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j NUCLEAR REGULATORY COMMISSION y
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- j WASHINGTON, D. C. 20555
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APR 1 0 37a MEMORANDUM FOR: Frank Schroeder, Deputy Director Division of Systems Safety FROM:
R. L. Tedesco, Assistant Director for Plant Systems, DSS f
SUBJECT:
ALAB-444 WRITEUPS FOR CATEGORY A TASK. ACTION PLANS s
We have reviewed the following subject Task Action Plan writeups as requested by your note dated April 3,1978:
A-7, A-8, A-17, A-19, A-21, A-22, A-23, A-24 A-25, A-28, A-30, A-32, A-35, A-36 and A-39.
We concur with the writeups if the changes indicated on the attached writeups are made.
Writeups attached: A-17, A-22, A-25 and A-39.
4<LO R. L. Tedesco,' Assistant Director for Plant Systeds Division of Systems Safety cc:
R. Mattson W. Minners V. Benaroya F. Rosa G. Lainas
. T. Ippolito
. J. Kudrick N. Su v J. Calvo R. Fitzpatrick J. Shapaker C. Anderson
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Z. Szukiewicz P. Baranowsky C. Long 1: G 73,!_: 19 9
l 8604010269 860114 PDR FOIA FIRESTOB5-665 PDR
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m TASK A-17 O
3.
Basis for Continued Plant Operation and Licensing Pending Completion of Task.
As discussed in Section 1, this task' addresses the development of a systematic process to review plant systems to determine their impact on other plant systems. The purpose of the task is to improve the
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licensing process. As discussed in Section 2 it is anticipated that this task will confim that current licensing requirements and pro-cedures acceptably control the potential for adverse systems interactions, even though some modifications for improvement in the review procedures and licensing requirements may be made.
Current ifcensing requirements are founded on the principle of defense in depth against credible occurrences. Adherence to this I
principle results in requirements such as physical separation and independence of redundant safety systems, and protection against events such as high energy line ruptures, missiles, high winds, flooding, seismic events, fires, operator errors, and sabotage.
N-These design provisions supplemented by the current review procedures of the Standard Review Plan (NUREG-75/087) which require interdisci-plinary reviews and which account, to a large extent, for review of potential systems interactions provide, for an adequately safe situation with respect to such interactions. The quality assurance program which is followed during the design, e
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o 2-construction, and operatianal phases for each plant is expected to provide added assurance against the potential for adverse systems
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interactions.
Plant licensing can continue pending ultimate resolution of this task because current licensing requirements provide an acceptable level of assurance against potentially adverse systems interactions. Pre-vious licensing procedures that were followed for those plants now operating also provided assurance against potentially adverse i
systems interactions, although perhaps to a lesser degree than current procedures. liewever.,Mperiencetodatehasdemonstrated that operating plants have been designed to provide reasonable 1
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assurance that adverse systems interactions will not occur.j/h
]t those instances such as fire protection and high-energy line break outside containment where the potential for adverse systems inter-action has been identified, corrective measures have been or are being i
taken on each plant to attain an acceptable level of assurance against adverse systems interaction.
In sumary, the staff considers that present plant design and review
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procedures which have been developed and refined from those procedures A' mon We.
followed for plants now in operation, provide assurance that unaccept-4 able adverse systems interactions will no occur. The results of this trisk are expected to confirm this view, although so.1e modifications
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ps review procedures,may be recomended. Accordingly, we conclude i
that while this task is being performed, continued operation and plant e
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e 3-licensing can proceed with reasonable assurance of protection to the health and safety of the public.
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m Task A-22_
10 33 sis 'sa C:ntinued Plant Oceration ard Licensine Pendino
_Cc.9tletion of Task
"-...r'catst n Ie:tior.1, several aspects of the v in Steam Line Break a
.s' 2: for ?:.?s as :urrer.tly provided iy acclic:r.ts and reviewed by the N.RC s t:f' hav.. baen questioned. Inis task is t evaluate these questions o* c:ncare.: anc c:nfirm or mo:1fy the present ositions. The concerns darive crir: : ally fr:m Issues
- an 15 o'.NUc.EG-C13S.
- see 1 cuc,: cati cre;it for tne opera:i:r. c' conss'ety-grade equipment s
as a tacku: 'er assu. ad sir.gle active fa' lures in safety-grade equipment foll:.<'ng a main steam lina break. This task will evaluate plant response seas't'vity to the oceration or nonoperacility Of various nonsafety-grade I
syste s aa. ::.ic:1ontt.
- will also develee cgU if.h ty risesscent C#
o' su:r. e af: y t.
