, and all Priority 1 Material Licensees 78-08 Radiation Levels fram 6/12/78 All Power and b i Fuel Element Transfer Rerearch Reactor k Tubes Facilities with a i~~~ Fuel Element i transfer tube and an OL. Enc 1c.eure 2 Page 1 cf 2 [ t 5C63.15 i i. L
IE Bulletin No. 79-01
- P February 8,1979 LISTING OF IE BULLETINS
~:. - - ISSUED IN LAST WELVE MONTHS Bulletin Subject Dat'e Issued Issued To k No. 78-09 BWR Drywell Leakage 6/14/78 All BWR Power in = ~ ~ Paths Associated with Reactor Facilities Inadequate Drywell with an OL or CP [: Closures i 78-10 Bergen-Paterson 6/27/78 All BWR Power l -- Hydraulic Shock Reactor Facilities Suppressor Accumulator with an OL or CP b.. Spring Coils p.. 7 78-11 Examination of Mark I 7/21/78 BWR Power Reactor f Containment Torus Facilities for F Welds action: Peach E Bottom 2 and 3 j. Quad Cities 1 and 2, Hatch 1, Pcnti-i cello and Vermont l-Yankee 78-12 Atypical Weld Paterial 9/29/78 All Power Reactor j( in React:r Pressure Facilities with an j( Vessel Welds OL or CP 78-12A Atypical Weld Pateri4L --11/24/78 All Power Reactor in Reacter Pressure Facilities with an
- t..
Vessel Welds OL or CP L
- g. =.
78-13 Failures In Source Heads 10/27/78 All general and 0 ~- of Kay-Ray, Inc., Gauges specific licensees Pcdels 7050, 7050B, 7051, with the subject L3= 7051B, 7050, 7050B, 7CGI Kay-Ray, Inc. r and 7051B gauges i i 78-14 Deterioration of Buna-N 12/19/78 All GE BWR facilities Ccepenents In ASCO with an OL or CP Solenoids e G.: Encicsure 2 Page 2 of 2 j.... "~~ _' ~ ~ ~ I
, Enclosure No. '3 ~ SEP Plants t :. !.E P1 ant Region L:: -. Dresden 1 III ri-L: l-Yankee Rowe I i Big Rock Point III San Onofre 1 V e Haddam Neck I Lacrosse III Oyster Creek I l R. E. Ginna I l-t i Dresden 2 III F: Millstone I = Palisades III t Is t i y {- / ^ gg 311 L
Ouestion 4 "Page 12 of the Comission's April 13, 1978 Memorandum and Order states: r.- " Fundamental to NRC regulation of nuclear power reactors is the n principle that safety systems must perform their intended functions in spite of the environment which may result from pos-
- =7 tuhted accidents.
(The controlling regulation here is 10 CFR 50, 2-Appendix A, General Design Criterion 4.) For example, if an ~ electrical ccmponent is required to function in a safety system which was designed to mitigate the consequences of certain [ accidents, that com onent must perform its intended function for postulated accidents such as: (a) loss-of-coolant accident (LOCA), (b) main steam line break (MSLB), or (c) failure of any other high-energy confining system." [ t Is there any inconsistency between that statement and the staff's conclusion that no regulation was violated by the licensees which installed unqualified equipment because the licensees were under no such requirement? Please [ elaborate." i1p Question 5 l "Are all licensees now under a duty or comitment to nave full environmental qualification for their electrical equipment so that any failure to have such [ qualification could result in an enforcement action?" i; I Resconse (Cuestions 4 and 5) {l 1 Questions 4 and 5 are related in that they both deal with the que tion of a licensee's obligation under :ast and present NRC regulations to install i qualified ecuipment in " plants licensed by the NRC. Before the question bp l of whether a licensee is, or has been, in violation of a Commission i regulation can be.:.nswered (Question 4), the extent of a 1icensee's obli-i l gation to comply with the regulation must be defined (Cuestion 5). Therefo re, a response to Question 5 will be pmvided first, followed by a response to Cuestien 4. !? L. With regsrd to Question 5, all licensees are new and have in the past been recuired to assure themselves that equipment im;ortant to safety is g envircrmen ally qualifiec for its service environment. The regulatory 7 basis for -his requirement a ra General Design Criteria 1 and a wnicn l~ s I % ~ ev i-mabn318 -w i t. l
2 r articulate 1cng standing regulatory practices of the NRC and the AEC before ,1r ~' it. Simply stated, Criterion 4 requires that, structures, systecs, and com-ponents important to safety be designed to function in both normal and E accident environments. Cricerion 1 requires that a quality assurance pro-gram be established to assure that these structures, systems, and compunents ~ h-are designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. To the extent that licensees are not in compliance with these general recuirements, they are in violation of the Commission's regulations and are i subject to corrective enforcement action. ,e These recuirements have been fundamental to the development and regulation [ of the nuclear industry and are generally understood by individuals familiar with the industry. The issue in question is what constitutes i-i "'ull" or adequate environmental qualification and the extent to which the i.c. I i qualification must be documented. L-This question has been addressed by the staff in several previous submittals [- to the Commission / In summary, industry practices and commitments which .I E-have baen found acceptable by the NRC staff for satisfying qualification E I I 3/ ll of the staff's memorandum for the Commissioner's dated E March 23,1973, subject: Union of Concerned Scientists Petition. [... 3 "Recort on the Historical Evolution of Environmnntal Cualication r-Recuirements for Safety-Related Electrical Ecuipment", dated Cecember 15, 1977. This reocr was part of the overall "NRC staff Recort on Union of Concerned Scientist's Petition for Emergency and Remedial Action", also dated Cecember 15, 1977, t 56S319, = ..= = = =.... a
3 requirements in the past have ranged from a simple specification of the hi ghes t industrial quality components available at the time for the oldest f-- plants, to a licensing commitment to a comprehensive program of environmental ~ qualification for. newer plants in the CP stage of licensing which is in 2 accordance with industry standards such as IEEE Std. 323-1974 / and i.zjl~ documented in accordance with 10 CFR 50, Appendix B requirements for quality I assurance crograms. L General Design Criteria 1 and 4 contain no detailed systematic requirements [ L for either quality assurance methods or for systematically maintaining quali ty records. The specific oualification and documentation require-h i ments aaplicable to individual plants ara set forth in the license hL aaplications aaprovec by the staff at the tite of initial licensing and in all current conditions of tne licenses, including technical specifications.