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6 Is:ue 15 (: ::n:erned with the mecnanical res:ense of the cressure vessel folicwir.g s *c r. steam line break. Inis task will consider safety systems
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and ::erator setier.: racuested te (3}.raintain acceotable p* essure vessel s
S t.*e s s *. vel s, a'.: ft) s:nieve long. e*
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!a NL'E3-T.:!. tre s:sf' c cviceh '.ns :ases far its '. tarts ::sitic ; :n l
- nese two 's:;ts. F:r Issue 1, the sts" state: tna :res%s in.tne r
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- *::ndary.; stem. are considered to have less :>0tential for major release
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of fission :roducts than breaks in the primary system.
(Typically, doses c.-1culated for main steamline break accidents are a small fraction of 10 CF,R J2rt100 limits.) Consequently, less stringent recuirements are' imposed on t'a qusMty and casign of systems ecquired to es:e with secondary system
.: tares. This to: reach, in the staff's judgme.t.
results in a proper weight-s i.1; c' :or.sagt.encas and safety requirements ir ceder to assure a balanced l
level cf safety ever the entire spectram of pestulated design basis accidents.
}& cule :( tra :r.;cnants in t..a sac:ncary s,:te:.1 are essantial to plant operatien ce svailability, and are in a state c' c:ntinuous or frequent i
ooerction. The :ensidera:1e experience from both fossil and nuclear plant u d,ea le d c:eratice. Mas d6me.a44*e4ee a nigh reliability of such c:mponents. Aware-ness of tnis reif ability level led to tne position of permitting credit for ?:nsafaty-.;rade equiement even thougn :ocumented evidence for all systers an: cc.r:ve.ar.t: was not availatie. Th':
/*:et 1: ex;e:ted
- o 009ff *m th s,adga.cnt tnr ugn the sensitivit;. and reliability studies referred to a: eve.
Relative t: 1ssue 15b, we noted that a potential safety proolem related to reactor,es54' tr.tegrit/ does not ec:xc impor ant until the vessel u! b21r. tu.l 2ctW ts exte.ded r.eutror, ir*1d'at's#. caring plant operation.
The irraddat10n s#fett is te reduce the e* iowa',*.4 ateess st red 6ced a
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tet.oeratures late in the life of the vessel. There must also be a potential fer return to high pressure at reduced vessal temperature. This task will crovide a c:oclusion regarding this latter aspect (discussed belcw) and ili prev';& in;.t tc 7ask A-11 which includes evaluation of the effects d
- . a steamlire
- reak en vessel intagrity.
t When c:nsidaring -he saquence of conditions folle. sing a main steam line break, the :rica.y rystsm is first deprestari:ec by overcoolinc through tne secondary sy: tem. The redaction in primary syste..1 Ortssure causes a ret::cetripancactuationofthesr.argencycoreecclingsystem(ECCS).
- . res;are reductions in the primary system are acccm:ar.ied by temperature decrease wi:n snrinkage of the ifquid volume. Actuation of the ECCS realenisnes tne volume of liquid. Unless :erminated or controlled by t.9e c: erat:r, tbt IC05 c:uld ever.tually ra'ili and e: pressurize the sel.rary sys:i= t.. '. t sefety valva set po'et 7:J.s ass will evaluate the timing re:uirentn:: for ocerator actions, tre nature of the actions.
and the Itkalfheed of accomplishment and devel:o a position relative to the recuirsments for such operator actions.
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The i teri =csi:1or.2 desc*ibed cacve with regars to credit fgr nonsafety-Grade e:uir or.t are.:entinuing to be used in the rev'4w of construction pe--it*a, t operating license a:;1(cations and in 2viivating t.Me continued coeratio.9 of coerating reactors and,'or the reasons stated above, we have
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4 conchded that ooeration does not present an undue risk to the health r d safety of the pubite. The other issue addresses the need for operator action following a main steam 1tr.e break late in reactor life to assure I
vessel integrity. The informatfor. in this regard derived from this task c'11 provide inout to Task A-ll which specifically addresses reactor 5
j vessel integrity.
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m TASK A-25 NON-SAFETY LOADS ON CLASS IE POWER SOURCES 3.
Basis for Continued Plant Operation and t.icensing Pendino Completion of Task.
' As discussed in Section 1, the safety concern addressed by this task is whether or not the reif ability of Class IE power sources 1s signi-ficantly affected by allowing the sharing of these sources by loads that perform safety functions and loads that perform normal plant functions (non-safety loads).,The Class IE power sources are those on-site sources that are called upon only in the event of.a loss of the preferred off-site power.
Present regulatory practice allows the connection of non-safety loads in addition to the required saf' ty loads to Class IE power e
sources by imposing some restrictions. The results of this task will be used by the regulatory staff in evaluating construction pennit applications submitted after 1978.
It is anticipated that the results of this task will not signifi-cantly alter current licensing criteria set forth in Section 8.3.1 Nfd.
of the 4sausssmener Standard Review Plan. In fact, this task may
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yield relaxed licensing requirements that would permit circuit breakers and fuses as isolation devices', under certain conditions, between non-safety loads and Class IE power sources.
Prior to the completion of the task, current Itcensing requirements set forth in the Commission's Standard Review Plan will be t
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2-utilized to assure safe plant operation. These requirements include the automatic disconnection of non-safety loads from Class IE buses at the onset of emergency conditions such that the non-safety i
loads do not affect the ability of the Class IE power sources to supply the required safety loads.
By assessing and quantifying the reliability assured by current regulatory requirements, this task will establish firm bases for connecting non-safety loads to Class IE sources without degrading the emergency sources below an acceptable level. This task will also I
determine whether some of the current licensing criteria can be relaxed and still provide adequate protection of public health and safety.
In swnnary, it is the staff's judgment that current licensing Pw de.
requirements set forth in the "yg '- Standard Review Plan,wh4eb su gie.r. ar e ss.ne sac
---"'- : :::::- -- of the reliability of Class IE power sources eset sweets *55T to continue with plant licensing or operation pending the i
ultimate resolution of the issue addressed by the task action plan.
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TASK A-39 DETERMINATION OF SAFETY REFIEF VALVE (SRV) POOL DYNAMIC LOADS AND TEMPERATURE LIMITS FOR BWR CONTAINMENT 3.
Basis for Continued Plant Operation and Licensino Pendino Concletion [
of Task.
As discussed in Sec' tion 1, the safety concern addressed by this task is the possible damage to wetwell internal structures and the pool boundary that could occur due to air-clearing and steam quenching
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phenomena resulting from safety relief valve (SRV) discharge into the' suppression pools of BWR plants. It is of concern to all BWR plants using the Mark I, Mark !!, or $rk III pressure suppression type containments.
As discussed in Section 2, this task will provide the basis for
' establishing acceptance criteria for safety relief valve loads and for suppression pool temperature limits.
In conjunction with Task A-7 (Mark I Long Term Program) and Task A-8 (Mark !! Contain-ment Pool Dynamic Loads), a complete evaluation will be provided of suppression pool dynamic loads for BWR containments.
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For plants with Mark I containments either in uperation or not yet licensed for operation, the justification for continued operation and licensing is based on our evaluation of operating experiences taloc h dmen hu 4s l
y the plant capability to tolerate SRV loads in the short term.
SRV operating experience has shown that in all but a few instances, t
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SRV discharges have performed satisfactorily without any evidence of damage either due to the hydrodynamic loads or pool temperature effect.
In those isolated cases where localized damage has been
,encoun ere, the damage did not result in a loss of the containment t
d function, or release of radio-activity, or undue risk to the health and safety of the public.
In those cases, repairs were made and additional margin was included in ti1e structures. With respect to the plant capability, the staff has concluded that the plants have the capability to tolerate SRV loads because the loads are related to the f
structural fatigue life. However, all plants will be required to demonstrate the capability to meet the SRY loads criteria and pool temperature Ifmit which will be established by this task.
i Plants with Mark !! containments, of which there are none in operation, will be required to' demonstrate the capability of accomodating the criteria developed under this t:sk program. The lead plant for an ope' rating license is the Zimer Plant, which has a projected fuel loading date of October,1978. The completion date for the interim criteria is June, 1978. We will require the Zimer 1
Plant to meet the interim criteria before an operating license can be N
issued.
For Mark III containments, we have issued acceptance criteria for SRV with quencher device. Although we believe that the loads criteria are conservative, we will require in-plant tests for con-firma tion.
It should be noted that the criteria were established x.-
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. on the basis of information previously provided by GE. However, recent GE studies indicate that there.are some changes in the pre-viously provided information. These changes would affect the load criteria. GE, however, has proposed an. approach to modi.fy the SRV control logic such that the current load criteria can be maintained.
We have included this concern in a revision to this task action plan and have actively reviewed the GE proposed approach. Although we have not completed the review, we believe that such an approach is techni-cally feasible. Since all Mark III containments use quencher devices, the. pool temperature Ifmits will not be an area of concern on the basis of current Mark III design.
In summary, the staff concludes that the SRY, loads are related to the structural fatigue life. Therefore, we feel that the plants with. '
Mark I containment can be allowed to continue operation until comple-tion of the Mark I Long Term program.
Interim criteria will be developed for use on Mark II containments prior to the issuance of an operating license for the first U. S. plant (Zimmer) to use such a containment. For Mark III containment, we believe that the current criteria are adequately conservative if the GE proposed approach is found acceptable as a result of our evaluation. Accordingly, we conclude that while this task is being performed, continued operation and plant licensing can proceed with reasonable assurance of pro-tection to the health and safety of the public.
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