- t.. -
For older piants the informtion contained in license applications and technical specifications tends to be general in natura, consisten* i with the more generalized provisions of Criteria 1 and 4. Consequently, j_ U it is often very difficult to identify a " violation" of tnese generalized [L!' specifications. However, enforcement action in the form of Orders can { be taken even wnen no specific regulatory requirement exists wherever [ cotentially hazardous conditions are identified. As discussed in response to Question 6 beluw, the staff is not aware of any such conditions. ~ E E..V= . ::= U nstitute of Electrical and Electronics Engineers Stancard 323-1974, I "IEEE Standard for Qualifying Class IE Ecuipment for Nuclear power Generating Stations." c. F i~ St;6320 ]- E_. -
Z.~. 4 With regard to Question 4, in view of the above discussion, the staff believes that there is no inconsistency between the statement on page 12 of the in 7 p.=.s Corm 11 s s i o,n 's Ap ril 13, 1978, Memorandum and Order and the staff's conclusion ~ that no regulation was violated by the licenseeswho installed equipment that f: was later detemined to be of questionable qualification. This conclusion
- 7..,.
was based on the staff's finding that the equipment was installed in j-.. compliance with the general requirements of General Design Criteria 1 and 4 and no identified violation nf Appendir R. The staterrent that the [: "the licensees were under no such requirement" was intended to mean that [T some of the licensees were uncer no specific regulatory requirements to [p use a specific method of qualification (e.g., testing) or to maintain h detailed qualification documentation. Ec:p_ As stated above, all plants are required to be in compliance with General jk. t:. Cesign Criteria 1 and 4 If for any reason a licensee determines that a [._ - r::. [:: plant is not in comoliance with these regulations, the licensee must take p p appropriate action to bring the plant into compliance and to seek any f _... p= n! quired regulatory approvals if any changes are evolved. The license. Ji n.. can alternatively seek an exemption from such requirements. ~ 5.=. a:1 EEi. j Even though many of the licensees are not under a regulatory reouf re ea+ to i maint,in specific qualification documentation, the staff has initiated p a program to determine the extent to which the qualification of safety-related equipment in all plants can be documented. The preliminary results of i this program are discussed in the response co Question s. .y u L u c.w. [.' } At this time the emphasis in the staff effort is being placed on detemining the status of qualification and on identifying possible deficiercies. Formal enforcement action will be used if necessary to assure that any needed corrective actions are taken in a timely manner. II-M l J. L L R ~ I- [ _- t. (: 566322 l i e . __~ ' ~ ^
! a-a Question 6 l "The July 6,1978 response to Item 11, Enclosure 1 stated that the staff continues to believe that adequate protection for the public health and i safety exists despite the six plant shutdown. Does the staff believe ~ that some plants in operation use equipment which will fail when exposed to design basis event conditions? What is the basis for the staff's ~' judgment?" Resconse The staff does not believe that equipment required for safety installed in operating reactors will fail before performing its safety function when i exposed to design basis event (DBE) conditions. In making this statement the staff does not mean to imply that test results are available which demonstrate that all safety related equipment would remain functional for the full time duration of all tests that have been devised to envelope C3E conditions. To the contrary, as the Commission is aware, there have been tests performed i-in which safety related equipment did not remain functional for the full i-l- a duration of the test. In each of these cases the staff has reviewed the particular circumstances involved and appropriate action was taken.1/ In C: !~ some cases equipment was replaced; in others it was not. In those cases where i the equipment was not replaced, the staff concluded that the equipment would [ i have performed its safety function prior to failure in actual DBE conditions. The staff conclusion was based on factors such as the length of time the equipment remained functional during the test, the severity of the test conditions as compared to expected DBE conditions, and the actual installed a locaticn of the equipment. The staff acknowledges that if the equipment discussed above had remained 1/ - See the staff's memorandum for the Commissioners dated March 23, 1973,
Subject:
Union of Concerned Scientist's Petition, for examples of staff reviews and actions in connection with recent test results. l 5 % 223
i.--i: e 2_ m=: functional during the entire test, there would be greater assurance that the equipment would perform its safety function in actual D3E conditions. ~ This statement speaks to the fundamental issue in question, i.e., the p t.. adequacy of the level of assurance that exists. The staff believes thtt - F_ there is adequate assurance for the present that safety related equipment is qualified to perform its safety function. This belief is not based on a rigorous component-by-ccmponent review and evaluation by. the staff of specific [ test results for all safety-related equipment. Test results do not exist for many components, nor were they required at the time of initiai licensing t: of the currently operating reactors. Inis belief is a matter of technical [ judgment based on the staff's consideration of all the available test results, L. including test failures as discussed, above, the previous staff reviews conducted at the time the plants were initially licensed to operatek . I I' subsequent operating experience, and reviews connected with staf backfitted kp requirements (e.g.,10 CFR 50.46). j. bb: E Al though the staff bel'es o. that the level of assurance that equipe.ent is [ t.. qualified is adequate f;r the present (i.e., in the short tern), i t also b~ E believes, as stated in its July 6,1978, response to Item 11, that it is [ desirable to increase its level of assurance for the future (i.e., in the long term). The objective of the ongoing staff actions-discussed in the } i response is to increase the existing level of assurance. The following para-I graphs summarize and update earlier staff discussions of these activities. [ f i f. g I lAccendix A to NURE3-0413 provides a detailed discussicr. o ~ the NRC environ-mental qualification and documentation recuirements that r:rmed tne casis j fer ne initial licensing of all currently ocerating re?ct:rs. i i L bbS324 r I. /
- ..i..
The Systematic Evaluation Program (SEP) is an important part of the ongoing ]-
- w=-
staff actions. The results of the initial phase of the SEP for equipment qualification are reported in NUREG-0458. As a result of this activity the [ staff instituted a prograa of augmented IE inspections and issued IE Circular [p 78-08 to feed back lessons learned and initiate action by the licensees to tl f-examine the installation and qualification documentation that exists for equipment located inside containment. (See response to question 3 for a [ A.._. summary of the results of IE Circular 78-C8 to date and proposed followp ipn. ac tions. ) [ Li Subsequent to the issuance of NUP,EG-0453, the SE? evaluatinns have continued and additional information is becoming available. The staff has requested [ each of the eleven SEP licensees to identify the specific conditions and i:: t method by which each equipment was qualified. The licensees are providing [L_ r the informat.an using a tabular format recommended by the staff that !j [ 3.= facilitates a cross comparison of 'he qualification infor aticn for the h= p same equipment used in different facilities. I:. ~ The major staff effort under the SEP progran is being directed to more f. i E i precisely define adequate methods of environmental qualification. Type i-i l testing of equipment used in the SEP facilities was not performed to the t E:M l axtent that it is mployed in more recent facilities. Mcwever, an evaluaticr.
- = r I
of the environmentcl qualification, frequently based on materials or partial testing of the ecuipment, has been perfor ed by the licensees in F many cases. The staff is currently reviewing these evaluations and plan-p =.
J -- p 7 :- to perform an onsite audit of the appropriate documentation for the SEP facilities to determine the adequacy and the margin of the licensees' fE I:' qualification methods. r e 'With respect to the environmental conditions associated with a postulated main steam line break accident, the staff is continuing its effort to improve the analytical models used to establish these conditions (Generic Technical Activities Program Task A-21). Particular attention is being given I to facilities without automatic containment spray initiation because the j t-operation of containment sprays decreases the duration of the peak containment temperature transient. i In addition ongoing staff reviews of operating license applications, which I l _. are being conducted in considerably more detail that in the past, are, in [ 1: general, confirming the adequacy of the qualification of equipment installed [ t in these reactors. The few instances of questionable qualification have i l f resulted in the requalification or replacement of equipment. The fact i that these more detailed reviews have generally confirmed the adequacy of qualification of equipment also contributes ot our confidence that equipment installed in operating reactors can perform'its safety function i rt an accident environment, b [ The staff actions already underway discussed above will result in the develocment of additional and more detailed information on the environmentai 6L6326 r ....L_........ e
l.. 4 r;; [:. .. [(~ i::...... l qualifications of existing safety-related electrical equipment in operating , [__. E plants, and will provide the basis for a staff judgment of the longer-term actions that are needed to increase the level of assurance of r.. the adequacy of environmental qualifications of such equipment. p
- f. :
pr i f L h;;i 7 h.' l_... =
- C6227 f
y.. k-t ~}}